WorldWideScience

Sample records for reactor conversion program

  1. Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs

    Energy Technology Data Exchange (ETDEWEB)

    Yoder, G.L.

    2005-10-03

    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.

  2. University Reactor Conversion Lessons Learned Workshop for Purdue University Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Eric C. Woolstenhulme; Dana M. Hewit

    2008-09-01

    The Department of Energy’s Idaho National Laboratory, under its programmatic responsibility for managing the University Research Reactor Conversions, has completed the conversion of the reactor at Purdue University Reactor. With this work completed and in anticipation of other impending conversion projects, the INL convened and engaged the project participants in a structured discussion to capture the lessons learned. The lessons learned process has allowed us to capture gaps, opportunities, and good practices, drawing from the project team’s experiences. These lessons will be used to raise the standard of excellence, effectiveness, and efficiency in all future conversion projects.

  3. Light water reactor program

    Energy Technology Data Exchange (ETDEWEB)

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  4. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Neil Todreas; Pavel Hejzlar

    2008-06-30

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

  5. SNTP program reactor design

    Science.gov (United States)

    Walton, Lewis A.; Sapyta, Joseph J.

    1993-06-01

    The Space Nuclear Thermal Propulsion (SNTP) program is evaluating the feasibility of a particle bed reactor for a high-performance nuclear thermal rocket engine. Reactors operating between 500 MW and 2,000 MW will produce engine thrusts ranging from 20,000 pounds to 80,000 pounds. The optimum reactor arrangement depends on the power level desired and the intended application. The key components of the reactor have been developed and are being tested. Flow-to-power matching considerations dominate the thermal-hydraulic design of the reactor. Optimal propellant management during decay heat cooling requires a three-pronged approach. Adequate computational methods exist to perform the neutronics analysis of the reactor core. These methods have been benchmarked to critical experiment data.

  6. Conversion program in Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Jonsson, E.B. [Studsvik Nuclear AB, Nykoeping (Sweden)

    1997-08-01

    The conversion of the Swedish 50 MW R2 reactor from HEU to LEU fuel has been successfully accomplished over a 16 cycles long process. The conversion started in January 1991 with the introduction of 6 LEU assemblies in the 8*8 core. The first all LEU core was loaded in March 1993 and physics measurements were performed for the final licensing reports. A total of 142 LEU fuel assemblies have been irradiated up until September 1994 without any fuel incident. The operating licence for the R2 reactor was renewed in mid 1994 taking into account new fuel type. The Swedish Nuclear Inspectorate (SKI) pointed out one crucial problem with the LEU operation, that the back end of the LEU fuel cycle has not yet been solved. For the HEU fuel Sweden had the reprocessing alternative. The country is now relying heavily on the success of the USDOEs Off Site Fuels Policy to take back the spent fuel from the research reactors. They have in the meantime increased their intermediate storage facilities. There is, however, a limit both in time and space for storage of MTR-type of assemblies in water. The penalty of the lower thermal neutron flux in LEU cores has been reduced by improvements of the new irradiation rigs and by fine tuning the core calculations. The Studsvik code package, CASMO-SIMULATE, widely used for ICFM in LWRs has been modified to suit the compact MTR type of core.

  7. Conversion Preliminary Safety Analysis Report for the NIST Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, D. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Baek, J. S. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hanson, A. L. [Brookhaven National Lab. (BNL), Upton, NY (United States); Cheng, L-Y [Brookhaven National Lab. (BNL), Upton, NY (United States); Brown, N. [Brookhaven National Lab. (BNL), Upton, NY (United States); Cuadra, A. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2015-01-30

    The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the NIST research reactor (aka NBSR); a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in an aluminum alloy, and the development of the fabrication techniques. This report is a preliminary version of the Safety Analysis Report (SAR) that would be submitted to the U.S. Nuclear Regulatory Commission (NRC) for approval prior to conversion. The report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis in any conversion SAR is to explain the differences between the LEU and HEU cores and to show the acceptability of the new design; there is no need to repeat information regarding the current reactor that will not change upon conversion. Hence, as seen in the report, the bulk of the SAR is devoted to Chapter 4, Reactor Description, and Chapter 13, Safety Analysis.

  8. Study on water cooled high conversion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    As a part of study on advanced reactors for the future, conceptual design of high conversion water cooled reactors is being studied, aiming at the contribution to nuclear fuel cycle by the LWR technology, since the utilization of LWRs will extend over a long period of time . We are studying on the reactor core concepts for BWR and PWR reactor systems. As for BWR system, three types of reactor cores are investigating for three different design goals; long operation period, high conversion ratio and high applicability for the existing BWR system. In all the cases, we have obtained a fair prospect of a large core concept with a capacity of 1,000 MWe class having negative void reactivity coefficient. This study is a part of JAERI-JAPCO (Japan Atomic Power Company) cooperative studies. Various kinds of conceptual designs will be created until the end of FY 1999. The designs will be checked and reviewed at that time, then experimental studies on the realization of the concepts will start with further design works from FY 2000. (author)

  9. Conversational English Program, 1.

    Science.gov (United States)

    Instituto de Idiomas Yazigi, Sao Paulo (Brazil). Centro de Linguistica Aplicada.

    This first book of a conversational English program for adults contains an introductory section in Portuguese and exercises in English. The text centers around an English-speaking family from the United States that goes to live in Brazil. It contains color photographs with captions followed by exercises. The exercises are in English and involve…

  10. Conversational English Program, 2.

    Science.gov (United States)

    Instituto de Idiomas Yazigi, Sao Paulo (Brazil). Centro de Linguistica Aplicada.

    This second book of a conversational English program for adults contains an introductory section in Portuguese and exercises in English. The text centers around an English-speaking family from the United States that goes to live in Brazil. It contains color photographs with captions followed by exercises. The exercises are in English and involve…

  11. Reactor Vessel Surveillance Program for Advanced Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

    2008-10-15

    This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

  12. Direct Energy Conversion for Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brown, N.; Cooper, J.; Vogt, D.; Chapline, G.; Turchi, P.; Barbee Jr., T.; Farmer, J.

    2000-07-01

    Thermoelectric generators (TEG) are a well-established technology for compact low power output long-life applications. Solid state TEGs are the technology of choice for many space missions and have also been used in remote earth-based applications. Since TEGs have no moving parts and can be hermetically sealed, there is the potential for nuclear reactor power systems using TEGs to be safe, reliable and resistant to proliferation. Such power units would be constructed in a manner that would provide decades of maintenance-free operation, thereby minimizing the possibility of compromising the system during routine maintenance operations. It should be possible to construct an efficient direct energy conversion cascade from an appropriate combination of solid-state thermoelectric generators, with each stage in the cascade optimized for a particular range of temperature. Performance of cascaded thermoelectric devices could be further enhanced by exploitation of compositionally graded p-n couples, as well as radial elements to maximize utilization of the heat flux. The Jet Propulsion Laboratory in Pasadena has recently reported segmented unicouples that operate between 300 and 975 K and have conversion efficiencies of 15 percent [Caillat, 2000]. TEGs are used in nuclear-fueled power sources for space exploration, in power sources for the military, and in electrical generators on diesel engines. Second, there is a wide variety of TE materials applicable to a broad range of temperatures. New materials may lead to new TEG designs with improved thermoelectric properties (i.e. ZT approaching 3) and significantly higher efficiencies than in designs using currently available materials. Computational materials science (CMS) has made sufficient progress and there is promise for using these techniques to reduce the time and cost requirements to develop such new TE material combinations. Recent advances in CMS, coupled with increased computational power afforded by the Accelerated

  13. Reactor service life extension program

    Energy Technology Data Exchange (ETDEWEB)

    Caskey, G.R.; Sindelar, R.L.; Ondrejcin, R.S.; Baumann, E.W.

    1990-12-31

    A review of the Savannah River Site production reactor systems was initiated in 1980 and led to implementation of the Reactor Materials Program in 1984 to assess reactor safety and reactor service life. The program evaluated performance of the reactor tanks, primary coolant piping, and thermal shields, components of welded construction that were fabricated from Type 304 stainless steel. The structural integrity analysis of the primary coolant system has shown that the pressure boundary is not susceptible to gross rupture, including a double ended guillotine break or equivalent large area bank. Residual service life is potentially limited by two material degradation modes, irradiation damage and intergranular stress corrosion cracking. Analysis of the structural integrity of the tanks and piping has shown that continued safe operation of the reactors for several additional decades is not limited by the material performance of the primary coolant system. Although irradiation damage has not degraded material behavior to an unacceptable level, past experience has revealed serious difficulties with repair welding on irradiated stainless steel. Stress corrosion can be mitigated by newly identified limits on impurity concentrations in the coolant water and by stress mitigation of weld residual stresses. Work continues in several areas: the effects of helium on mechanical behavior of irradiated stainless steel; improved weld methods for piping and the reactor tanks; and a surveillance program to track irradiation effects on the tank walls.

  14. Reactor service life extension program

    Energy Technology Data Exchange (ETDEWEB)

    Caskey, G.R.; Sindelar, R.L.; Ondrejcin, R.S.; Baumann, E.W.

    1990-01-01

    A review of the Savannah River Site production reactor systems was initiated in 1980 and led to implementation of the Reactor Materials Program in 1984 to assess reactor safety and reactor service life. The program evaluated performance of the reactor tanks, primary coolant piping, and thermal shields, components of welded construction that were fabricated from Type 304 stainless steel. The structural integrity analysis of the primary coolant system has shown that the pressure boundary is not susceptible to gross rupture, including a double ended guillotine break or equivalent large area bank. Residual service life is potentially limited by two material degradation modes, irradiation damage and intergranular stress corrosion cracking. Analysis of the structural integrity of the tanks and piping has shown that continued safe operation of the reactors for several additional decades is not limited by the material performance of the primary coolant system. Although irradiation damage has not degraded material behavior to an unacceptable level, past experience has revealed serious difficulties with repair welding on irradiated stainless steel. Stress corrosion can be mitigated by newly identified limits on impurity concentrations in the coolant water and by stress mitigation of weld residual stresses. Work continues in several areas: the effects of helium on mechanical behavior of irradiated stainless steel; improved weld methods for piping and the reactor tanks; and a surveillance program to track irradiation effects on the tank walls.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-04-01

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

  16. Biological conversion of biomass to methane, the effect of reactor design on kinetics

    Energy Technology Data Exchange (ETDEWEB)

    Geisser, H R; Pfeffer, J T

    1977-12-01

    An experimental program was conducted to evaluate the effect of reactor type on methane production. This study showed that if a balanced population of organisms can be maintained in the initial stage, multi-stage fermentation is more efficient than a complex mix system. However, when the system is stressed, failure in the multi-stage system is more rapid. If the objective is to maximize the conversion of solids to methane, a staged system will produce more methane per unit volume of reactor. If the objective is to maximize the methane production per unit volume of reactor, a single stage complete-mix reactor operating near the critical retention time is required.

  17. Review of selected aspects of the Army Gas-Cooled Reactor Systems Program

    Energy Technology Data Exchange (ETDEWEB)

    None

    1965-08-27

    Information is presented concerning the AGCRS program; ML-1 reactor skid refurbishing program; ML-1-IM fabrication status; power conversion system component testing program; ML-1 demonstration test program; and applications of ML-1 technology.

  18. University Reactor Conversion Lessons Learned Workshop for Texas A&M University Nuclear Science Center Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Eric C. Woolstenhulme; Dana M. Meyer

    2007-04-01

    The objectives of this meeting were to capture the observations, insights, issues, concerns, and ideas of those involved in the Texas A&M University Nuclear Science Center (TAMU NSC) TRIGA Reactor Conversion so that future efforts can be conducted with greater effectiveness, efficiency, and with fewer challenges. This workshop was held in conjunction with a similar workshop for the University of Florida Reactor Conversion. Some of the generic lessons from that workshop are included in this report for completeness.

  19. Conversion of Methanogenic Substrates in Anaerobic Reactors

    OpenAIRE

    Gonzalez-Gil, G.

    2000-01-01

    The EGSB systems represents an attractive option to extend further the use of anaerobic technology for wastewater treatment, particularly with respect to waste streams originating from chemical industries. Frequently chemical waste streams are unbalanced with respect to nutrients and/or micronutrients and furthermore these streams may contain toxic-biodegradable compounds. To reduce toxicity high recycle ratios may be applied as in the case of EGSB reactors however, this at the same time may ...

  20. Conversion of Biomass Syngas to DME Using a Microchannel Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Jianli; Wang, Yong; Cao, Chunshe; Elliott, Douglas C.; Stevens, Don J.; White, James F.

    2005-03-01

    The capability of a microchannel reactor for direct synthesis of dimethylether (DME) from biomass syngas was explored. The reactor was operated in conjunction with a hybrid catalyst system consisting of methanol synthesis and dehydration catalysts, and the influence of reaction parameters on syngas conversion was investigated. The activities of different dehydration catalysts were compared under DME synthesis conditions. Reaction temperature and pressure exhibited similar positive effects on DME formation. A catalytic stability test of the hybrid catalyst system was performed for 880 hours, during which CO conversion only decreased from 88% to 81%. In the microchannel reactor, the catalyst deactivation rate appeared to be much slower than in a tubular fixed-bed reactor tested for comparison. Test results also indicated that the dehydration reaction rate and the water depletion rate via a water-gas-shift reaction should be compatible in order to achieve high selectivity to DME. Using the microchannel reactor, it was possible to achieve a space time yield almost three times higher than commercially demonstrated performance results. A side-by-side comparison indicated that the heat removal capability of the microchannel reactor was at least six times greater than that of a commercial slurry reactor under similar reaction conditions.

  1. Methanol conversion in high-rate anaerobic reactors

    NARCIS (Netherlands)

    Weijma, J.; Stams, A.J.M.

    2001-01-01

    An overview on methanol conversion in high-rate anaerobic reactors is presented, with the focus on technological as well as microbiological aspects. The simple C1-compound methanol can be degraded anaerobically in a complex way, in which methanogens, sulfate reducing bacteria and homoacetogens

  2. Energy Conversion and Storage Program

    Energy Technology Data Exchange (ETDEWEB)

    Cairns, E.J.

    1992-03-01

    The Energy Conversion and Storage Program applies chemistry and materials science principles to solve problems in (1) production of new synthetic fuels, (2) development of high-performance rechargeable batteries and fuel cells, (3) development of advanced thermochemical processes for energy conversion, (4) characterization of complex chemical processes, and (5) application of novel materials for energy conversion and transmission. Projects focus on transport-process principles, chemical kinetics, thermodynamics, separation processes, organic and physical chemistry, novel materials, and advanced methods of analysis. Electrochemistry research aims to develop advanced power systems for electric vehicle and stationary energy storage applications. Topics include identification of new electrochemical couples for advanced rechargeable batteries, improvements in battery and fuel-cell materials, and the establishment of engineering principles applicable to electrochemical energy storage and conversion. Chemical Applications research includes topics such as separations, catalysis, fuels, and chemical analyses. Included in this program area are projects to develop improved, energy-efficient methods for processing waste streams from synfuel plants and coal gasifiers. Other research projects seek to identify and characterize the constituents of liquid fuel-system streams and to devise energy-efficient means for their separation. Materials Applications research includes the evaluation of the properties of advanced materials, as well as the development of novel preparation techniques. For example, the use of advanced techniques, such as sputtering and laser ablation, are being used to produce high-temperature superconducting films.

  3. Technical basis in support of the conversion of the University of Missouri Research Reactor (MURR) core from highly-enriched to low-enriched uranium - core neutron physics

    Energy Technology Data Exchange (ETDEWEB)

    Stillman, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Foyto, L [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Kutikkad, K [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; McKibben, J C [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Peters, N. [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Stevens, J. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2012-09-01

    This report contains the results of reactor design and performance for conversion of the University of Missouri Research Reactor (MURR) from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL) and the MURR Facility. The core conversion to LEU is being performed with financial support of the U. S. government.

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

    Energy Technology Data Exchange (ETDEWEB)

    Pablo Rubiolo, Principal Investigator

    2003-03-21

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

  5. NCSU reactor sharing program. Final technical report

    Energy Technology Data Exchange (ETDEWEB)

    Perez, P.B.

    1997-01-10

    The Nuclear Reactor Program at North Carolina State University provides the PULSTAR Research Reactor and associated facilities to eligible institutions with support, in part, from the Department of Energy Reactor Sharing Program. Participation in the NCSU Reactor Sharing Program continues to increase steadily with visitors ranging from advance high school physics and chemistry students to Ph.D. level research from neighboring universities. This report is the Final Technical Report for the DOE award reference number DE-FG05-95NE38136 which covers the period September 30, 1995 through September 30, 1996.

  6. Effectiveness Factors and Conversion in a Biocatalytic Membrane Reactor.

    Science.gov (United States)

    Godongwana, Buntu

    2016-01-01

    Analytical expressions of the effectiveness factor of a biocatalytic membrane reactor, and its asymptote as the Thiele modulus becomes large, are presented. The evaluation of the effectiveness factor is based on the solution of the governing equations for solute transport in the two regions of the reactor, i.e. the lumen and the matrix (with the biofilm immobilized in the matrix). The lumen solution accounts for both axial diffusion and radial convective flow, while the matrix solution is based on Robin-type boundary conditions. The effectiveness factor is shown to be a function of the Thiele modulus, the partition coefficient, the Sherwood number, the Peclet number, and membrane thickness. Three regions of Thiele moduli are defined in the effectiveness factor graphs. These correspond with reaction rate limited, internal-diffusion limited, and external mass transfer limited solute transport. Radial convective flows were shown to only improve the effectiveness factor in the region of internal diffusion limitation. The assumption of first order kinetics is shown to be applicable only in the Thiele modulus regions of internal and external mass transfer limitation. An iteration scheme is also presented for estimating the effectiveness factor when the solute fractional conversion is known. The model is validated with experimental data from a membrane gradostat reactor immobilised with Phanerochaete chrysosporium for the production of lignin and manganese peroxidases. The developed model and experimental data allow for the determination of the Thiele modulus at which the effectiveness factor and fractional conversion are optimal.

  7. Biomass thermochemical conversion program. 1985 annual report

    Energy Technology Data Exchange (ETDEWEB)

    Schiefelbein, G.F.; Stevens, D.J.; Gerber, M.A.

    1986-01-01

    Wood and crop residues constitute a vast majority of the biomass feedstocks available for conversion, and thermochemical processes are well suited for conversion of these materials. The US Department of Energy (DOE) is sponsoring research on this conversion technology for renewable energy through its Biomass Thermochemical Conversion Program. The Program is part of DOE's Biofuels and Municipal Waste Technology Division, Office of Renewable Technologies. This report briefly describes the Thermochemical Conversion Program structure and summarizes the activities and major accomplishments during fiscal year 1985. 32 figs., 4 tabs.

  8. A reload and startup plan for conversion of the NIST research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, D. J. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2016-03-31

    The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts. The reload portion of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.

  9. Fostering multiculturalism: the English Conversation Pairs Program.

    Science.gov (United States)

    Smith, B E; Elander, E

    1995-01-01

    Exposure to and personal interactions with people of diverse cultures foster an acceptance of multiculturalism. To this end, the English Conversation Pairs Program provides weekly conversation sessions between students who are native speakers of English (NSE) and students who speak English as a second language (ESL). Attitudes of twenty-two NSE students are surveyed before and after participation in the Conversation Pairs Program. This paper discusses (a) NSE students' perceptions of the program, (b) perceived changes in cultural sensitivity of NSE students, (c) activities which enhance and/or hamper program effectiveness, and (d) variables which influence program effectiveness.

  10. The DOE advanced gas reactor fuel development and qualification program

    Science.gov (United States)

    Petti, David; Maki, John; Hunn, John; Pappano, Pete; Barnes, Charles; Saurwein, John; Nagley, Scott; Kendall, Jim; Hobbins, Richard

    2010-09-01

    The high outlet temperatures and high thermal-energy conversion efficiency of modular high-temperature gas-cooled reactors (HTGRs) enable an efficient and cost-effective integration of the reactor system with non-electricity-generation applications, such as process heat and/or hydrogen production, for the many petrochemical and other industrial processes that require temperatures between 300°C and 900°C. The U.S. Department of Energy (DOE) has selected the HTGR concept for the Next Generation Nuclear Plant (NGNP) Project as a transformative application of nuclear energy that will demonstrate emissions-free nuclear-assisted electricity, process heat, and hydrogen production, thereby reducing greenhouse-gas emissions and enhancing energy security. The objective of the DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification program is to qualify tristructural isotropic (TRISO)-coated particle fuel for use in HTGRs. An overview of the program and recent progress is presented.

  11. NASA-OAST photovoltaic energy conversion program

    Science.gov (United States)

    Mullin, J. P.; Loria, J. C.

    1984-01-01

    The NASA program in photovoltaic energy conversion research is discussed. Solar cells, solar arrays, gallium arsenides, space station and spacecraft power supplies, and state of the art devices are discussed.

  12. Advanced Reactors Transition Program Resource Loaded Schedule

    Energy Technology Data Exchange (ETDEWEB)

    GANTT, D.A.

    2000-01-12

    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.

  13. Advanced Reactors Transition Program Resource Loaded Schedule

    Energy Technology Data Exchange (ETDEWEB)

    BOWEN, W.W.

    1999-11-08

    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.

  14. Neutronic feasibility studies for LEU conversion of the HFR Petten Reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Hanan, N. A.; Deen, J. R.; Matos, J. E.

    2000-09-26

    Design and safety analyses to determine an optimum LEU fuel assembly design using U{sub 3}Si{sub 2}-Al fuel with up to 4.8 g/cm{sup 3} for conversion of the HFR Petten reactor were performed by the RERTR program in cooperation with the Joint Research Centre and NRG. Credibility of the calculational methods and models were established by comparing calculations with recent measurements by NRG for a core configuration setup for this purpose. This model and methodology were then used to study various LEU fissile loading and burnable poison options that would satisfy specific design criteria.

  15. Accident Analyses for Conversion of the University of Missouri Research Reactor (MURR) from Highly-Enriched to Low-Enriched Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Stillman, J. A. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Div., Research and Test Reactor Dept.; Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Div., Research and Test Reactor Dept.; Wilson, E. H. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Div., Research and Test Reactor Dept.; Foyto, L. P. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; Kutikkad, K. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; McKibben, J. C. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; Peters, N. J. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; Cowherd, W. M. [Univ. of Missouri, Columbia, MO (United States). College of Engineering, Nuclear Engineering Program; Rickman, B. [Univ. of Missouri, Columbia, MO (United States). College of Engineering, Nuclear Engineering Program

    2014-12-01

    This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. In the framework of non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context most research and test reactors, both domestic and international, have started a program of conversion to the use of LEU fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MURR. This report presents the results of a study of core behavior under a set of accident conditions for MURR cores fueled with HEU U-Alx dispersion fuel or LEU monolithic U-Mo alloy fuel with 10 wt% Mo

  16. Low-Enriched Uranium Fuel Conversion Activities for the High Flux Isotope Reactor, Annual Report for FY 2011

    Energy Technology Data Exchange (ETDEWEB)

    Renfro, David G [ORNL; Cook, David Howard [ORNL; Freels, James D [ORNL; Griffin, Frederick P [ORNL; Ilas, Germina [ORNL; Sease, John D [ORNL; Chandler, David [ORNL

    2012-03-01

    This report describes progress made during FY11 in ORNL activities to support converting the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum (UMo) alloy. With both radial and axial contouring of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current levels achieved with HEU fuel. Studies are continuing to demonstrate that the fuel thermal safety margins can be preserved following conversion. Studies are also continuing to update other aspects of the reactor steady state operation and accident response for the effects of fuel conversion. Technical input has been provided to Oregon State University in support of their hydraulic testing program. The HFIR conversion schedule was revised and provided to the GTRI program. In addition to HFIR conversion activities, technical support was provided directly to the Fuel Fabrication Capability program manager.

  17. Investigation of the low enrichment conversion of the Texas A and M Nuclear Science Center Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Reuscher, J.A.

    1988-01-01

    The use of highly enriched uranium as a fuel research reactors is of concern due to the possibility of diversion for nuclear weapons applications. The Texas A M TRIGA reactor currently uses 70% enriched uranium in a FLIP (Fuel Life Improvement Program) fuel element manufactured by General Atomics. Thus fuel also contains 1.5 weight percent of erbium as a burnable poison to prolong useful core life. US university reactors that use highly enriched uranium will be required to covert to 20% or less enrichment to satisfy Nuclear Regulatory Commission requirements for the next core loading if the fuel is available. This investigation examined the feasibility of a material alternate to uranium-zirconium hydride for LEU conversion of a TRIGA reactor. This material is a beryllium oxide uranium dioxide based fuel. The theoretical aspects of core physics analyses were examined to assess the potential advantages of the alternative fuel. A basic model was developed for the existing core configuration since it is desired to use the present fuel element grid for the replacement core. The computing approach was calibrated to the present core and then applied to a core of BeO-UO{sub 2} fuel elements. Further calculations were performed for the General Atomics TRIGA low-enriched uranium zirconium hydride fuel.

  18. The Sloping Land Conversion Program in China

    DEFF Research Database (Denmark)

    Liu, Zhen

    conversion program. Our results show that SLCP works as a valid external policy intervention on rural livelihood diversification. In addition, the findings demonstrate that there exist heterogeneous effects of SLCP implementation on livelihood diversification across different rural income groups. The lower......By overcoming the barriers that limit access to financial liquidity and human resource, the Sloping Land Conversion Program (SLCP) can promote rural livelihood diversification. This paper examines this effect using a household survey data set spanning the 1999 implementation of the Sloping land...... income group was more affected by the program in terms of income diversification....

  19. Energy conversion & storage program. 1994 annual report

    Energy Technology Data Exchange (ETDEWEB)

    Cairns, E.J.

    1995-04-01

    The Energy Conversion and Storage Program investigates state-of-the-art electrochemistry, chemistry, and materials science technologies for: (1) development of high-performance rechargeable batteries and fuel cells; (2) development of high-efficiency thermochemical processes for energy conversion; (3) characterization of complex chemical processes and chemical species; (4) study and application of novel materials for energy conversion and transmission. Research projects focus on transport process principles, chemical kinetics, thermodynamics, separation processes, organic and physical chemistry, novel materials, and advanced methods of analysis.

  20. Energy Conversion & Storage Program, 1993 annual report

    Energy Technology Data Exchange (ETDEWEB)

    Cairns, E.J.

    1994-06-01

    The Energy Conversion and Storage Program applies chemistry and materials science principles to solve problems in: production of new synthetic fuels; development of high-performance rechargeable batteries and fuel cells; development of high-efficiency thermochemical processes for energy conversion; characterization of complex chemical processes and chemical species; and the study and application of novel materials for energy conversion and transmission. Projects focus on transport-process principles, chemical kinetics, thermodynamics, chemical kinetics, thermodynamics, separation processes, organic and physical chemistry, novel materials, and advanced methods of analysis.

  1. Conversion of methanol to gasoline-range hydrocarbons in a ZSM-5 coated monolithic reactor

    Energy Technology Data Exchange (ETDEWEB)

    Antia, J.E.; Govind, R. (Univ. of Cincinnati, OH (United States). Dept. of Chemical Engineering)

    1995-01-01

    Novel reactor configurations featuring catalysts supported on monolithic or honeycomb structures are being increasingly used for a number of applications. In this work, a zeolite-coated monolithic reactor is employed for the conversion of methanol to gasoline-range hydrocarbons. Experimental results show that the conversion and hydrocarbon product distribution compare favorably with data reported for fixed and fluid beds. Mathematical modeling shows that the conversion here is controlled by diffusion in the molecule-sized intracrystalline pores of the zeolite structure. This finding is of considerable important because it demonstrates that monolithic reactors are well-suited to zeolite-based catalytic processes.

  2. DIRECT ENERGY CONVERSION (DEC) FISSION REACTORS - A U.S. NERI PROJECT

    Energy Technology Data Exchange (ETDEWEB)

    D. BELLER; G. POLANSKY; ET AL

    2000-11-01

    The direct conversion of the electrical energy of charged fission fragments was examined early in the nuclear reactor era, and the first theoretical treatment appeared in the literature in 1957. Most of the experiments conducted during the next ten years to investigate fission fragment direct energy conversion (DEC) were for understanding the nature and control of the charged particles. These experiments verified fundamental physics and identified a number of specific problem areas, but also demonstrated a number of technical challenges that limited DEC performance. Because DEC was insufficient for practical applications, by the late 1960s most R&D ceased in the US. Sporadic interest in the concept appears in the literature until this day, but there have been no recent programs to develop the technology. This has changed with the Nuclear Energy Research Initiative that was funded by the U.S. Congress in 1999. Most of the previous concepts were based on a fission electric cell known as a triode, where a central cathode is coated with a thin layer of nuclear fuel. A fission fragment that leaves the cathode with high kinetic energy and a large positive charge is decelerated as it approaches the anode by a charge differential of several million volts, it then deposits its charge in the anode after its kinetic energy is exhausted. Large numbers of low energy electrons leave the cathode with each fission fragment; they are suppressed by negatively biased on grid wires or by magnetic fields. Other concepts include magnetic collimators and quasi-direct magnetohydrodynamic generation (steady flow or pulsed). We present the basic principles of DEC fission reactors, review the previous research, discuss problem areas in detail and identify technological developments of the last 30 years relevant to overcoming these obstacles. A prognosis for future development of direct energy conversion fission reactors will be presented.

  3. Testing of an advanced thermochemical conversion reactor system

    Science.gov (United States)

    1990-01-01

    This report presents the results of work conducted by MTCI to verify and confirm experimentally the ability of the MTCI gasification process to effectively generate a high-quality, medium-Btu gas from a wider variety of feedstock and waste than that attainable in air-blown, direct gasification systems. The system's overall simplicity, due to the compact nature of the pulse combustor, and the high heat transfer rates attainable within the pulsating flow resonance tubes, provide a decided and near-term potential economic advantage for the MTCI indirect gasification system. The primary objective 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.

  4. Testing of an advanced thermochemical conversion reactor system

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

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

  5. High conversion pressurized water reactor with boiling channels

    Energy Technology Data Exchange (ETDEWEB)

    Margulis, M., E-mail: maratm@post.bgu.ac.il [The Unit of Nuclear Engineering, Ben Gurion University of the Negev, POB 653, Beer Sheva 84105 (Israel); Shwageraus, E., E-mail: es607@cam.ac.uk [Department of Engineering, University of Cambridge, CB2 1PZ Cambridge (United Kingdom)

    2015-10-15

    Highlights: • Conceptual design of partially boiling PWR core was proposed and studied. • Self-sustainable Th–{sup 233}U fuel cycle was utilized in this study. • Seed-blanket fuel assembly lattice optimization was performed. • A coupled Monte Carlo, fuel depletion and thermal-hydraulics studies were carried out. • Thermal–hydraulic analysis assured that the design matches imposed safety constraints. - Abstract: Parametric studies have been performed on a seed-blanket Th–{sup 233}U fuel configuration in a pressurized water reactor (PWR) with boiling channels to achieve high conversion ratio. Previous studies on seed-blanket concepts suggested substantial reduction in the core power density is needed in order to operate under nominal PWR system conditions. Boiling flow regime in the seed region allows more heat to be removed for a given coolant mass flow rate, which in turn, may potentially allow increasing the power density of the core. In addition, reduced moderation improves the breeding performance. A two-dimensional design optimization study was carried out with BOXER and SERPENT codes in order to determine the most attractive fuel assembly configuration that would ensure breeding. Effects of various parameters, such as void fraction, blanket fuel form, number of seed pins and their dimensions, on the conversion ratio were examined. The obtained results, for which the power density was set to be 104 W/cm{sup 3}, created a map of potentially feasible designs. It was found that several options have the potential to achieve end of life fissile inventory ratio above unity, which implies potential feasibility of a self-sustainable Thorium fuel cycle in PWRs without significant reduction in the core power density. Finally, a preliminary three-dimensional coupled neutronic and thermal–hydraulic analysis for a single seed-blanket fuel assembly was performed. The results indicate that axial void distribution changes drastically with burnup. Therefore

  6. Irradiation Tests Supporting LEU Conversion of Very High Power Research Reactors in the US

    Energy Technology Data Exchange (ETDEWEB)

    Woolstenhulme, N. E.; Cole, J. I.; Glagolenko, I.; Holdaway, K. K.; Housley, G. K.; Rabin, B. H.

    2016-10-01

    The US fuel development team is developing a high density uranium-molybdenum alloy monolithic fuel to enable conversion of five high-power research reactors. Previous irradiation tests have demonstrated promising behavior for this fuel design. A series of future irradiation tests will enable selection of final fuel fabrication process and provide data to qualify the fuel at moderately-high power conditions for use in three of these five reactors. The remaining two reactors, namely the Advanced Test Reactor and High Flux Isotope Reactor, require additional irradiation tests to develop and demonstrate the fuel’s performance with even higher power conditions, complex design features, and other unique conditions. This paper reviews the program’s current irradiation testing plans for these moderately-high irradiation conditions and presents conceptual testing strategies to illustrate how subsequent irradiation tests will build upon this initial data package to enable conversion of these two very-high power research reactors.

  7. The Sloping Land Conversion Program in China

    DEFF Research Database (Denmark)

    Liu, Zhen

    By overcoming the barriers that limit access to financial liquidity and human resource, the Sloping Land Conversion Program (SLCP) can promote rural livelihood diversification. This paper examines this effect using a household survey data set spanning the 1999 implementation of the Sloping land...

  8. Development of a Reactor Model for Chemical Conversion of Lunar Regolith

    Science.gov (United States)

    Hegde, U.; Balasubramaniam, R.; Gokoglu, S.

    2009-01-01

    Lunar regolith will be used for a variety of purposes such as oxygen and propellant production and manufacture of various materials. The design and development of chemical conversion reactors for processing lunar regolith will require an understanding of the coupling among the chemical, mass and energy transport processes occurring at the length and time scales of the overall reactor with those occurring at the corresponding scales of the regolith particles. To this end, a coupled transport model is developed using, as an example, the reduction of ilmenite-containing regolith by a continuous flow of hydrogen in a flow-through reactor. The ilmenite conversion occurs on the surface and within the regolith particles. As the ilmenite reduction proceeds, the hydrogen in the reactor is consumed, and this, in turn, affects the conversion rate of the ilmenite in the particles. Several important quantities are identified as a result of the analysis. Reactor scale parameters include the void fraction (i.e., the fraction of the reactor volume not occupied by the regolith particles) and the residence time of hydrogen in the reactor. Particle scale quantities include the time for hydrogen to diffuse into the pores of the regolith particles and the chemical reaction time. The paper investigates the relationships between these quantities and their impact on the regolith conversion. Application of the model to various chemical reactor types, such as fluidized-bed, packed-bed, and rotary-bed configurations, are discussed.

  9. SAFARI-1: adjusting priorities during the LEU conversion program

    Energy Technology Data Exchange (ETDEWEB)

    Piani, C.S.B. [SAFARI-1 Research Reactor, South African Nuclear Energy Corporation (Necsa) (South Africa)

    2007-07-01

    In July 2005, the South African Department of Minerals and Energy authorised the conversion to Low Enriched Uranium (LEU) of the South African Research Reactor (SAFARI-1) and the associated fuel manufacturing at Pelindaba. At that stage the proposed scheduling allowed approximately three years for the full conversion of the reactor, anticipating simultaneous manufacturing ability from the fuel production plant. Initial priorities and regulatory agreements were allocated with the intention to manufacture and produce two Lead Test Assemblies (LTAs) from the Pelindaba plant (Phase I) and use these as qualification of manufacturer as well as initiation of the SAFARI-1 conversion (Phase II). Delays in the demonstration of sufficient confidence in the manufacturing ability to enable local fuel licensing and qualification have resulted in minor readjustments of these Phases. Delays in the initial schedule that allowed for the insertion of the two South African LTAs during the first quarter of 2006 were preempted by the acquisition of 2 LEU silicide elements of South African design manufactured by AREVA-CERCA. These two LTAs are currently undergoing testing in SAFARI-1 and have to-date completed up to 8 cycles of irradiation. As a further precaution to the potential delays in the fuel-manufacturing Phase, a reload (760 plates) of LEU silicide element fuel plates were purchased and will be assembled locally to enable the SAFARI-1 conversion program to continue according to schedule. This paper will trace the developments of the above in order to reflect the current status and the planned correlation of the Phase I and Phase II programs according to latest expectations. (author)

  10. Thermophotovoltaic Energy Conversion in Space Nuclear Reactor Power Systems

    Science.gov (United States)

    2004-12-01

    contrasted with nuclear thermal rockets which use the heat from a nuclear fission reactor to heat propellant to provide rocket thrust and radioisotope...K. Note that the highest temperature (2550 K by the Pewee reactor) was for a nuclear thermal rocket application and has the shortest duration (40 min

  11. Conversion of hydrocarbon fuel in thermal protection reactors of hypersonic aircraft

    Science.gov (United States)

    Kuranov, A. L.; Mikhaylov, A. M.; Korabelnikov, A. V.

    2016-07-01

    Thermal protection of heat-stressed surfaces of a high-speed vehicle flying in dense layers of atmosphere is one of the topical issues. Not of a less importance is also the problem of hydrocarbon fuel combustion in a supersonic air flow. In the concept under development, it is supposed that in the most high-stressed parts of airframe and engine, catalytic thermochemical reactors will be installed, wherein highly endothermic processes of steam conversion of hydrocarbon fuel take place. Simultaneously with heat absorption, hydrogen generation will occur in the reactors. This paper presents the results of a study of conversion of hydrocarbon fuel in a slit reactor.

  12. Energy conversion & storage program. 1995 annual report

    Energy Technology Data Exchange (ETDEWEB)

    Cairns, E.J.

    1996-06-01

    The 1995 annual report discusses laboratory activities in the Energy Conversion and Storage (EC&S) Program. The report is divided into three categories: electrochemistry, chemical applications, and material applications. Research performed in each category during 1995 is described. Specific research topics relate to the development of high-performance rechargeable batteries and fuel cells, the development of high-efficiency thermochemical processes for energy conversion, the characterization of new chemical processes and complex chemical species, and the study and application of novel materials related to energy conversion and transmission. Research projects focus on transport-process principles, chemical kinetics, thermodynamics, separation processes, organic and physical chemistry, novel materials and deposition technologies, and advanced methods of analysis.

  13. Preliminary Accident Analyses for Conversion of the Massachusetts Institute of Technology Reactor (MITR) from Highly Enriched to Low Enriched Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Dunn, Floyd E. [Argonne National Lab. (ANL), Argonne, IL (United States); Olson, Arne P. [Argonne National Lab. (ANL), Argonne, IL (United States); Wilson, Erik H. [Argonne National Lab. (ANL), Argonne, IL (United States); Sun, Kaichao S. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Newton, Jr., Thomas H. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Hu, Lin-wen [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    2013-09-30

    The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context most research and test reactors, both domestic and international, have started a program of conversion to the use of LEU fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MITR. This report presents the preliminary accident analyses for MITR cores fueled with LEU monolithic U-Mo alloy fuel with 10 wt% Mo. Preliminary results demonstrate adequate performance, including thermal margin to expected safety limits, for the LEU accident scenarios analyzed.

  14. Methane conversion into higher hydrocarbons with dielectric barrier discharge micro-plasma reactor

    Institute of Scientific and Technical Information of China (English)

    Baowei; Wang; Wenjuan; Yan; Wenjie; Ge; Xiaofei; Duan

    2013-01-01

    We reported a coaxial,micro-dielectric barrier discharge(micro-DBD)reactor and a conventional DBD reactor for the direct conversion of methane into higher hydrocarbons at atmospheric pressure.The effects of input power,residence time,discharge gap and external electrode length were investigated for methane conversion and product selectivity.We found the conversion of methane in a micro-DBD reactor was higher than that in a conventional DBD reactor.And at an input power of 25.0 W,the conversion of methane and the total C2+C3 selectivity reached 25.10% and 80.27%,respectively,with a micro-DBD reactor of 0.4 mm discharge gap.Finally,a nonlinear multiple regression model was used to study the correlations between both methane conversion and product selectivity and various system variables.The calculated data were obtained using SPSS 12.0 software.The regression analysis illustrated the correlations between system variables and both methane conversion and product selectivity.

  15. A generalized perturbation program for CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Yang, Won Sik [Chosun University, Kwangju (Korea, Republic of)

    1998-12-31

    A generalized perturbation program has been developed for the purpose of estimating zonal power variation of a CANDU reactor upon refueling operation. The forward and adjoint calculation modules of RFSP code were used to construct the generalized perturbation program. The numerical algorithm for the generalized adjoint flux calculation was verified by comparing the zone power estimates upon refueling with those of forward calculation. It was, however, noticed that the truncation error from the iteration process of the generalized adjoint flux is not negligible. 2 refs., 1 figs., 1 tab. (Author)

  16. Accident Analyses for Conversion of the University of Missouri Research Reactor (MURR) from Highly-Enriched to Low-Enriched Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Stillman, J. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jaluvka, D. [Argonne National Lab. (ANL), Argonne, IL (United States); Wilson, E. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Foyto, L. P. [Univ. of Missouri, Columbia, MO (United States); Kutikkad, K. [Univ. of Missouri, Columbia, MO (United States); McKibben, J. C. [Univ. of Missouri, Columbia, MO (United States); Peters, N. J. [Univ. of Missouri, Columbia, MO (United States)

    2017-02-01

    This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members in the Research and Test Reactor Department at the Argonne National Laboratory (ANL) and the MURR Facility. MURR LEU conversion is part of an overall effort to develop and qualify high-density fuel within the U.S. High Performance Research Reactor Conversion (USHPRR) program conducted by the U.S. Department of Energy National Nuclear Security Administration’s Office of Material Management and Minimization (M3).

  17. University Reactor Conversion Lessons Learned Workshop for the University of Florida

    Energy Technology Data Exchange (ETDEWEB)

    Eric C. Woolstenhulme; Dana M. Meyer

    2007-04-01

    The Department of Energy’s (DOE) Idaho National Laboratory (INL), under its programmatic responsibility for managing the University Research Reactor Conversions, has completed the conversion of the reactor at the University of Florida. This project was successfully completed through an integrated and collaborative effort involving the INL, Argonne National Laboratory (ANL), DOE (Headquarters and Field Office), the Nuclear Regulatory Commission, the Universities, and contractors involved in analyses, fuel design and fabrication, and SNF shipping and disposition. With the work completed with these two universities, and in anticipation of other impending conversion projects, INL convened and engaged the project participants in a structured discussion to capture lessons learned. The objectives of this meeting were to capture the observations, insights, issues, concerns, and ideas of those involved in the reactor conversions so that future efforts can be conducted with greater effectiveness, efficiency, and with fewer challenges.

  18. The sloping land conversion program in China

    DEFF Research Database (Denmark)

    Liu, Zhen; Lan, Jing

    2015-01-01

    Through addressing the motivations behind rural households’ livelihood diversification, this paper examines the effect of the Sloping Land Conversion Program (SLCP) on livelihood diversification using a longitudinal household survey data set spanning the overall implementation of the SLCP. Our...... results show that the SLCP works as a valid external policy intervention to increase rural livelihood diversification. In addition, the findings demonstrate that the implementation of the SLCP has had heterogeneous effects on livelihood diversification across different rural income groups. The lower...... income group was more affected by the program in terms of income diversification....

  19. CO2 conversion by plasma technology: insights from modeling the plasma chemistry and plasma reactor design

    Science.gov (United States)

    Bogaerts, A.; Berthelot, A.; Heijkers, S.; Kolev, St.; Snoeckx, R.; Sun, S.; Trenchev, G.; Van Laer, K.; Wang, W.

    2017-06-01

    In recent years there has been growing interest in the use of plasma technology for CO2 conversion. To improve this application, a good insight into the underlying mechanisms is of great importance. This can be obtained from modeling the detailed plasma chemistry in order to understand the chemical reaction pathways leading to CO2 conversion (either in pure form or mixed with another gas). Moreover, in practice, several plasma reactor types are being investigated for CO2 conversion, so in addition it is essential to be able to model these reactor geometries so that their design can be improved, and the most energy efficient CO2 conversion can be achieved. Modeling the detailed plasma chemistry of CO2 conversion in complex reactors is, however, very time-consuming. This problem can be overcome by using a combination of two different types of model: 0D chemical reaction kinetics models are very suitable for describing the detailed plasma chemistry, while the characteristic features of different reactor geometries can be studied by 2D or 3D fluid models. In the first instance the latter can be developed in argon or helium with a simple chemistry to limit the calculation time; however, the ultimate aim is to implement the more complex CO2 chemistry in these models. In the present paper, examples will be given of both the 0D plasma chemistry models and the 2D and 3D fluid models for the most common plasma reactors used for CO2 conversion in order to emphasize the complementarity of both approaches. Furthermore, based on the modeling insights, the paper discusses the possibilities and limitations of plasma-based CO2 conversion in different types of plasma reactors, as well as what is needed to make further progress in this field.

  20. Conversion regular patterns of acetic acid,propionic acid and butyric acid in UASB reactor

    Institute of Scientific and Technical Information of China (English)

    LIU Min; REN Nan-qi; CHEN Ying; ZHU Wen-fang; DING Jie

    2004-01-01

    On the basis of continuous tests and batch tests, conversion regular patterns of acetate, propionate and butyrate in activated sludge at different heights of the UASB reactor were conducted. Results indicated that the conversion capacity of the microbial is decided by the substrate characteristic when sole VFA is used as the only substrate. But when mixed substrates are used,the conversion regulations would have changed accordingly. Relationships of different substrates vary according to their locations. In the whole reactor, propionate's conversion is restrained by acetate and butyrate of high concentration. On the top and at the bottom of the reactor, conversion of acetate, but butyrate, is restrained by propionate. And in the midst, acetate's conversion is accelerated by propionate while that of butyrate is restrained. It is proved, based on the analysis of specific conversion rate, that the space distribution of the microbe is the main factor that affects substrates' conversion. The ethanol-type fermentation of the acidogenic-phase is the optimal acid-type fermentation for the two-phase anaerobic process.

  1. Kinetics of propionate conversion in anaerobic continuously stirred tank reactors

    DEFF Research Database (Denmark)

    Bangsø Nielsen, Henrik; Mladenovska, Zuzana; Ahring, Birgitte Kiær

    2008-01-01

    (max), and the half saturation constant, K-m, were initially estimated by applying the integrated Michaelis-Menten equation. A(max) was in the range from 22.8 to 29.1 mu mol gVS(-1) h(-1) while K-m, was in the range from 0.46-0.95 mM. In general, A(max) gave a good reflection of the reactor performances. Secondly...

  2. Disk magnetohydrodynamic power conversion system for NERVA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jackson, W.D. (HMJ Corporation. 10400 Connecticut Ave., Kensington, Maryland 20895 (United States)); Bernard, F.E. (Westinghouse Corp., P.O. Box 355, Pittsburgh, Pennsylvania 15230 (United States)); Holman, R.R. (HMJ Corporation, 10400 Connecticut Ave., Kensington, Maryland 20895 (United States)); Maxwell, C.D. (STD Research Corp., P.O. Box C, Arcadia, California 91006 (United States)); Seikel, G.R. (SeiTec, Inc., P.O. Box 81264, Cleveland, Ohio 44181 (United States))

    1993-01-15

    The combination of a magnetohydrodynamic (MHD) generator of the disk type with a NERVA reactor yields an advanced power system particularly suited to space applications with the capability of producing up to gigawatt pulses and multi-megawatt continuous operation. Several unique features result from the combination of this type of reactor and a disk MHD generator in which hydrogen serves as the plasma working fluid. Cesium seedings is utilized under conditions which enable the generator to operate stably in the non-equilibrium electrical conduction mode. In common with all practical MHD generators, the disk output is DC and voltages in the range 20--100 kV are attainable. This leads to a simplification of the power conditioning system and a major reduction in specific mass. Taken together with the high performance capabilities of the NERVA reactor, the result is an attractively low overall system specific mass. Further, the use of non-equilibrium ionization enables system specific enthalpy extractions in excess of 40% to be attained. This paper reports the results of a study to establish the basis for the design of a cesium seeded hydrogen MHD disk generator. Generator performance results are presented in terms of a stability factor which is related to cesium seeded hydrogen plasma behavior. It is shown that application of the results already obtained with cesium seeded noble gases (argon and helium) to the case of hydrogen as the working fluid in a disk MHD generator enables a high performance power system to be defined.

  3. Catalytic-Dielectric Barrier Discharge Plasma Reactor For Methane and Carbon Dioxide Conversion

    Directory of Open Access Journals (Sweden)

    Istadi Istadi

    2007-10-01

    Full Text Available A catalytic - DBD plasma reactor was designed and developed for co-generation of synthesis gas and C2+ hydrocarbons from methane. A hybrid Artificial Neural Network - Genetic Algorithm (ANN-GA was developed to model, simulate and optimize the reactor. Effects of CH4/CO2 feed ratio, total feed flow rate, discharge voltage and reactor wall temperature on the performance of catalytic DBD plasma reactor was explored. The Pareto optimal solutions and corresponding optimal operating parameters ranges based on multi-objectives can be suggested for catalytic DBD plasma reactor owing to two cases, i.e. simultaneous maximization of CH4 conversion and C2+ selectivity, and H2 selectivity and H2/CO ratio. It can be concluded that the hybrid catalytic DBD plasma reactor is potential for co-generation of synthesis gas and higher hydrocarbons from methane and carbon dioxide and showed better than the conventional fixed bed reactor with respect to CH4 conversion, C2+ yield and H2 selectivity for CO2 OCM process. © 2007 BCREC UNDIP. All rights reserved.[Presented at Symposium and Congress of MKICS 2007, 18-19 April 2007, Semarang, Indonesia][How to Cite: I. Istadi, N.A.S. Amin. (2007. Catalytic-Dielectric Barrier Discharge Plasma Reactor For Methane and Carbon Dioxide Conversion. Bulletin of Chemical Reaction Engineering and Catalysis, 2 (2-3: 37-44.  doi:10.9767/bcrec.2.2-3.8.37-44][How to Link/DOI: http://dx.doi.org/10.9767/bcrec.2.2-3.8.37-44 || or local: http://ejournal.undip.ac.id/index.php/bcrec/article/view/8][Cited by: Scopus 1 |

  4. Power conversion systems based on Brayton cycles for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Linares, J.I., E-mail: linares@upcomillas.es [Rafael Marino Chair on New Energy Technologies. Comillas Pontifical University, Alberto Aguilera, 25-28015 Madrid (Spain); Herranz, L.E. [Unit of Nuclear Safety Research. CIEMAT, Madrid (Spain); Moratilla, B.Y.; Serrano, I.P. [Rafael Marino Chair on New Energy Technologies. Comillas Pontifical University, Alberto Aguilera, 25-28015 Madrid (Spain)

    2011-10-15

    This paper investigates Brayton power cycles for fusion reactors. Two working fluids have been explored: helium in classical configurations and CO{sub 2} in recompression layouts (Feher cycle). Typical recuperator arrangements in both cycles have been strongly constrained by low temperature of some of the energy thermal sources from the reactor. This limitation has been overcome in two ways: with a combined architecture and with dual cycles. Combined architecture couples the Brayton cycle with a Rankine one capable of taking advantage of the thermal energy content of the working fluid after exiting the turbine stage (iso-butane and steam fitted best the conditions of the He and CO{sub 2} cycles, respectively). Dual cycles set a specific Rankine cycle to exploit the lowest quality thermal energy source, allowing usual recuperator arrangements in the Brayton cycle. The results of the analyses indicate that dual cycles could reach thermal efficiencies around 42.8% when using helium, whereas thermal performance might be even better (46.7%), if a combined CO{sub 2}-H{sub 2}O cycle was set.

  5. DOE reactor-pumped laser program

    Energy Technology Data Exchange (ETDEWEB)

    Felty, J.R. [USDOE, Germantown, MD (United States). Defense Programs; Lipinski, R.J.; McArthur, D.A.; Pickard, P.S. [Sandia National Labs., Albuquerque, NM (United States)

    1993-12-31

    FALCON is a high-power, steady-state, nuclear reactor-pumped laser (RPL) concept that is being developed by the Department of Energy. The FALCON program has experimentally demonstrated reactor-pumped lasing in various mixtures of xenon, argon, neon, and helium at at wavelengths of 585, 703, 725, 1271, 1733, 1792, 2032, 2630, 2650, and 3370 nm with intrinsic efficiency as high as 2.5%. The major strengths of a reactor-pumped laser are continuous high-power operation, modular construction, self-contained power, compact size, and a variety of wavelengths (from visible to infrared). These characteristics suggest numerous applications not easily accessible to other laser types. A ground-based RPL could beam its power to space for such activities as illuminating geosynchronous communication satellites in the earth`s shadow to extend their lives, beaming power to orbital transfer vehicles, removing space debris, and providing power (from earth) to a lunar base during the long lunar night. The compact size and self-contained power also makes an RPL very suitable for ship basing so that power-beaming activities could be situated around the globe. The continuous high power of an RPL opens many potential manufacturing applications such as deep-penetration welding and cutting of thick structures, wide-area hardening of metal surfaces by heat treatment or cladding application, wide-area vapor deposition of ceramics onto metal surfaces, production of sub-micron sized particles for manufacturing of ceramics, wide-area deposition of diamond-like coatings, and 3-D ceramic lithography.

  6. DOE reactor-pumped laser program

    Science.gov (United States)

    Felty, James R.; Lipinski, Ronald J.; McArthur, David A.; Pickard, Paul S.

    1994-05-01

    FALCON is a high-power, steady-state, nuclear reactor-pumped laser (RPL) concept that is being developed by the Department of Energy. The FALCON program has experimentally demonstrated reactor-pumped lasing in various mixtures of xenon, argon, neon, and helium at wavelengths of 585, 703, 725, 1271, 1733, 1792, 2032, 2630, 2650, and 3370 nm with intrinsic efficiency as high as 2.5%. The major strengths of a reactor-pumped laser are continuous high-power operation, modular construction, self-contained power, compact size, and a variety of wavelengths (from visible to infrared). These characteristics suggest numerous applications not easily accessible to other laser types. A ground-based RPL could beam its power to space for such activities as illuminating geosynchronous communication satellites in the earth's shadow to extend their lives, beaming power to orbital transfer vehicles, removing space debris, and providing power (from earth) to a lunar base during the long lunar night. The compact size and self-contained power also makes an RPL very suitable for ship basing so that power-beaming activities could be situated around the globe. The continuous high power of an RPL opens many potential manufacturing applications such as deep-penetration welding and cutting of thick structures, wide-area hardening of metal surfaces by heat treatment or cladding application, wide-area vapor deposition of ceramics onto metal surfaces, production of sub-micron sized particles for manufacturing of ceramics, wide-area deposition of diamond- like coatings, and 3-D ceramic lithography.

  7. A novel condensation reactor for efficient CO2 to methanol conversion for storage of renewable electric energy

    NARCIS (Netherlands)

    Bos, Martin Johan; Brilman, Derk Willem Frederik

    2015-01-01

    A novel reactor design for the conversion of CO2 and H2 to methanol is developed. The conversion limitations because of thermodynamic equilibrium are bypassed via in situ condensation of a water/methanol mixture. Two temperatures zones inside the reactor ensure optimal catalyst activity (high

  8. A novel condensation reactor for efficient CO2 to methanol conversion for storage of renewable electric energy

    NARCIS (Netherlands)

    Bos, M.J.; Brilman, D.W.F.

    2015-01-01

    A novel reactor design for the conversion of CO2 and H2 to methanol is developed. The conversion limitations because of thermodynamic equilibrium are bypassed via in situ condensation of a water/methanol mixture. Two temperatures zones inside the reactor ensure optimal catalyst activity (high temper

  9. Direct conversion of methane to aromatics in a catalytic co-ionic membrane reactor.

    Science.gov (United States)

    Morejudo, S H; Zanón, R; Escolástico, S; Yuste-Tirados, I; Malerød-Fjeld, H; Vestre, P K; Coors, W G; Martínez, A; Norby, T; Serra, J M; Kjølseth, C

    2016-08-01

    Nonoxidative methane dehydroaromatization (MDA: 6CH4 ↔ C6H6 + 9H2) using shape-selective Mo/zeolite catalysts is a key technology for exploitation of stranded natural gas reserves by direct conversion into transportable liquids. However, this reaction faces two major issues: The one-pass conversion is limited by thermodynamics, and the catalyst deactivates quickly through kinetically favored formation of coke. We show that integration of an electrochemical BaZrO3-based membrane exhibiting both proton and oxide ion conductivity into an MDA reactor gives rise to high aromatic yields and improved catalyst stability. These effects originate from the simultaneous extraction of hydrogen and distributed injection of oxide ions along the reactor length. Further, we demonstrate that the electrochemical co-ionic membrane reactor enables high carbon efficiencies (up to 80%) that improve the technoeconomic process viability.

  10. Status of reduced enrichment programs for research reactors in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kanda, Keiji; Nishihara, Hedeaki [Kyoto Univ., Osaka (Japan); Shirai, Eiji; Oyamada, Rokuro; Sanokawa, Konomo [Japan Atomic Energy Research Institute, Tokyo (Japan)

    1997-08-01

    The reduced enrichment programs for the JRR-2, JRR-3, JRR-4 and JMTR of Japan Atomic Energy Research Institute (JAERI), and the KUR of Kyoto University Research Reactor Institute (KURRI) have been partially completed and are mostly still in progress under the Joint Study Programs with Argonne National Laboratory (ANL). The JMTR and JRR-2 have been already converted to use MEU aluminide fuels in 1986 and 1987, respectively. The operation of the upgraded JRR-3(JRR-3M) has started in March 1990 with the LEU aluminide fuels. Since May 1992, the two elements have been inserted in the KUR. The safety review application for the full core conversion to use LEU silicide in the JMTR was approved in February 1992 and the conversion has been done in January 1994. The Japanese Government approved a cancellation of the KUHFR Project in February 1991, and in April 1994 the U.S. Government gave an approval to utilize HEU in the KUR instead of the KUHFR. Therefore, the KUR will be operated with HEU fuel until 2001. Since March 1994, Kyoto University is continuing negotiation with UKAEA Dounreay on spent fuel reprocessing and blending down of recovered uranium, in addition to that with USDOE.

  11. Ocean Thermal Energy Conversion (OTEC) program. FY 1977 program summary

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-01-01

    An overview is given of the ongoing research, development, and demonstration efforts. Each of the DOE's Ocean Thermal Energy Conversion projects funded during fiscal year 1977 (October 1, 1976 through September 30, 1977) is described and each project's status as of December 31, 1977 is reflected. These projects are grouped as follows: program support, definition planning, engineering development, engineering test and evaluation, and advanced research and technology. (MHR)

  12. Programming Guidelines for FBD Programs in Reactor Protection System Software

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Se Jin; Lee, Dong Ah; Kim, Eui Sub; Yoo, Jun Beom [Division of Computer Science and Engineering College of Information and Communication, Konkuk University, Seoul (Korea, Republic of); Lee, Jang Su [Man-Machine Interface System team Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Properties of programming languages, such as reliability, traceability, etc., play important roles in software development to improve safety. Several researches are proposed guidelines about programming to increase the dependability of software which is developed for safety critical systems. Misra-c is a widely accepted programming guidelines for the C language especially in the sector of vehicle industry. NUREG/CR-6463 helps engineers in nuclear industry develop software in nuclear power plant systems more dependably. FBD (Function Block Diagram), which is one of programming languages defined in IEC 61131-3 standard, is often used for software development of PLC (programmable logic controllers) in nuclear power plants. Software development for critical systems using FBD needs strict guidelines, because FBD is a general language and has easily mistakable elements. There are researches about guidelines for IEC 61131-3 programming languages. They, however, do not specify details about how to use languages. This paper proposes new guidelines for the FBD based on NUREG/CR-6463. The paper introduces a CASE (Computer-Aided Software Engineering) tool to check FBD programs with the new guidelines and shows availability with a case study using a FBD program in a reactor protection system. The paper is organized as follows.

  13. Experimental study on temperature characteristics and energy conversion in packed bed reactor with dielectric barrier discharge

    Science.gov (United States)

    Li, Sen; Tang, Zuchen; Gu, Fan

    2010-10-01

    The temperature characteristics and energy conversion in packed bed reactor combined with a dielectric barrier discharge (DBD) plasma was investigated experimentally. The pellet temperatures of two types packed bed reactor, cylindrical reactor and parallel-plate reactor, was measured in conditions of various inlet voltage of DBD plasma. The relationship between pellet temperature of the packed bed and applied voltage of DBD plasma was discovered. The experimental result indicates a tendency that the pellet temperature of packed bed increases as the applied voltage of inlet plasma increases. When the voltage of inlet plasma is high enough, the pellet temperature increment decreases. Simultaneously,the packed bed temperature is sensitive to the inlet plasma energy and there is a potential application to heat exchanger. Moreover the proportion of energy consumption of plasma inputting into packed bed reactor was analyzed and calculated. The mechanisms that electrical energy of inlet plasma is transformed into heat energy in the two phases, gaseous and pellets of the packed bed reactor are different. The energy consumption in pellet phase is dielectric polarization loss and depends on packed bed geometry and DBD plasma etc. The energy consumption in gaseous phase is plasma sheath procedure. The important factors effecting on gas discharge are gaseous component and voltage, frequency of power.

  14. Role of research reactors for nuclear power program in Indonesia

    Energy Technology Data Exchange (ETDEWEB)

    Soentono, S.; Arbie, B. [National Atomic Energy Agency, Batan (Indonesia)

    1994-12-31

    The main objectives of nuclear development program in Indonesia are to master nuclear science and technology, as well as to utilise peaceful uses of nuclear know-how, aiming at stepwisely socioeconomic development. A Triga Mark II, previously of 250 kW, reactor in Bandung has been in operation since 1965 and its design power has been increased to 1000 kW in 1972. Using core grid of the Triga 250 kW, BATAN designed and constructed the Kartini Reactor in Yogyakarta which started its operation in 1979. Both of these Triga reactors have served a wide spectrum of utilisation, such as training of manpower in nuclear engineering as well as radiochemistry, isotope production and beam research activities in solid state physics. In order to support the nuclear power development program in general and to suffice the reactor experiments further, simultaneously meeting the ever increasing demand for radioisotope, the third reactor, a multipurpose reactor of 30 MW called GA. Siwabessy (RSG-GAS) has been in operation since 1987 at Serpong near Jakarta. Each of these reactors has strong cooperation with Universities, namely the Bandung Institute of Technology at Bandung, the Gadjah Mada University at Yogyakarta, and the Indonesia University at Jakarta and has facilitated the man power development required. The role of these reactors, especially the multipurpose GA. Siwabessy reactor, as essential tools in nuclear power program are described including the experience gained during preproject, construction and commissioning, as well as through their operation, maintenance and utilisation.

  15. Light Water Reactor Sustainability Program Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    George Griffith; Robert Youngblood; Jeremy Busby; Bruce Hallbert; Cathy Barnard; Kathryn McCarthy

    2012-01-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline - even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy's Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration's energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program's plans.

  16. High Temperature Fusion Reactor Cooling Using Brayton Cycle Based Partial Energy Conversion

    Science.gov (United States)

    Juhasz, Albert J.; Sawicki, Jerzy T.

    2003-01-01

    For some future space power systems using high temperature nuclear heat sources most of the output energy will be used in other than electrical form, and only a fraction of the total thermal energy generated will need to be converted to electrical work. The paper describes the conceptual design of such a partial energy conversion system, consisting of a high temperature fusion reactor operating in series with a high temperature radiator and in parallel with dual closed cycle gas turbine (CCGT) power systems, also referred to as closed Brayton cycle (CBC) systems, which are supplied with a fraction of the reactor thermal energy for conversion to electric power. Most of the fusion reactor's output is in the form of charged plasma which is expanded through a magnetic nozzle of the interplanetary propulsion system. Reactor heat energy is ducted to the high temperature series radiator utilizing the electric power generated to drive a helium gas circulation fan. In addition to discussing the thermodynamic aspects of the system design the authors include a brief overview of the gas turbine and fan rotor-dynamics and proposed bearing support technology along with performance characteristics of the three phase AC electric power generator and fan drive motor.

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

    Energy Technology Data Exchange (ETDEWEB)

    1979-12-01

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

  18. Light Water Reactor Sustainability Program Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Kathryn McCarthy; Jeremy Busby; Bruce Hallbert; Shannon Bragg-Sitton; Curtis Smith; Cathy Barnard

    2013-04-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.

  19. Light Water Reactor Sustainability Program Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    McCarthy, Kathryn A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Busby, Jeremy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hallbert, Bruce [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bragg-Sitton, Shannon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Barnard, Cathy [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-04-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.

  20. Conversion enhancement of tubular fixed-bed reactor for Fischer-Tropsch synthesis using static mixer

    Institute of Scientific and Technical Information of China (English)

    Phavanee Narataruksa; Sabaithip Tungkamani; Karn Pana-Suppamassadu; Phongsak Keeratiwintakorn; Siriluck Nivitchanyong; Piyapong Hunpinyo; Hussanai Sukkathanyawat; Prayut Jiamrittiwong; Visarut Nopparat

    2012-01-01

    Recently,Fischer-Tropsch synthesis (FTS) has become an interesting technology because of its potential role in producing biofuels via Biomassto-Liquids (BTL) processes.In Fischer-Tropsch (FT) section,biomass-derived syngas,mainly composed of a mixture of carbon monoxide (CO)and hydrogen (H2),is converted into various forms of hydrocarbon products over a catalyst at specified temperature and pressure.Fixed-bed reactors are typically used for these processes as conventional FT reactors.The fixed-bed or packed-bed type reactor has its drawbacks,which are heat transfer limitation,i.e.a hot spot problem involved highly exothermic characteristics of FT reaction,and mass transfer limitation due to the condensation of liquid hydrocarbon products occurred on catalyst surface.This work is initiated to develop a new chemical reactor design in which a better distribution of gaseous reactants and hydrocarbon products could be achieved,and led to higher throughput and conversion.The main goal of the research is the enhancement of a fixed-bed reactor,focusing on the application of KenicsTM static mixer insertion in the tubular packed-bed reactor.Two FTS experiments were carried out using two reactors i.e.,with and without static mixer insertion within catalytic beds.The modeled syngas used was a mixed gas composed of H2/CO in 2 ∶ 1 molar ratio that was fed at the rate of 30 mL(STP)·min-1 (GHSV ≈ 136 mL·g-1cat·h-1) into the fixed Ru supported aluminum catalyst bed of weight 13.3 g.The reaction was carried out at 180 ℃ and atmospheric pressure continuously for 36 h for both experiments.Both transient and steady-state conversions (in terms of time on stream) were reported.The results revealed that the steady-state CO conversion for the case using the static mixer was approximately 3.5 times higher than that of the case without static mixer.In both cases,the values of chain growth probability of hydrocarbon products (α) for Fischer-Tropsch synthesis were 0.92 and 0.89 for

  1. Distributed secondary gas injection via a fractal injector: a nature-inspired approach to improving conversion in fluidized bed reactors

    NARCIS (Netherlands)

    Christensen, D.O.

    2008-01-01

    The conversion in bubbling fluidized bed reactors is suppressed because the interphase mass transfer and gas-solid contact in bubbling fluidized bed reactors are often poor. Most of the gas is present in the form of bubbles, which have low surface-to-volume ratios and are nearly devoid of catalyst

  2. Effective conversion of biomass tar into fuel gases in a microwave reactor

    Science.gov (United States)

    Anis, Samsudin; Zainal, Z. A.

    2016-06-01

    This work deals with conversion of naphthalene (C10H8) as a biomass tar model compound by means of thermal and catalytic treatments. A modified microwave oven with a maximum output power of 700 W was used as the experimental reactor. Experiments were performed in a wide temperature range of 450-1200°C at a predetermined residence time of 0.24-0.5 s. Dolomite and Y-zeolite were applied to convert naphthalene catalytically into useful gases. Experimental results on naphthalene conversion showed that conversion efficiency and yield of gases increased significantly with the increase of temperature. More than 90% naphthalene conversion efficiency was achieved by thermal treatment at 1200°C and 0.5 s. Nevertheless, this treatment was unfavorable for fuel gases production. The main product of this treatment was soot. Catalytic treatment provided different results with that of thermal treatment in which fuel gases formation was found to be the important product of naphthalene conversion. At a high temperature of 900°C, dolomite had better conversion activity where almost 40 wt.% of naphthalene could be converted into hydrogen, methane and other hydrocarbon gases.

  3. Effective conversion of biomass tar into fuel gases in a microwave reactor

    Energy Technology Data Exchange (ETDEWEB)

    Anis, Samsudin, E-mail: samsudin-anis@yahoo.com [Department of Mechanical Engineering, Universitas Negeri Semarang, Kampus Sekaran, Gunungpati, 50229 Semarang, 8508101 (Indonesia); Zainal, Z. A., E-mail: mezainal@usm.my [School of Mechanical Engineering, Universiti Sains Malaysia, Engineering Campus, 14300 Nibong Tebal, Penang (Malaysia)

    2016-06-03

    This work deals with conversion of naphthalene (C{sub 10}H{sub 8}) as a biomass tar model compound by means of thermal and catalytic treatments. A modified microwave oven with a maximum output power of 700 W was used as the experimental reactor. Experiments were performed in a wide temperature range of 450-1200°C at a predetermined residence time of 0.24-0.5 s. Dolomite and Y-zeolite were applied to convert naphthalene catalytically into useful gases. Experimental results on naphthalene conversion showed that conversion efficiency and yield of gases increased significantly with the increase of temperature. More than 90% naphthalene conversion efficiency was achieved by thermal treatment at 1200°C and 0.5 s. Nevertheless, this treatment was unfavorable for fuel gases production. The main product of this treatment was soot. Catalytic treatment provided different results with that of thermal treatment in which fuel gases formation was found to be the important product of naphthalene conversion. At a high temperature of 900°C, dolomite had better conversion activity where almost 40 wt.% of naphthalene could be converted into hydrogen, methane and other hydrocarbon gases.

  4. Light Water Reactor Sustainability Program: Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2017-05-01

    proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans. For the LWRS Program, sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for a longer-than-initially-licensed lifetime. It has two facets with respect to long-term operations: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the industry to implement technology to exceed the performance of the current labor-intensive business model.

  5. Advanced gas cooled nuclear reactor materials evaluation and development program

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    Results of work performed from January 1, 1977 through March 31, 1977 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed.

  6. Reactors

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1988-01-01

    This standard applies to the following types of reactors: shunt reactors, current-limiting reactors including neutral-earthing reactors, damping reactors, tuning (filter) reactors, earthing transformers (neutral couplers), arc-suppression reactors, smoothing reactors, with the exception of the following reactors: small reactors with a rating generally less than 2 kvar single-phase and 10 kvar three-phase, reactors for special purposes such as high-frequency line traps or reactors mounted on rolling stock.

  7. Balanced program plan. Volume IV. Coal conversion

    Energy Technology Data Exchange (ETDEWEB)

    Richmond, C. R.; Reichle, D. E.; Gehrs, C. W.

    1976-05-01

    This document contains a description of the biomedical and environmental research necessary to ensure the timely attainment of coal conversion technologies amenable to man and his environment. The document is divided into three sections. The first deals with the types of processes currently being considered for development; the data currently available on composition of product, process and product streams, and their potential effects; and problems that might arise from transportation and use of products. Section II is concerned with a description of the necessary research in each of the King-Muir categories, while the third section presents the research strategies necessary to assess the potential problems at the conversion plant (site specific) and those problems that might effect the general public and environment as a result of the operation of large-scale coal conversion plants. (auth)

  8. Documented Safety Analysis Addendum for the Neutron Radiography Reactor Facility Core Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Boyd D. Christensen

    2009-05-01

    The Neutron Radiography Reactor Facility (NRAD) is a Training, Research, Isotope Production, General Atomics (TRIGA) reactor which was installed in the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) in the mid 1970s. The facility provides researchers the capability to examine both irradiated and non-irradiated materials in support of reactor fuel and components programs through non-destructive neutron radiography examination. The facility has been used in the past as one facet of a suite of reactor fuels and component examination facilities available to researchers at the INL and throughout the DOE complex. The facility has also served various commercial research activities in addition to the DOE research and development support. The reactor was initially constructed using Fuel Lifetime Improvement Program (FLIP)- type highly enriched uranium (HEU) fuel obtained from the dismantled Puerto Rico Nuclear Center (PRNC) reactor. In accordance with international non-proliferation agreements, the NRAD core will be converted to a low enriched uranium (LEU) fuel and will continue to utilize the PRNC control rods, control rod drives, startup source, and instrument console as was previously used with the HEU core. The existing NRAD Safety Analysis Report (SAR) was created and maintained in the preferred format of the day, combining sections of both DOE-STD-3009 and Nuclear Regulatory Commission Regulatory Guide 1.70. An addendum was developed to cover the refueling and reactor operation with the LEU core. This addendum follows the existing SAR format combining required formats from both the DOE and NRC. This paper discusses the project to successfully write a compliant and approved addendum to the existing safety basis documents.

  9. Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

    2010-01-01

    A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

  10. Low-temperature conversion of high-moisture biomass: Continuous reactor system results

    Energy Technology Data Exchange (ETDEWEB)

    Elliott, D.C.; Sealock, L.J. Jr.; Butner, R.S.; Baker, E.G.; Neuenschwander, G.G.

    1989-10-01

    Pacific Northwest Laboratory (PNL) is developing a low-temperature, catalytic process for converting high-moisture biomass feedstocks and other wet organic substances to useful gaseous fuels. This system, in which thermocatalytic conversion takes place in an aqueous environment, was designed to overcome the problems usually encountered with high-water-content feedstocks. The process uses a reduced nickel catalyst at temperatures as low as 350{degree}C and pressures ranging from 2000 to 4000 psig -- conditions favoring the formation of gas consisting mostly of methane. The results of numerous batch tests showed that the system could convert feedstocks not readily converted by conventional methods. Fifteen tests were conducted in a continuous reactor system to further evaluate the effectiveness of the process for high-moisture biomass gasification and to obtain conversion rate data needed for process scaleup. During the tests, the complex gasification reactions were evaluated by several analytical methods. The results of these tests show that the heating value of the gas ranged from 400 to 500 Btu/scf, and if the carbon dioxide is removed, the product gas is pipeline quality. Conversion of the feedstocks was high. Engineering analysis indicates that, based on these results, a tubular reactor can be designed that should convert greater than 99% of the carbon fed as high-moisture biomass to a gaseous product in a reaction time of less than 11 min.

  11. NASA-OAST program in photovoltaic energy conversion

    Science.gov (United States)

    Mullin, J. P.; Flood, D. J.

    1982-01-01

    The NASA program in photovoltaic energy conversion includes research and technology development efforts on solar cells, blankets, and arrays. The overall objectives are to increase conversion efficiency, reduce mass, reduce cost, and increase operating life. The potential growth of space power requirements in the future presents a major challenge to the current state of technology in space photovoltaic systems.

  12. High Temperature and Pressure Alkaline Electrochemical Reactor for Conversion of Power to Chemicals

    DEFF Research Database (Denmark)

    Chatzichristodoulou, Christodoulos

    2016-01-01

    Moving away from fossil fuels requires harvesting more and more intermittent renewable energy resources and establishing a sustainable system for the production of chemicals. This brings forward the need for efficient large scale energy storage technologies 1-3 and technologies for the conversion...... densities. This work will provide an overview of our efforts to develop components of such high temperature alkaline electrochemical reactors for different applications. Low-cost large-scale production methods have been successfully employed for the production of ceramic diaphragms and full cells...... of renewable electricity to chemicals. Electrochemical reactors can play a crucial role in this endeavor, since they can efficiently and reversibly transform electricity to high-value chemicals, and thus serve as energy storage and recovery devices for balancing the grid, while offering a means...

  13. A green approach to ethyl acetate: Quantitative conversion of ethanol through direct dehydrogenation in a Pd-Ag membrane reactor

    KAUST Repository

    Zeng, Gaofeng

    2012-11-07

    Pincers do the trick: The conversion of ethanol to ethyl acetate and hydrogen was achieved using a pincer-Ru catalyst in a Pd-Ag membrane reactor. Near quantitative conversions and yields could be achieved without the need for acid or base promoters or hydrogen acceptors (see scheme). © 2012 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  14. NCSU Reactor Sharing Program. Final technical report, [September 1, 1980--August 29, 1992

    Energy Technology Data Exchange (ETDEWEB)

    Perez, P.B.

    1993-11-10

    The Nuclear Reactor Program at North Carolina State University provides the PULSTAR Research Reactor and associated facilities to eligible institutions with support, in part, from the Department of Energy Reactor Sharing Program. Participation in the NCSU Reactor Sharing Program continues to increase steadily with visitors ranging from advance high school physics and chemistry students to Ph.D. level research from neighboring universities.

  15. High Temperature and Pressure Alkaline Electrochemical Reactor for Conversion of Power to Chemicals

    DEFF Research Database (Denmark)

    Chatzichristodoulou, Christodoulos

    2016-01-01

    Moving away from fossil fuels requires harvesting more and more intermittent renewable energy resources and establishing a sustainable system for the production of chemicals. This brings forward the need for efficient large scale energy storage technologies 1-3 and technologies for the conversion...... of renewable electricity to chemicals. Electrochemical reactors can play a crucial role in this endeavor, since they can efficiently and reversibly transform electricity to high-value chemicals, and thus serve as energy storage and recovery devices for balancing the grid, while offering a means...

  16. The Socioeconomic Assessment of Sloping Land Conversion Program in China

    DEFF Research Database (Denmark)

    Liu, Zhen

    , reduction of compensation payment rate lowers the treatment effect of the SLCP on participating households from both River basins. Lastly, the third paper, ― The effect of the Sloping Land Conversion Program on farm household productivity in rural China, investigates the treatment effect (the causal effect......Abstract This thesis mainly focuses on the socioeconomic impact of the largest Ecological Recovery Program ― the Sloping Land Conversion Program (SLCP), also called Grain for Green Program (GFG) in China. The central government initiated this program in 1999 and it was launched nationwide in 2002...... of household survey data, this study aims to improve our understanding of the treatment effect of the SLCP on farm households, which is split into three parts. The first paper ― The Sloping Land Conversion Program in China: Effects on Rural Households’ Livelihood Diversification, evaluates the effects...

  17. Feasibility analyses for HEU to LEU fuel conversion of the LAUE Langivin Institute (ILL) High Flux Reactor (RHF).

    Energy Technology Data Exchange (ETDEWEB)

    Stevens, J.; Tentner. A.; Bergeron, A.; Nuclear Engineering Division

    2010-08-19

    The High Flux Reactor (RHF) of the Laue Langevin Institute (ILL) based in Grenoble, France is a research reactor designed primarily for neutron beam experiments for fundamental science. It delivers one of the most intense neutron fluxes worldwide, with an unperturbed thermal neutron flux of 1.5 x 10{sup 15} n/cm{sup 2}/s in its reflector. The reactor has been conceived to operate at a nuclear power of 57 MW but currently operates at 52 MW. The reactor currently uses a Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most worldwide research and test reactors have already started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on a mixture of uranium and molybdenum (UMo) is expected to allow the conversion of compact high performance reactors like the RHF. This report presents the results of reactor design, performance and steady state safety analyses for conversion of the RHF from the use of HEU fuel to the use of UMo LEU fuel. The objective of this work was to show that is feasible, under a set of manufacturing assumptions, to design a new RHF fuel element that could safely replace the HEU element currently used. The new proposed design has been developed to maximize performance, minimize changes and preserve strong safety margins. Neutronics and thermal-hydraulics models of the RHF have been developed and qualified by benchmark against experiments and/or against other codes and models. The models developed were then used to evaluate the RHF performance if LEU UMo were to replace the current HEU fuel 'meat' without any geometric change to the fuel plates. Results of these direct replacement analyses have shown a significant degradation of the RHF performance, in terms of both neutron flux and cycle

  18. Conversion of L-lactate into n-caproate by a continuously fed reactor microbiome.

    Science.gov (United States)

    Kucek, Leo A; Nguyen, Mytien; Angenent, Largus T

    2016-04-15

    Conversion of lactate to n-caproate had been described for the type strain Megasphaera elsdenii in batch systems. Recently, investigators have also described production of n-caproate from endogenous or exogenous lactate with batch-fed reactor microbiome systems. However, no reports exist of lactate to n-caproate conversion within a continuously fed bioreactor. Since continuously fed systems are advantageous for biotechnology production platforms, our objective was to develop such a system. Here, we demonstrated continuous lactate to n-caproate conversion for more than 165 days. The volumetric n-caproate production rate (productivity) was improved when we decreased the operating pH from 5.5 to 5.0, and was again improved when we utilized in-line product recovery via pertraction (membrane-based liquid-liquid extraction). We observed a maximum n-caproate productivity of 6.9 g COD/L-d for a period of 17 days at an L-lactate loading rate of 9.1 g COD/L-d, representing the highest sustained lactate to n-caproate conversion rate ever reported. We had to manage two competing lactate conversion pathways: 1) the reverse β-oxidation pathway to n-caproate; and 2) the acrylate pathway to propionate. We found that maintaining a low residual lactate concentration in the bioreactor broth was necessary to direct lactate conversion towards n-caproate instead of propionate. These findings provide a foundation for the development of new resource recovery processes to produce higher-value liquid products (e.g., n-caproate) from carbon-rich wastewaters containing lactate or lactate precursors (e.g., carbohydrates).

  19. EBR-2 (Experimental Breeder Reactor-2), IFR (Integral Fast Reactor) prototype testing programs

    Energy Technology Data Exchange (ETDEWEB)

    Lehto, W.K.; Sackett, J.I.; Lindsay, R.W. (Argonne National Lab., Idaho Falls, ID (USA). EBR-II Div. Argonne National Lab., IL (USA)); Planchon, H.P.; Lambert, J.D.B. (Argonne National Lab., IL (USA))

    1990-01-01

    The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development. (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs.

  20. Solar spectral conversion for improving the photosynthetic activity in algae reactors

    Science.gov (United States)

    Wondraczek, Lothar; Batentschuk, Miroslaw; Schmidt, Markus A.; Borchardt, Rudolf; Scheiner, Simon; Seemann, Benjamin; Schweizer, Peter; Brabec, Christoph J.

    2013-06-01

    Sustainable biomass production is expected to be one of the major supporting pillars for future energy supply, as well as for renewable material provision. Algal beds represent an exciting resource for biomass/biofuel, fine chemicals and CO2 storage. Similar to other solar energy harvesting techniques, the efficiency of algal photosynthesis depends on the spectral overlap between solar irradiation and chloroplast absorption. Here we demonstrate that spectral conversion can be employed to significantly improve biomass growth and oxygen production rate in closed-cycle algae reactors. For this purpose, we adapt a photoluminescent phosphor of the type Ca0.59Sr0.40Eu0.01S, which enables efficient conversion of the green part of the incoming spectrum into red light to better match the Qy peak of chlorophyll b. Integration of a Ca0.59Sr0.40Eu0.01S backlight converter into a flat panel algae reactor filled with Haematococcus pluvialis as a model species results in significantly increased photosynthetic activity and algae reproduction rate.

  1. Main activities carried out for the conversion of the reactor core TRIGA, from HEU 8.5/70 / LEU 8.5/20 to LEU 30/20; Principales actividades llevadas a cabo para la conversion del nucleo del reactor TRIGA, de HEU 8.5/70 / LEU 8.5/20 a LEU 30/20

    Energy Technology Data Exchange (ETDEWEB)

    Flores C, J., E-mail: jorge.floresc@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    In agreement with the policies of the global initiative of threats reduction (GTRI), Mexico committed that inside the reduction program of the fuel enrichment in research and test reactors (RERTR), the conversion of the core reactor TRIGA (in the nuclear centre) would be made, to use solely fuel with low enrichment ({<=} 20% U{sup 235}). To support to the execution of this commitment, a series of accords and agreements were established. The Project Agreement and Supply among the IAEA, the United States of America and Mexico was the more relevant. In this work the main activities carried out in the Instituto Nacional de Investigaciones Nucleares (ININ) with this purpose are presented. (Author)

  2. TRIGA Mark-III reactor dismantling program

    Energy Technology Data Exchange (ETDEWEB)

    Chung, U. S.; Lee, B. J.; Paik, S. T.; Jung, K. J. [TRIGA Research Reactor D and D Project Team, Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-07-01

    The activation assessment of the main parts of the TRIGA Mark-III (KRR-2) was estimated to effectively dismantle the activated and contaminated areas. All of the method and the order for decommissioning the KRR-2 have been chosen as a result of the examination of the physical structure and radiological conditions of the reactor component. These decommissioning methods and orders were reviewed as part of the Hazard and Operability (HAZOP) studies for the project. Radiological assessment is also done to protect the workers and the environment from the dismantling work. License documents were submitted to the Ministry of Science and Technology (MOST) at the end of 1998. Practical work of the D and D will start at the end of 1999 once the government issues the license. Radiation protection plan was also set up to control the workers and environment. This paper summarized the main lines of those studies. (author)

  3. Numerical simulation of ion transport membrane reactors: Oxygen permeation and transport and fuel conversion

    KAUST Repository

    Hong, Jongsup

    2012-07-01

    Ion transport membrane (ITM) based reactors have been suggested as a novel technology for several applications including fuel reforming and oxy-fuel combustion, which integrates air separation and fuel conversion while reducing complexity and the associated energy penalty. To utilize this technology more effectively, it is necessary to develop a better understanding of the fundamental processes of oxygen transport and fuel conversion in the immediate vicinity of the membrane. In this paper, a numerical model that spatially resolves the gas flow, transport and reactions is presented. The model incorporates detailed gas phase chemistry and transport. The model is used to express the oxygen permeation flux in terms of the oxygen concentrations at the membrane surface given data on the bulk concentration, which is necessary for cases when mass transfer limitations on the permeate side are important and for reactive flow modeling. The simulation results show the dependence of oxygen transport and fuel conversion on the geometry and flow parameters including the membrane temperature, feed and sweep gas flow, oxygen concentration in the feed and fuel concentration in the sweep gas. © 2012 Elsevier B.V.

  4. Catalytic and Noncatalytic Conversion of Methane to Olefins and Synthesis Gas in an AC Parallel Plate Discharge Reactor

    Directory of Open Access Journals (Sweden)

    Mohammad Ali Khodagholi

    2013-01-01

    Full Text Available Direct conversion of methane to ethylene, acetylene, and synthesis gas at ambient pressure and temperature in a parallel plate discharge reactor was investigated. The experiments were carried out using a quartz reactor of outer diameter of 9 millimeter and a driving force of ac current of 50 Hz. The input power to the reactor to establish a stable gas discharge varied from 9.6 to maximum 15.3 watts (w. The effects of ZSM5, Fe–ZSM5, and Ni–ZSM5 catalysts combined with corona discharge for conversion of methane to more valued products have been addressed. It was found that in presence or absence of a catalyst in gas discharge reactor, the rate of methane and oxygen conversion increased upon higher input power supplied to the reactor. The effect of Fe–ZSM5 catalyst combined with gas discharge plasma yields C2 hydrocarbons up to 21.9%, which is the highest productions of C2 hydrocarbons in this work. The effect of combined Ni–ZSM5 and gas discharge plasma was mainly production of synthesis gas. The advantage of introducing ZSM5 to the plasma zone was increase in synthesis gas and acetylene production. The highest energy efficiency was 0.22 mmol/kJ, which belongs to lower rate of energy injection to the reactor.

  5. Integral Fast Reactor Program annual progress report, FY 1991

    Energy Technology Data Exchange (ETDEWEB)

    1992-06-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1991. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R&D.

  6. Integral Fast Reactor Program annual progress report, FY 1991

    Energy Technology Data Exchange (ETDEWEB)

    1992-06-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1991. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R D.

  7. Integral Fast Reactor Program annual progress report, FY 1994

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, J.J.

    1994-12-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1994. Technical accomplishments are presented in the following areas of the IFR technology development activities: metal fuel performance; pyroprocess development; safety experiments and analyses; core design development; fuel cycle demonstration; and LMR technology R&D.

  8. Integral Fast Reactor Program. Annual progress report, FY 1992

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, M.J.

    1993-06-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1992. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R&D.

  9. Integral Fast Reactor Program. Annual progress report, FY 1993

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, M.J.

    1994-10-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1993. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R and D.

  10. Neutronic analysis for core conversion (HEU–LEU) of the low power research reactor using the MCNP4C code

    OpenAIRE

    Aldawahra Saadou; Khattab Kassem; Saba Gorge

    2015-01-01

    Comparative studies for conversion of the fuel from HEU to LEU in the miniature neutron source reactor (MNSR) have been performed using the MCNP4C code. The HEU fuel (UAl4-Al, 90% enriched with Al clad) and LEU (UO2 12.6% enriched with zircaloy-4 alloy clad) cores have been analyzed in this study. The existing HEU core of MNSR was analyzed to validate the neutronic model of reactor, while the LEU core was studied to prove the possibility of fuel conversion of the existing HEU core. The propos...

  11. The coupling effect of gas-phase chemistry and surface reactions on oxygen permeation and fuel conversion in ITM reactors

    KAUST Repository

    Hong, Jongsup

    2015-08-01

    © 2015 Elsevier B.V. The effect of the coupling between heterogeneous catalytic reactions supported by an ion transport membrane (ITM) and gas-phase chemistry on fuel conversion and oxygen permeation in ITM reactors is examined. In ITM reactors, thermochemical reactions take place in the gas-phase and on the membrane surface, both of which interact with oxygen permeation. However, this coupling between gas-phase and surface chemistry has not been examined in detail. In this study, a parametric analysis using numerical simulations is conducted to investigate this coupling and its impact on fuel conversion and oxygen permeation rates. A thermochemical model that incorporates heterogeneous chemistry on the membrane surface and detailed chemical kinetics in the gas-phase is used. Results show that fuel conversion and oxygen permeation are strongly influenced by the simultaneous action of both chemistries. It is shown that the coupling somewhat suppresses the gas-phase kinetics and reduces fuel conversion, both attributed to extensive thermal energy transfer towards the membrane which conducts it to the air side and radiates to the reactor walls. The reaction pathway and products, in the form of syngas and C2 hydrocarbons, are also affected. In addition, the operating regimes of ITM reactors in which heterogeneous- or/and homogeneous-phase reactions predominantly contribute to fuel conversion and oxygen permeation are elucidated.

  12. Neutron flux reduction programs for reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, C.S. [Korea Atomic Energy Research Inst. KAERI, 150 Deogjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Kim, B.C. [Korea Reactor Integrity Surveillance Technology KRIST, 150 Deogjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2011-07-01

    The objective of this work is to implement various fast neutron flux reduction programs on the belt-line region of the reactor pressure vessel to reduce the increasing rate of reference temperature for pressurized thermal shock (RT PTS) for Korea Nuclear Unit 1. A pressurized thermal shock (PTS) event is an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. A PTS concern arises if one of these transients acts in the belt-line region of a reactor vessel where a reduced fracture resistance exists because of neutron irradiation. Generally, the RT PTS value is continuously increasing according to the fast neutron irradiation during the reactor operation, and it can reach the screening criterion prior to the expiration of the operating license. To reduce the increasing rate of RT PTS, various neutron flux reduction programs can be implemented, which are focused on license renewal. In this paper, neutron flux reduction programs, such as low leakage loading pattern strategy, loading of neutron absorber rods, and dummy fuel assembly loading are considered for Korea Nuclear Unit 1, of which the RT PTS value of the leading material (circumferential weld) is going to reach the screening criterion in the near future. To evaluate the effects of the neutron flux reduction programs, plant and cycle specific forward neutron transport calculations for the various neutron flux reduction programs were carried out. For the analysis, all transport calculations were carried out by using the DORT 3.1 discrete ordinate code and BUGLE-96 cross-section library. (authors)

  13. Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Katoh, Yutai [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Wilson, Dane F [ORNL

    2008-08-01

    Since 2002, the Department of Energy's (DOE's) Generation IV Nuclear Energy Systems (Gen IV) Program has addressed the research and development (R&D) necessary to support next-generation nuclear energy systems. The six most promising systems identified for next-generation nuclear energy are described within this roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor-SCWR and the Very High Temperature Reactor-VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor-GFR, the Lead-cooled Fast Reactor-LFR, and the Sodium-cooled Fast Reactor-SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides and may provide an alternative to accelerator-driven systems. At the inception of DOE's Gen IV program, it was decided to significantly pursue five of the six concepts identified in the Gen IV roadmap to determine which of them was most appropriate to meet the needs of future U.S. nuclear power generation. In particular, evaluation of the highly efficient thermal SCWR and VHTR reactors was initiated primarily for energy production, and evaluation of the three fast reactor concepts, SFR, LFR, and GFR, was begun to assess viability for both energy production and their potential contribution to closing the fuel cycle. Within the Gen IV Program itself, only the VHTR class of reactors was selected for continued development. Hence, this document will address the multiple activities under the Gen IV program that contribute to the development of the VHTR. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of

  14. Overview of the Lockheed Martin Compact Fusion Reactor (CFR) Program

    Science.gov (United States)

    McGuire, Thomas

    2015-11-01

    The Lockheed Martin Compact Fusion Reactor (CFR) Program endeavors to quickly develop a compact fusion power plant with favorable commercial economics and military utility. An overview of the concept and its diamagnetic, high beta magnetically encapsulated linear ring cusp confinement scheme will be given. The analytical model of the major loss mechanisms and predicted performance will be discussed, along with the major physics challenges. Key features of an operational CFR reactor will be highlighted. The proposed developmental path following the current experimental efforts will be presented. ©2015 Lockheed Martin Corporation. All Rights Reserved.

  15. Test Results From a Direct Drive Gas Reactor Simulator Coupled to a Brayton Power Conversion Unit

    Science.gov (United States)

    Hervol, David S.; Briggs, Maxwell H.; Owen, Albert K.; Bragg-Sitton, Shannon M.

    2009-01-01

    The Brayton Power Conversion Unit (BPCU) located at NASA Glenn Research Center (GRC) in Cleveland, OH is a closed cycle system incorporating a turboaltemator, recuperator, and gas cooler connected by gas ducts to an external gas heater. For this series of tests, the BPCU was modified by replacing the gas heater with the Direct Drive Gas heater or DOG. The DOG uses electric resistance heaters to simulate a fast spectrum nuclear reactor similar to those proposed for space power applications. The combined system thermal transient behavior was the focus of these tests. The BPCU was operated at various steady state points. At each point it was subjected to transient changes involving shaft rotational speed or DOG electrical input. This paper outlines the changes made to the test unit and describes the testing that took place along with the test results.

  16. A neutronic feasibility study for LEU conversion of the WWR-SM research reactor in Uzbekistan.

    Energy Technology Data Exchange (ETDEWEB)

    Rakhmanov, A.

    1998-10-19

    The WWR-SM research reactor in Uzbekistan has operated at 10 MW since 1979, using Russian-supplied IRT-3M fuel assemblies containing 90% enriched uranium. Burnup tests of three full-sized IRT-3M FA with 36% enrichment were successfully completed to a burn up of about {approximately}50% in 1987-1989. In August 1998, four IRT-3M FA with 36% enriched uranium were loaded into the core to initiate conversion of the entire core to 36% enriched fuel. This paper presents the results of equilibrium fuel cycle comparisons of the reactor using HEU (90%) and HEU (36%) IRT-3M fuel and compares results with the performance of IRT-4M FA containing LEU (19.75%). The results show that an LEU (19.75%) density of 3.8 g/cm{sup 3} is required to match the cycle length of the HEU (90%) core and an LEU density 3.9 g/cm{sup 3} is needed to match the cycle length of the HEU (36%) core.

  17. Sandwich-structured enzyme membrane reactor for efficient conversion of maltose into isomaltooligosaccharides.

    Science.gov (United States)

    Zhang, Lei; Su, Yanlei; Zheng, Yang; Jiang, Zhongyi; Shi, Jiafu; Zhu, Yuanyuan; Jiang, Yanjun

    2010-12-01

    A novel enzyme membrane reactor with sandwich structure has been developed by confining glucosidase between two sheets of ultrafiltration membranes to effectively convert maltose to isomaltooligosaccharides (IMOs). The hydrophilic ultrafiltration membranes, which were prepared by phase inversion method using PES as bulk polymer and Pluronic F127 as both surface modification and pore formation agent, exhibited the desirable enzyme adsorption-resistant property. The scanning electron microscopy (SEM) photographs showed that two sheets of PES/Pluronic F127 membranes were packed tightly and glucosidase was kept in a free state within a nanoscale space. When the weight ratio of Pluronic F127 to PES was 30%, glucosidase could be completely rejected by the membranes. Due to the sandwich structuring of the membrane reactor and the high hydrophilicity of the PES/Pluronic F127 membrane surface, maltose conversion and yield reached 100% and 58% under the optimum experimental conditions (pH 6.0, 50 degrees C), respectively. 2010 Elsevier Ltd. All rights reserved.

  18. Gas-cooled reactor programs. High-temperature gas-cooled reactor technology development program. Annual progress report, December 31, 1983

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.; Sanders, J.P.

    1984-06-01

    ORNL continues to make significant contributions to the national program. In the HTR fuels area, we are providing detailed statistical information on the fission product retention performance of irradiated fuel. Our studies are also providing basic data on the mechanical, physical, and chemical behavior of HTR materials, including metals, ceramics, graphite, and concrete. The ORNL has an important role in the development of improved HTR graphites and in the specification of criteria that need to be met by commercial products. We are also developing improved reactor physics design methods. Our work in component development and testing centers in the Component Flow Test Loop (CFTL), which is being used to evaluate the performance of the HTR core support structure. Other work includes experimental evaluation of the shielding effectiveness of the lower portions of an HTR core. This evaluation is being performed at the ORNL Tower Shielding Facility. Researchers at ORNL are developing welding techniques for attaching steam generator tubing to the tubesheets and are testing ceramic pads on which the core posts rest. They are also performing extensive testing of aggregate materials obtained from potential HTR site areas for possible use in prestressed concrete reactor vessels. During the past year we continued to serve as a peer reviewer of small modular reactor designs being developed by GA and GE with balance-of-plant layouts being developed by Bechtel Group, Inc. We have also evaluated the national need for developing HTRs with emphasis on the longer term applications of the HTRs to fossil conversion processes.

  19. Energy Conversion and Storage Program: 1992 Annual report

    Energy Technology Data Exchange (ETDEWEB)

    Cairns, E.J.

    1993-06-01

    This report is the 1992 annual progress report for the Energy Conversion and Storage Program, a part of the Energy and Environment Division of the Lawrence Berkeley Laboratory. Work described falls into three broad areas: electrochemistry; chemical applications; and materials applications. The Energy Conversion and Storage Program applies principles of chemistry and materials science to solve problems in several areas: (1) production of new synthetic fuels, (2) development of high-performance rechargeable batteries and fuel cells, (3) development of advanced thermochemical processes for energy conversion, (4) characterization of complex chemical processes and chemical species, and (5) study and application of novel materials for energy conversion and transmission. Projects focus on transport-process principles, chemical kinetics, thermodynamics, separation processes, organic and physical chemistry, novel materials, and advanced methods of analysis. Electrochemistry research aims to develop advanced power systems for electric vehicle and stationary energy storage applications. Chemical applications research includes topics such as separations, catalysis, fuels, and chemical analyses. Included in this program area are projects to develop improved, energy-efficient methods for processing product and waste streams from synfuel plants, coal gasifiers, and biomass conversion processes. Materials applications research includes evaluation of the properties of advanced materials, as well as development of novel preparation techniques. For example, techniques such as sputtering, laser ablation, and poised laser deposition are being used to produce high-temperature superconducting films.

  20. The Socioeconomic Assessment of Sloping Land Conversion Program in China

    DEFF Research Database (Denmark)

    Liu, Zhen

    Abstract This thesis mainly focuses on the socioeconomic impact of the largest Ecological Recovery Program ― the Sloping Land Conversion Program (SLCP), also called Grain for Green Program (GFG) in China. The central government initiated this program in 1999 and it was launched nationwide in 2002...... of which support the policy intention of central government according to our own understanding, whereas the effects differ depending on the group, region and period. This research provides a detailed understanding of the treatment effects of the SLCP and thus, contributes to the on-going political debate...

  1. Programming-Free Form Conversion, Design, and Processing

    OpenAIRE

    Fan, Ting-Jun; Machlin, Rona S.; Wang, Christopher P.; Chang, Ifay F.

    1990-01-01

    In this paper, we present the requirements and design considerations for programming-free form conversion, design, and processing. A set of object-oriented software tools are also presented to help users convert a paper form into an electronic form, design an electronic form, and fill in an electronic form directly on screen.

  2. 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

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Sean R. Morrell

    2012-09-01

    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

  3. Biomass Program 2007 Program Peer Review - Thermochemical Conversion Platform Summary

    Energy Technology Data Exchange (ETDEWEB)

    none,

    2009-10-27

    This document discloses the comments provided by a review panel at the U.S. Department of Energy Office of the Biomass Program Peer Review held on November 15-16, 2007 in Baltimore, MD and the Biomass Program Peer Review for the Thermochemical Platform, held on July 9th and 10th in Golden, Colorado.

  4. Modification of reference temperature program in reactor regulating system

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  5. Compact conversion of natural gas and biomass to DME in microstructured reactors

    Energy Technology Data Exchange (ETDEWEB)

    Myrstad, Rune

    2010-07-01

    Efficient production of easily distributable fuel from natural gas or biomass in the small-to-medium scale calls for a more compact and efficient process than using conventional technology. Microstructured reactors have improved heat and mass transfer properties which make them suitable for process intensified production of liquid fuel from synthesis gas and demonstration plants using such technology are announced. Dimethyl ether (DME) can be used as an intermediate in the production of several industrial chemicals and DME is also used as an aerosol propellant because of its environmentally benign properties. Since DME has high cetane number and is considered as an ultra-clean fuel with reduced NO{sub x}, SO{sub x}, and PM emissions, DME has emerged as a substitute for auto diesel fuel and bio-DME is one of the most promising second-generation biofuels. DME can be prepared in a one-step process from synthesis gas, which is thermodynamically and economically favourable to the two step process consisting of methanol synthesis followed by dehydration of methanol to DME. As the direct process is strongly exothermic, the reaction heat has to be effectively removed from the reaction system in order to maintain a safe and economic operational mode. Direct DME synthesis possesses a high volumetric heat production rate and hence the temperature control is a main challenge. Besides this, parameters such as syngas composition, pressure, contact time and catalytic system affect the conversion and yield. In this work direct DME synthesis from syngas in a microstructured packed bed reactor was demonstrated to operate at practically isothermal conditions. The performance of the catalyst was enhanced by elimination of the undesired phenomena related to the exothermic process, such as hot spot formation and side reactions. The influence of process parameters on methanol selectivity and DME productivity was studied. The highest CO conversion was achieved by a H{sub 2}-rich syngas at

  6. Participation in the U.S. Department of Energy Reactor Sharing Program

    Energy Technology Data Exchange (ETDEWEB)

    Mulder, R. U.; Benneche, P. E.; Hosticka, B.

    1998-09-30

    The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these users institutions is enhanced by the use of the nuclear facilities.

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

    Science.gov (United States)

    2012-06-15

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

  8. Participation in the U.S. Department of Energy Reactor Sharing Program

    Energy Technology Data Exchange (ETDEWEB)

    Mulder, R. U.; Benneche, P. E.; Hosticka, B.

    1998-09-30

    The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these users institutions is enhanced by the use of the nuclear facilities.

  9. A design study of reactor core optimization for direct nuclear heat-to-electricity conversion in a space power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Hidekazu; Takahashi, Makoto; Shimoda, Hiroshi; Takeoka, Satoshi [Kyoto Univ. (Japan); Nakagawa, Masayuki; Kugo, Teruhiko

    1998-01-01

    To propose a new design concept of a nuclear reactor used in the space, research has been conducted on the conceptual design of a new nuclear reactor on the basis of the following three main concepts: (1) Thermionic generation by thermionic fuel elements (TFE), (2) reactivity control by rotary reflector, and (3) reactor cooling by liquid metal. The outcomes of the research are: (1) A calculation algorithm was derived for obtaining convergent conditions by repeating nuclear characteristic calculation and thermal flow characteristic calculation for the space nuclear reactor. (2) Use of this algorithm and the parametric study established that a space nuclear reactor using 97% enriched uranium nitride as the fuel and lithium as the coolant and having a core with a radius of about 25 cm, a height of about 50 cm and a generation efficiency of about 7% can probably be operated continuously for at least more than ten years at 100 kW only by reactivity control by rotary reflector. (3) A new CAD/CAE system was developed to assist design work to optimize the core characteristics of the space nuclear reactor comprehensively. It is composed of the integrated design support system VINDS using virtual reality and the distributed system WINDS to collaboratively support design work using Internet. (N.H.)

  10. Energy Conversion and Storage Program. 1990 annual report

    Energy Technology Data Exchange (ETDEWEB)

    Cairns, E.J.

    1992-03-01

    The Energy Conversion and Storage Program applies chemistry and materials science principles to solve problems in (1) production of new synthetic fuels, (2) development of high-performance rechargeable batteries and fuel cells, (3) development of advanced thermochemical processes for energy conversion, (4) characterization of complex chemical processes, and (5) application of novel materials for energy conversion and transmission. Projects focus on transport-process principles, chemical kinetics, thermodynamics, separation processes, organic and physical chemistry, novel materials, and advanced methods of analysis. Electrochemistry research aims to develop advanced power systems for electric vehicle and stationary energy storage applications. Topics include identification of new electrochemical couples for advanced rechargeable batteries, improvements in battery and fuel-cell materials, and the establishment of engineering principles applicable to electrochemical energy storage and conversion. Chemical Applications research includes topics such as separations, catalysis, fuels, and chemical analyses. Included in this program area are projects to develop improved, energy-efficient methods for processing waste streams from synfuel plants and coal gasifiers. Other research projects seek to identify and characterize the constituents of liquid fuel-system streams and to devise energy-efficient means for their separation. Materials Applications research includes the evaluation of the properties of advanced materials, as well as the development of novel preparation techniques. For example, the use of advanced techniques, such as sputtering and laser ablation, are being used to produce high-temperature superconducting films.

  11. Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Katoh, Yutai [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Wilson, Dane F [ORNL

    2008-08-01

    Since 2002, the Department of Energy's (DOE's) Generation IV Nuclear Energy Systems (Gen IV) Program has addressed the research and development (R&D) necessary to support next-generation nuclear energy systems. The six most promising systems identified for next-generation nuclear energy are described within this roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor-SCWR and the Very High Temperature Reactor-VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor-GFR, the Lead-cooled Fast Reactor-LFR, and the Sodium-cooled Fast Reactor-SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides and may provide an alternative to accelerator-driven systems. At the inception of DOE's Gen IV program, it was decided to significantly pursue five of the six concepts identified in the Gen IV roadmap to determine which of them was most appropriate to meet the needs of future U.S. nuclear power generation. In particular, evaluation of the highly efficient thermal SCWR and VHTR reactors was initiated primarily for energy production, and evaluation of the three fast reactor concepts, SFR, LFR, and GFR, was begun to assess viability for both energy production and their potential contribution to closing the fuel cycle. Within the Gen IV Program itself, only the VHTR class of reactors was selected for continued development. Hence, this document will address the multiple activities under the Gen IV program that contribute to the development of the VHTR. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of

  12. Neutronics Conversion Analyses of the Laue-Langevin Institute (ILL) High Flux Reactor (RHF)

    Energy Technology Data Exchange (ETDEWEB)

    Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Calzavara, Y. [Inst. Laue-Langevin (ILL), Grenoble (France)

    2014-09-30

    The following report describes the neutronics results obtained with the MCNP model of the RHF U7Mo LEU reference design that has been established in 2010 during the feasibility analysis. This work constitutes a complete and detailed neutronics analysis of that LEU design using models that have been significantly improved since 2010 and the release of the feasibility report. When possible, the credibility of the neutronics model is tested by comparing the HEU model results with experimental data or other codes calculations results. The results obtained with the LEU model are systematically compared to the HEU model. The changes applied to the neutronics model lead to better comparisons with experimental data or improved the calculation efficiency but do not challenge the conclusion of the feasibility analysis. If the U7Mo fuel is commercially available, not cost prohibitive, a back-end solution is established and if it is possible to manufacture the proposed element, neutronics analyses show that the performance of the reactor would not be challenged by the conversion to LEU fuel.

  13. Simplifying Chemical Reactor Design by using Molar Quantities Instead of Fractional Conversion.

    Science.gov (United States)

    Brown, Lee F.; Falconer, John L.

    1987-01-01

    Explains the advantages of using molar quantities in chemical reactor design. Advocates the use of differential versions of reactor mass balances rather than the integrated forms. Provides specific examples and cases to illustrate the principles. (ML)

  14. Materials Inventory Database for the Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Kazi Ahmed; Shannon M. Bragg-Sitton

    2013-08-01

    Scientific research involves the purchasing, processing, characterization, and fabrication of many sample materials. The history of such materials can become complicated over their lifetime – materials might be cut into pieces or moved to various storage locations, for example. A database with built-in functions to track these kinds of processes facilitates well-organized research. The Material Inventory Database Accounting System (MIDAS) is an easy-to-use tracking and reference system for such items. The Light Water Reactor Sustainability Program (LWRS), which seeks to advance the long-term reliability and productivity of existing nuclear reactors in the United States through multiple research pathways, proposed MIDAS as an efficient way to organize and track all items used in its research. The database software ensures traceability of all items used in research using built-in functions which can emulate actions on tracked items – fabrication, processing, splitting, and more – by performing operations on the data. MIDAS can recover and display the complete history of any item as a simple report. To ensure the database functions suitably for the organization of research, it was developed alongside a specific experiment to test accident tolerant nuclear fuel cladding under the LWRS Advanced Light Water Reactor Nuclear Fuels Pathway. MIDAS kept track of materials used in this experiment from receipt at the laboratory through all processes, test conduct and, ultimately, post-test analysis. By the end of this process, the database proved to be right tool for this program. The database software will help LWRS more efficiently conduct research experiments, from simple characterization tests to in-reactor experiments. Furthermore, MIDAS is a universal tool that any other research team could use to organize their material inventory.

  15. Light Water Reactor Sustainability Program Reactor Safety Technologies Pathway Technical Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, M. L. [Univ. of Wisconsin, Madison, WI (United States); Peko, D. [US Dept. of Energy, Washington, DC (United States); Farmer, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Rempe, J. [Rempe and Associates LLC, Idaho Falls, ID (United States); Humrickhouse, P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); O' Brien, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Robb, K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauntt, R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Osborn, D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-06-01

    In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safety initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary

  16. Light Water Reactor Sustainability Program: Reactor Safety Technologies Pathway Technical Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, M. L. [Univ. of Wisconsin, Madison, WI (United States)

    2015-06-01

    In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safety initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary

  17. Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Baek J.; Diamond D.; Cuadra, A.; Hanson, A.L.; Cheng, L-Y.; Brown, N.R.

    2012-09-30

    Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a model of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.

  18. Closed-Cycle Engine Program Used to Study Brayton Power Conversion

    Science.gov (United States)

    Johnson, Paul K.

    2005-01-01

    One form of power conversion under consideration in NASA Glenn Research Center's Thermal Energy Conversion Branch is the closed-Brayton-cycle engine. In the tens-of-kilowatts to multimegawatt class, the Brayton engine lends itself to potential space nuclear power applications such as electric propulsion or surface power. The Thermal Energy Conversion Branch has most recently concentrated its Brayton studies on electric propulsion for Prometheus. One piece of software used for evaluating such designs over a limited tradeoff space has been the Closed Cycle Engine Program (CCEP). The CCEP originated in the mid-1980s from a Fortran aircraft engine code known as the Navy/NASA Engine Program (NNEP). Components such as a solar collector, heat exchangers, ducting, a pumped-loop radiator, a nuclear heat source, and radial turbomachinery were added to NNEP, transforming it into a high-fidelity design and performance tool for closed-Brayton-cycle power conversion and heat rejection. CCEP was used in the 1990s in conjunction with the Solar Dynamic Ground Test Demonstration conducted at Glenn. Over the past year, updates were made to CCEP to adapt it for an electric propulsion application. The pumped-loop radiator coolant can now be n-heptane, water, or sodium-potassium (NaK); liquid-metal pump design tables were added to accommodate the NaK fluid. For the reactor and shield, a user can now elect to calculate a higher fidelity mass estimate. In addition, helium-xenon working-fluid properties were recalculated and updated.

  19. Water cooled breeder program summary report (LWBR (Light Water Breeder Reactor) development program)

    Energy Technology Data Exchange (ETDEWEB)

    1987-10-01

    The purpose of the Department of Energy Water Cooled Breeder Program was to demonstrate pratical breeding in a uranium-233/thorium fueled core while producing electrical energy in a commercial water reactor generating station. A demonstration Light Water Breeder Reactor (LWBR) was successfully operated for more than 29,000 effective full power hours in the Shippingport Atomic Power Station. The reactor operated with an availability factor of 76% and had a gross electrical output of 2,128,943,470 kilowatt hours. Following operation, the expended core was examined and no evidence of any fuel element defects was found. Nondestructive assay of 524 fuel rods determined that 1.39 percent more fissile fuel was present at the end of core life than at the beginning, proving that breeding had occurred. This demonstrates the existence of a vast source of electrical energy using plentiful domestic thorium potentially capable of supplying the entire national need for many centuries. To build on the successful design and operation of the Shippingport Breeder Core and to provide the technology to implement this concept, several reactor designs of large breeders and prebreeders were developed for commercial-sized plants of 900--1000 Mw(e) net. This report summarizes the Water Cooled Breeder Program from its inception in 1965 to its completion in 1987. Four hundred thirty-six technical reports are referenced which document the work conducted as part of this program. This work demonstrated that the Light Water Breeder Reactor is a viable alternative as a PWR replacement in the next generation of nuclear reactors. This transition would only require a minimum of change in design and fabrication of the reactor and operation of the plant.

  20. Hydrogenogenic CO Conversion in a Moderately Thermophilic (55 C) Sulfate-Fed Gas Lift Reactor: Competition for CO-Derived H2

    NARCIS (Netherlands)

    Sipma, J.; Lettinga, G.; Stams, A.J.M.; Lens, P.N.L.

    2006-01-01

    Thermophilic (55 °C) sulfate reduction in a gas lift reactor fed with CO gas as the sole electron donor was investigated. The reactor was inoculated with mesophilic granular sludge with a high activity of CO conversion to hydrogen and carbon dioxide at 55 °C. Strong competition for H2 was observed b

  1. Large-scale surface dielectric barrier discharge type reactor : effect of the electric wind on the conversion effectiveness

    Energy Technology Data Exchange (ETDEWEB)

    Jolibois, J. [Univ. de Poitiers, Poitiers (France). Centre national de la recherche scientifique, Laboratoire de Catalyse en Chimie Organique; Poitiers Univ., Futuroscope Chasseneuil Cedex (France). Centre national de la recherche scientifique, Inst. Pprime; Zouzou, N.; Moreau, E. [Poitiers Univ., Futuroscope Chasseneuil Cedex (France). Centre national de la recherche scientifique, Inst. Pprime; Tatibouet, J.M. [Univ. de Poitiers, Poitiers (France). Centre national de la recherche scientifique, Laboratoire de Catalyse en Chimie Organique

    2010-07-01

    Non-thermal plasma (NTP) techniques offer an innovative approach for air pollution reduction. Most studies in NTP techniques use volumetric discharge reactors with small dimensions and low flow rates at laboratory scale. The objective of this study was to develop an air pollution control plasma reactor at industrial scale with surface discharge. Propene (C{sub 3}H{sub 6}) was oxidized at high flow rates in a large-scale plasma reactor based on surface dielectric barrier discharge (DBD). Three different configurations of surface discharges were tested with 15 ppm of C{sub 3}H{sub 6} in air at ambient temperature for a flow rate of 50 m{sup 3} per hour. The properties of these different surface discharges were analyzed using chemical measurements and 3 component particle image velocimetry (PIV) measurements. PIV measurements were used characterize the effect of the electric wind on the polluted gas airflow inside the reactor and to explain the differences of effectiveness of the three tested plasma generators. For the three plasma generators, a propene oxidation of up to 45 percent was obtained at one J per liter. The electric wind produced by the surface discharge resulted in the formation of vortices inside the plasma reactor. This electric wind can increase gas mixing inside the plasma reactor and therefore plays a key role in conversion efficiency. It was concluded that the electric wind produced by surface discharges enables the use of this type of discharge for VOC elimination at high flow rate, with the same effectiveness of volumetric discharges. 5 refs., 10 figs.

  2. Civilian Power Program. Part 1, Summary, Current status of reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Author, Not Given

    1959-09-01

    This study group covered the following: delineation of the specific objectives of the overall US AEC civilian power reactor program, technical objectives of each reactor concept, preparation of a chronological development program for each reactor concept, evaluation of the economic potential of each reactor type, a program to encourage the the development, and yardsticks for measuring the development. Results were used for policy review by AEC, program direction, authorization and appropriation requests, etc. This evaluation encompassed civilian power reactors rated at 25 MW(e) or larger and related experimental facilities and R&D. This Part I summarizes the significant results of the comprehensive effort to determine the current technical and economic status for each reactor concept; it is based on the 8 individual technical status reports (Part III).

  3. The upgrade and conversion of the ET-RR-1 research reactor using plate type fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Ashoub, N. [Reactor Physics Dept., Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt); Saleh, H.G. [Faculty of Girls for Arts and Education, Ain-Shams Univ., Cairo (Egypt)

    2001-11-01

    The ET-RR-1 research reactor has been operated at 2 MW since 1961 using EK-10 fuel elements with 10% enriched uranium. The reactor has been used for nuclear applied research and isotope production. In order to upgrade the reactor power to a reasonable limit facing up-to-date uses, core conversion by a new type of fuel element available is necessary. Two fuel elements in plate type are suggested in this study to be used in the ET-RR-1 reactor core rather than the utilized ones. The first element has a dimension of 8 x 8 x 50 cm and consists of 19.7% enriched uranium, which is typical for that utilized in the ET-RR-2 reactor, but with a different length. The other element is proposed with a dimension of 7 x 7 x 50 cm and has the same uranium enrichment. To accomplish safety requirements for these fuel elements, thermal-hydraulic evaluation has been carried out using the PARET code. To reach a core conversion of the ET-RR-1 reactor with the above two types of fuel elements, neutronic calculations have been performed using WIMSD4, DIXY2 and EREBUS codes. Some important nuclear parameters needed in the physical design of the reactor were calculated and included in this study. (orig.) [German] Der ET-RR-1 Forschungsreaktor wird seit 1961 unter Verwendung von EK-10 Brennelementen mit einer Leistung von 2 MW betrieben. Der Reaktor wird in der angewandten Forschung und zur Isotopenherstellung eingesetzt. Um die Reaktorleistung im Hinblick auf eine zeitgemaesse Nutzung der Anlage in einem vernuenftigen Mass zu erhoehen, ist eine Umwandlung des Kerns durch Verwendung neuartiger Brennelemente noetig. In der vorliegenden Untersuchung wird vorgeschlagen, anstelle der z. Z. verwendeten Elemente zwei neue, plattenfoermige Brennelemente zu verwenden. Das erste Element hat eine Groesse von 8 x 8 x 50 cm und besteht aus 19,7% angereichertem Uran, was den im ET-RR-2 Reaktor verwendeten Elementen entspricht, allerdings mit einer anderen Groesse. Das zweite Element hat die gleiche

  4. A study of concentrated acid hydrolysis conversion of lignocellulosic materials to sugars using a co-rotating twin-screw reactor extruder and plug flow reactor

    Science.gov (United States)

    Miller, William Scott

    Concerns about the ability of petroleum to continue supplying ever increasing global energy demands, at a price capable of generating continued economic growth, have spurred innovative research in the field of alternative energy. One alternative energy option that has the ability to provide long-term sustainable energy supplies for the global energy market is the conversion of lignocellulosic materials, via acid hydrolysis, to fermentable sugars for the production of fuel grade ethanol. This research demonstrates the ability of a co-rotating twin-screw reactor extruder and plug flow reactor to continuously convert lignocellulosic materials to fermentable sugars using high temperature concentrated acid hydrolysis. In addition to demonstrating continuous operation of the two-stage concentrated acid hydrolysis system, a number of design of experiments were conducted to model the twin-screw performance and maximize its ability to effectively solubilize lignocellulosic feedstocks in the high shear, elevated temperature, concentrated acid environment. These studies produced a base case twin-screw operating condition used to generate a standard extrudate composition for an extensive high temperature acid hydrolysis batch reactor kinetic modeling study. In this study a number of nonlinear and linear regression analyses were undertaken so that the concentration of less resistant cellulose, or the amount of solublilized extrudate cellulose, resistant cellulose, or non-solubilized extrudate cellulose, glucose, and decomposition products could be obtained as a function of time, temperature, and acid concentration. This study demonstrated that the theoretical cellulose conversion of 51% was limited by the amount of solubilized polysaccharides that could be produced in the twin-screw pretreatment. Further experimentation, showing twin-screw pretreatment lignocellulosic versatility, produced nearly identical results as the southern yellow pine sawdust experiments that were

  5. Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Kalcheva, S [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Sikik, E [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

    2016-09-01

    BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water. The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cm2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident. A feasibility study for the conversion of the BR2 reactor from highly-enriched uranium (HEU) to low-enriched uranium (LEU) fuel was previously performed to verify it can operate safely at the same maximum nominal steady-state heat flux. An assessment was also performed to quantify the heat fluxes at which the onset of flow instability and critical heat flux occur for each fuel type. This document updates and expands these results for the current representative core configuration (assuming a fresh beryllium matrix) by evaluating the onset of nucleate boiling (ONB), onset of fully developed nucleate boiling (FDNB), onset of flow instability (OFI) and critical heat flux (CHF).

  6. Plasmon-induced selective carbon dioxide conversion on earth-abundant aluminum-cuprous oxide antenna-reactor nanoparticles.

    Science.gov (United States)

    Robatjazi, Hossein; Zhao, Hangqi; Swearer, Dayne F; Hogan, Nathaniel J; Zhou, Linan; Alabastri, Alessandro; McClain, Michael J; Nordlander, Peter; Halas, Naomi J

    2017-06-21

    The rational combination of plasmonic nanoantennas with active transition metal-based catalysts, known as 'antenna-reactor' nanostructures, holds promise to expand the scope of chemical reactions possible with plasmonic photocatalysis. Here, we report earth-abundant embedded aluminum in cuprous oxide antenna-reactor heterostructures that operate more effectively and selectively for the reverse water-gas shift reaction under milder illumination than in conventional thermal conditions. Through rigorous comparison of the spatial temperature profile, optical absorption, and integrated electric field enhancement of the catalyst, we have been able to distinguish between competing photothermal and hot-carrier driven mechanistic pathways. The antenna-reactor geometry efficiently harnesses the plasmon resonance of aluminum to supply energetic hot-carriers and increases optical absorption in cuprous oxide for selective carbon dioxide conversion to carbon monoxide with visible light. The transition from noble metals to aluminum based antenna-reactor heterostructures in plasmonic photocatalysis provides a sustainable route to high-value chemicals and reaffirms the practical potential of plasmon-mediated chemical transformations.Plasmon-enhanced photocatalysis holds promise for the control of chemical reactions. Here the authors report an Al@Cu2O heterostructure based on earth abundant materials to transform CO2 into CO at significantly milder conditions.

  7. TERRAPOWER, LLC TRAVELING WAVE REACTOR DEVELOPMENT PROGRAM OVERVIEW

    Directory of Open Access Journals (Sweden)

    PAVEL HEJZLAR

    2013-11-01

    burden. This paper describes the origins and current status of the TWR development program at TerraPower, LLC. Some of the areas covered include the key TWR design challenges and brief descriptions of TWR-Prototype (TWR-P reactor. Selected information on the TWR-P core designs are also provided in the areas of neutronic, thermal hydraulic and fuel performance. The TWR-P plant design is also described in such areas as; system design descriptions, mechanical design, and safety performance.

  8. Neutronic analysis for core conversion (HEU–LEU of the low power research reactor using the MCNP4C code

    Directory of Open Access Journals (Sweden)

    Aldawahra Saadou

    2015-06-01

    Full Text Available Comparative studies for conversion of the fuel from HEU to LEU in the miniature neutron source reactor (MNSR have been performed using the MCNP4C code. The HEU fuel (UAl4-Al, 90% enriched with Al clad and LEU (UO2 12.6% enriched with zircaloy-4 alloy clad cores have been analyzed in this study. The existing HEU core of MNSR was analyzed to validate the neutronic model of reactor, while the LEU core was studied to prove the possibility of fuel conversion of the existing HEU core. The proposed LEU core contained the same number of fuel pins as the HEU core. All other structure materials and dimensions of HEU and LEU cores were the same except the increase in the radius of control rod material from 0.195 to 0.205 cm and keeping the outer diameter of the control rod unchanged in the LEU core. The effective multiplication factor (keff, excess reactivity (ρex, control rod worth (CRW, shutdown margin (SDM, safety reactivity factor (SRF, delayed neutron fraction (βeff and the neutron fluxes in the irradiation tubes for the existing and the potential LEU fuel were investigated. The results showed that the safety parameters and the neutron fluxes in the irradiation tubes of the LEU fuels were in good agreements with the HEU results. Therefore, the LEU fuel was validated to be a suitable choice for fuel conversion of the MNSR in the future.

  9. A calorimetric study of energy conversion efficiency of a sonochemical reactor at 500 kHz for organic solvents.

    Science.gov (United States)

    Toma, Maricela; Fukutomi, Satoshi; Asakura, Yoshiyuki; Koda, Shinobu

    2011-01-01

    It would seem that the economic viability is yet to be established for a great number of sonochemical processes, owning to their perfectible ultrasonic equipments. Industrial scale sonoreactors may become more important as a result of mastering the parameters with influence on their energy balance. This work related the solvent type to the energy efficiency as the first step of a complex study aiming to assess the energy balance of sonochemical reactors at 500 kHz. Quantitative measurements of ultrasonic power for water and 10 pure organic solvents were performed by calorimetry for a cylindrically shaped sonochemical reactor with a bottom mounted vibrating plate. It was found that the ultrasonic power is strongly related to the solvent, the energy conversion for organic liquids is half from that of water and there is a drop in energy efficiency for filling levels up to 250 mm organic solvents. Surface tension, viscosity and vapor pressure influence the energy conversion for organic solvents, but it is difficult explain these findings based on physical properties of solvents alone. The apparent intensity of the atomization process shows a good agreement with the experimentally determined values for energy conversion for water and the solvent group studied here. This study revealed that to attain the same ultrasonic power level, more electrical energy is need for organic solvents as compared to water. The energy balance equation has been defined based on these findings by considering an energy term for atomization.

  10. Reactor Materials Program: Mechanical properties of irradiated Types 304 and 304L stainless steel weldment components

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.L.; Caskey, G.R. Jr.

    1991-12-01

    The vessels (reactor tanks) of the Savannah River Site nuclear production reactors constructed in the 1950`s are comprised of Type 304 stainless steel with Type 308 stainless steel weld filler. Irradiation exposure to the reactor tank sidewalls through reactor operation has caused a change in the mechanical properties of these materials. A database of as-irradiated mechanical properties for site-specific materials and irradiation conditions has been produced for reactor tank structural analyses and to quantify the effects of radiation-induced materials degradation for evaluating reactor service life. The data has been collected from the SRL Reactor Materials Program (RMP) irradiations and testing of archival stainless steel weldment components and from previous SRL programs to measure properties of irradiated reactor Thermal Shield weldments and reactor tank (R-tank) sidewall material. Irradiation programs of the RMP are designed to quantify mechanical properties at tank operating temperatures following irradiation to present and future tank wall maximum exposure conditions. The exposure conditions are characterized in terms of fast neutron fluence (E{sub n} > 0.1 MeV) and displacements per atom (dpa){sup 3}. Tensile properties, Charpy-V notch toughness, and elastic-plastic fracture toughness were measured for base, weld, and weld heat-affected zone (HAZ) weldment components from archival piping specimens following a Screening Irradiation in the University of Buffalo Reactor (UBR) and following a Full-Term Irradiation in the High Flux Isotope Reactor (HFIR).

  11. Reactor Materials Program: Mechanical properties of irradiated Types 304 and 304L stainless steel weldment components

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.L.; Caskey, G.R. Jr.

    1991-12-01

    The vessels (reactor tanks) of the Savannah River Site nuclear production reactors constructed in the 1950's are comprised of Type 304 stainless steel with Type 308 stainless steel weld filler. Irradiation exposure to the reactor tank sidewalls through reactor operation has caused a change in the mechanical properties of these materials. A database of as-irradiated mechanical properties for site-specific materials and irradiation conditions has been produced for reactor tank structural analyses and to quantify the effects of radiation-induced materials degradation for evaluating reactor service life. The data has been collected from the SRL Reactor Materials Program (RMP) irradiations and testing of archival stainless steel weldment components and from previous SRL programs to measure properties of irradiated reactor Thermal Shield weldments and reactor tank (R-tank) sidewall material. Irradiation programs of the RMP are designed to quantify mechanical properties at tank operating temperatures following irradiation to present and future tank wall maximum exposure conditions. The exposure conditions are characterized in terms of fast neutron fluence (E{sub n} > 0.1 MeV) and displacements per atom (dpa){sup 3}. Tensile properties, Charpy-V notch toughness, and elastic-plastic fracture toughness were measured for base, weld, and weld heat-affected zone (HAZ) weldment components from archival piping specimens following a Screening Irradiation in the University of Buffalo Reactor (UBR) and following a Full-Term Irradiation in the High Flux Isotope Reactor (HFIR).

  12. Syngas Production By Thermochemical Conversion Of H2o And Co2 Mixtures Using A Novel Reactor Design

    Energy Technology Data Exchange (ETDEWEB)

    Pearlman, Howard [Advanced Cooling Technologies, Inc, Lancaster, PA (United States); Chen, Chien-Hua [Advanced Cooling Technologies, Inc, Lancaster, PA (United States)

    2014-08-27

    The Department of Energy awarded Advanced Cooling Technologies, Inc. (ACT) an SBIR Phase II contract (#DE-SC0004729) to develop a high-temperature solar thermochemical reactor for syngas production using water and/or carbon dioxide as feedstocks. The technology aims to provide a renewable and sustainable alternative to fossil fuels, promote energy independence and mitigate adverse issues associated with climate change by essentially recycling carbon from carbon dioxide emitted by the combustion of hydrocarbon fuels. To commercialize the technology and drive down the cost of solar fuels, new advances are needed in materials development and reactor design, both of which are integral elements in this program.

  13. ECUT (Energy Conversion and Utilization Technologies) program: Biocatalysis project

    Science.gov (United States)

    Baresi, Larry

    1989-03-01

    The Annual Report presents the fiscal year (FY) 1988 research activities and accomplishments, for the Biocatalysis Project of the U.S. Department of Energy, Energy Conversion and Utilization Technologies (ECUT) Division. The ECUT Biocatalysis Project is managed by the Jet Propulsion Laboratory, California Institute of Technology. The Biocatalysis Project is a mission-oriented, applied research and exploratory development activity directed toward resolution of the major generic technical barriers that impede the development of biologically catalyzed commercial chemical production. The approach toward achieving project objectives involves an integrated participation of universities, industrial companies and government research laboratories. The Project's technical activities were organized into three work elements: (1) The Molecular Modeling and Applied Genetics work element includes research on modeling of biological systems, developing rigorous methods for the prediction of three-dimensional (tertiary) protein structure from the amino acid sequence (primary structure) for designing new biocatalysis, defining kinetic models of biocatalyst reactivity, and developing genetically engineered solutions to the generic technical barriers that preclude widespread application of biocatalysis. (2) The Bioprocess Engineering work element supports efforts in novel bioreactor concepts that are likely to lead to substantially higher levels of reactor productivity, product yields and lower separation energetics. Results of work within this work element will be used to establish the technical feasibility of critical bioprocess monitoring and control subsystems. (3) The Bioprocess Design and Assessment work element attempts to develop procedures (via user-friendly computer software) for assessing the energy-economics of biocatalyzed chemical production processes, and initiation of technology transfer for advanced bioprocesses.

  14. Direct conversion of methane to aromatics in a catalytic co-ionic membrane reactor

    National Research Council Canada - National Science Library

    Morejudo, S H; Zanón, R; Escolástico, S; Yuste-Tirados, I; Malerød-Fjeld, H; Vestre, P K; Coors, W G; Martínez, A; Norby, T; Serra, J M; Kjølseth, C

    2016-01-01

    .... We show that integration of an electrochemical BaZrO3-based membrane exhibiting both proton and oxide ion conductivity into an MDA reactor gives rise to high aromatic yields and improved catalyst stability...

  15. Fabrication of uranium dioxide fuel pellets in support of a SLOWPOKE-2 research reactor HEU to LEU core conversion

    Energy Technology Data Exchange (ETDEWEB)

    Bergeron, A. [Aomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2014-07-01

    The International Centre for Environmental and Nuclear Sciences (ICENS) at the University of the West Indies in Jamaica operates a SLOWPOKE-2 research reactor that is currently fuelled with highly-enriched uranium (HEU). As part of the Global Threat Reduction Initiative, Atomic Energy of Canada Ltd. has been subcontracted to fabricate low-enriched uranium (LEU) fuel for the ICENS SLOWPOKE-2. The low enriched uranium core consists of a fuel cage containing uranium dioxide fuelled elements. This paper describes the fabrication of the low-enriched uranium dioxide fuel pellets for the SLOWPOKE-2 core conversion. (author)

  16. Gas Turbine Energy Conversion Systems for Nuclear Power Plants Applicable to LiFTR Liquid Fluoride Thorium Reactor Technology

    Science.gov (United States)

    Juhasz, Albert J.

    2014-01-01

    This panel plans to cover thermal energy and electric power production issues facing our nation and the world over the next decades, with relevant technologies ranging from near term to mid-and far term.Although the main focus will be on ground based plants to provide baseload electric power, energy conversion systems (ECS) for space are also included, with solar- or nuclear energy sources for output power levels ranging tens of Watts to kilo-Watts for unmanned spacecraft, and eventual mega-Watts for lunar outposts and planetary surface colonies. Implications of these technologies on future terrestrial energy systems, combined with advanced fracking, are touched upon.Thorium based reactors, and nuclear fusion along with suitable gas turbine energy conversion systems (ECS) will also be considered by the panelists. The characteristics of the above mentioned ECS will be described, both in terms of their overall energy utilization effectiveness and also with regard to climactic effects due to exhaust emissions.

  17. Conversion of Tobacco Biomass to Flavor Components by Means of Microwave and Parr Reactors

    Directory of Open Access Journals (Sweden)

    Ara Katayoun Mahdavi

    2017-04-01

    Full Text Available In the present work, microwave and Parr reactors were utilized for synthesis of pyrazines from plant-based biomass in the presence of ammonia and different amino acids. Using these techniques led to synthesis of a relatively wide range of pyrazines with sweet odor and chocolate-like smell. The optimum synthetic conditions to have maximum pyrazine yield for both the microwave and Parr reactions were 41 g of fructose/glucose syrup derived from cellulosic biomass, 28 mL NH4OH (30%, and 0.96 g L-threonine, 0.56 g L-valine, 0.5 g L-leucine, and 0.5 g L-isoleucine at 120 °C for 30 min. Quantitative results obtained via gas chromatography-mass spectrometry (GC-MS using the traditional open-heated oil bath method have been compared with data obtained via microwave and Parr reactors. In these two latter methods, sealed vessels under high pressure and higher temperature were used. The yield of synthesized pyrazines increased dramatically with both microwave and Parr reactors. Surprisingly, the yield of synthesized pyrazines was both reproducible and nearly two times higher via the Parr reactor than that observed with the microwave reactor under comparable conditions.

  18. Pretreated Landfill Gas Conversion Process via a Catalytic Membrane Reactor for Renewable Combined Fuel Cell-Power Generation

    Directory of Open Access Journals (Sweden)

    Zoe Ziaka

    2013-01-01

    Full Text Available A new landfill gas-based reforming catalytic processing system for the conversion of gaseous hydrocarbons, such as incoming methane to hydrogen and carbon oxide mixtures, is described and analyzed. The exit synthesis gas (syn-gas is fed to power effectively high-temperature fuel cells such as SOFC types for combined efficient electricity generation. The current research work is also referred on the description and design aspects of permreactors (permeable reformers carrying the same type of landfill gas-reforming reactions. Membrane reactors is a new technology that can be applied efficiently in such systems. Membrane reactors seem to perform better than the nonmembrane traditional reactors. The aim of this research includes turnkey system and process development for the landfill-based power generation and fuel cell industries. Also, a discussion of the efficient utilization of landfill and waste type resources for combined green-type/renewable power generation with increased processing capacity and efficiency via fuel cell systems is taking place. Moreover, pollution reduction is an additional design consideration in the current catalytic processors fuel cell cycles.

  19. Participation in the U.S. Department of Energy Reactor Sharing Program. Progress report

    Energy Technology Data Exchange (ETDEWEB)

    Mulder, R.U.; Benneche, P.E.; Hosticka, B.

    1997-03-01

    The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would not be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA for use at their institutions. These areas are discussed here.

  20. Light Water Reactor Sustainability Program. Digital Architecture Requirements

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, Kenneth [Idaho National Lab. (INL), Idaho Falls, ID (United States); Oxstrand, Johanna [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    The Digital Architecture effort is a part of the Department of Energy (DOE) sponsored Light-Water Reactor Sustainability (LWRS) Program conducted at Idaho National Laboratory (INL). The LWRS program is performed in close collaboration with industry research and development (R&D) programs that provides the technical foundations for licensing and managing the long-term, safe, and economical operation of current nuclear power plants (NPPs). One of the primary missions of the LWRS program is to help the U.S. nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. Therefore, a major objective of the LWRS program is the development of a seamless digital environment for plant operations and support by integrating information from plant systems with plant processes for nuclear workers through an array of interconnected technologies. In order to get the most benefits of the advanced technology suggested by the different research activities in the LWRS program, the nuclear utilities need a digital architecture in place to support the technology. A digital architecture can be defined as a collection of information technology (IT) capabilities needed to support and integrate a wide-spectrum of real-time digital capabilities for nuclear power plant performance improvements. It is not hard to imagine that many processes within the plant can be largely improved from both a system and human performance perspective by utilizing a plant wide (or near plant wide) wireless network. For example, a plant wide wireless network allows for real time plant status information to easily be accessed in the control room, field workers’ computer-based procedures can be updated based on the real time plant status, and status on ongoing procedures can be incorporated into smart schedules in the outage command center to allow for more accurate planning of critical tasks. The goal

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-10-15

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

  2. Final report. U.S. Department of Energy University Reactor Sharing Program

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, John A

    2003-01-21

    Activities supported at the MIT Nuclear Reactor Laboratory under the U.S. DOE University Reactor Sharing Program are reported for Grant DE FG02-95NE38121 (September 16, 1995 through May 31, 2002). These activities fell under four subcategories: support for research at thesis and post-doctoral levels, support for college-level laboratory exercises, support for reactor tours/lectures on nuclear energy, and support for science fair participants.

  3. Goals of care conversation teaching in residency - a cross-sectional survey of postgraduate program directors.

    Science.gov (United States)

    Roze des Ordons, Amanda; Kassam, Aliya; Simon, Jessica

    2017-01-06

    Residents are commonly involved in establishing goals of care for hospitalized patients. While education can improve the quality of these conversations, whether and how postgraduate training programs integrate such teaching into their curricula is not well established. The objective of this study was to characterize perceptions of current teaching and assessment of goals of care conversations, and program director interest in associated curricular integration. An electronic survey was sent to all postgraduate program directors at the University of Calgary. Quantitative data was analyzed using descriptive statistics and qualitative comments were analyzed using thematic analysis. The survey response rate was 34% (22/64). Formal goals of care conversation teaching is incorporated into 63% of responding programs, and most commonly involves lectures. Informal teaching occurs in 86% of programs, involving discussion, direct observation and role modeling in the clinical setting. Seventy-three percent of programs assess goals of care conversation skills, mostly in the clinical setting through feedback. Program directors believe that over two-thirds of clinical faculty are prepared to teach goals of care conversations, and are interested in resources to teach and assess goals of care conversations. Themes that emerged include 1) general perceptions, 2) need for teaching, 3) ideas for teaching, and 4) assessment of goals of care conversations. The majority of residency training programs at the University of Calgary incorporate some goals of care conversation teaching and assessment into their curricula. Program directors are interested in resources to improve teaching and assessment of goals of care conversations.

  4. The effect of sulphate on methanol conversion in mesophilic upflow anaerobic sludge bed reactors

    NARCIS (Netherlands)

    Weijma, J.; Chi, T.M.; Hulshoff Pol, L.W.; Stams, A.J.M.; Lettinga, G.

    2003-01-01

    Mesophilic (30 °C) upflow anaerobic sludge bed reactors were fed with an influent containing sulphate (2 g l-1) and methanol (1.33 g l-1). More than 90% of the methanol was mineralised to methane, while only ˜5–10% of the methanol was used for sulphate reduction. This pattern was independent of

  5. Characterization of the fast neutron irradiation facility of the Portuguese Research Reactor after core conversion.

    Science.gov (United States)

    Marques, J G; Sousa, M; Santos, J P; Fernandes, A C

    2011-08-01

    The fast neutron irradiation facility of the Portuguese Research Reactor was characterized after the reduction in uranium enrichment and rearrangement of the core configuration. In this work we report on the determination of the hardness parameter and the 1MeV equivalent neutron flux along the facility, in the new irradiation conditions, following ASTM E722 standard.

  6. Conversion of low density polyethylene into petrochemical feedstocks using a continuous screw kiln reactor

    Energy Technology Data Exchange (ETDEWEB)

    Serrano, D.P.; Aguado, J.; Escola, J.M. [ESCET, Rey Juan Carlos University, c/Tulipan s/n, 28933 Mostoles, Madrid (Spain); Garagorri, E. [Chemical Engineering Department, Complutense University of Madrid, Av. Complutense s/n, 28040 Madrid (Spain)

    2001-04-01

    Thermal and catalytic degradation of low-density polyethylene (LDPE) has been investigated using a screw kiln reactor provided with two zones of reaction temperature. Thermal degradation experiments carried out at different temperatures and screw speeds have shown that this continuous system is suitable for the LDPE degradation, product outputs up to c.a. 100 g/h being obtained. Compared to a conventional batch reactor, the screw kiln system leads to a lower formation of gaseous products, whereas overcracking of the heavy fractions is also reduced. These differences are probably originated by the intimate contact and the same residence times for all the product fractions that exist within the screw reactor, which is in contrast with the selective and fast withdrawal of volatile products taking place in the batch system. In the catalytic experiments, a mesoporous MCM-41 type aluminosilicate has been used as catalyst, being continuously fed to the screw reactor mixed with the raw plastic material. In these conditions, yields up to 80% towards hydrocarbons within the gasoline range (C{sub 5}-C{sub 12}) have been obtained. Moreover, high amounts of C{sub 7} and C{sub 8} hydrocarbons are present in the gasoline fractions, which is assigned to catalytic oligomerization reactions that selectively affect to C{sub 3} and C{sub 4} gaseous hydrocarbons.

  7. SEPARATION OF HYDROGEN AND CARBON DIOXIDE USING A NOVEL MEMBRANE REACTOR IN ADVANCED FOSSIL ENERGY CONVERSION PROCESS

    Energy Technology Data Exchange (ETDEWEB)

    Shamsuddin Ilias

    2005-02-03

    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.

  8. Biological CO2 conversion to acetate in subsurface coal-sand formation using a high-pressure reactor system

    Directory of Open Access Journals (Sweden)

    Yoko eOhtomo

    2013-12-01

    Full Text Available Geological CO2 sequestration in unmineable subsurface oil/gas fields and coal formations has been proposed as a means of reducing anthropogenic greenhouse gasses in the atmosphere. However, the feasibility of injecting CO2 into subsurface depends upon a variety of geological and economic conditions, and the ecological consequences are largely unpredictable. In this study, we developed a new flow-through-type reactor system to examine potential geophysical, geochemical and microbiological impacts associated with CO2 injection by simulating in situ pressure (0–100 MPa and temperature (0–70°C conditions. Using the reactor system, anaerobic artificial fluid and CO2 (flow rate: 0.002 and 0.00001 mL/min, respectively were continuously supplemented into a column comprised of bituminous coal and sand under a pore pressure of 40 MPa (confined pressure: 41 MPa at 40°C for 56 days. 16S rRNA gene analysis of the bacterial components showed distinct spatial separation of the predominant taxa in the coal and sand over the course of the experiment. Cultivation experiments using sub-sampled fluids revealed that some microbes survived, or were metabolically active, under CO2-rich conditions. However, no methanogens were activated during the experiment, even though hydrogenotrophic and methylotrophic methanogens were obtained from conventional batch-type cultivation at 20°C. During the reactor experiment, the acetate and methanol concentration in the fluids increased while the δ13Cacetate, H2 and CO2 concentrations decreased, indicating the occurrence of homo-acetogenesis. 16S rRNA genes of homo-acetogenic spore-forming bacteria related to the genus Sporomusa were consistently detected from the sandstone after the reactor experiment. Our results suggest that the injection of CO2 into a natural coal-sand formation preferentially stimulates homo-acetogenesis rather than methanogenesis, and that this process is accompanied by biogenic CO2 conversion to

  9. Magnetic Fustion Reactor Design Studies Program final report, 1 July 1986--30 September 1986

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1986-09-30

    This report presents progress reported during the period, 7/1/86 - 9/30/86 for the Technical Support Services (TSS) for the Magnetic Fusion Reactor Design Studies Program. Tasks reported include: systems studies work plan, normalization of reactor design studies, interpretation of design study activities, research and development plan, conference support, and reports generated.

  10. ECUT (Energy Conversion and Utilization Technologies Program). Biocatalysis Project

    Science.gov (United States)

    1986-01-01

    Presented are the FY 1985 accomplishments, activities, and planned research efforts of the Biocatalysis Project of the U.S. Department of Energy, Energy Conversion and Utilization Technologies (ECUT) Program. The Project's technical activities were organized as follows: In the Molecular Modeling and Applied Genetics work element, research focused on (1) modeling and simulation studies to establish the physiological basis of high temperature tolerance in a selected enzyme and the catalytic mechanisms of three species of another enzyme, and (2) determining the degree of plasmid amplification and stability of several DNA bacterial strains. In the Bioprocess Engineering work element, research focused on (1) studies of plasmid propagation and the generation of models, (2) developing methods for preparing immobilized biocatalyst beads, and (3) developing an enzyme encapsulation method. In the Process Design and Analysis work element, research focused on (1) further refinement of a test case simulation of the economics and energy efficiency of alternative biocatalyzed production processes, (2) developing a candidate bioprocess to determine the potential for reduced energy consumption and facility/operating costs, and (3) a techno-economic assessment of potential advancements in microbial ammonia production.

  11. Hydrogen-Permeable Tubular Membrane Reactor: Promoting Conversion and Product Selectivity for Non-Oxidative Activation of Methane over an Fe©SiO2 Catalyst.

    Science.gov (United States)

    Sakbodin, Mann; Wu, Yiqing; Oh, Su Cheun; Wachsman, Eric D; Liu, Dongxia

    2016-12-23

    Non-oxidative methane conversion over Fe©SiO2 catalyst was studied for the first time in a hydrogen (H2 ) permeable tubular membrane reactor. The membrane reactor is composed of a mixed ionic-electronic SrCe0.7 Zr0.2 Eu0.1 O3-δ thin film (≈20 μm) supported on the outer surface of a one-end capped porous SrCe0.8 Zr0.2 O3-δ tube. Significant improvement in CH4 conversion was achieved upon H2 removal from the membrane reactor compared to that in a fixed-bed reactor. The Fe©SiO2 catalyst in the H2 permeable membrane reactor demonstrated a stable ≈30 % C2+ single-pass yield, with up to 30 % CH4 conversion and 99 % selectivity to C2 (ethylene and acetylene) and aromatic (benzene and naphthalene) products, at the tested conditions. The selectivity towards C2 or aromatics was manipulated purposely by adding H2 into or removing H2 from the membrane reactor feed and permeate gas streams. © 2016 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  12. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Cyrus M [ORNL; Nanstad, Randy K [ORNL; Clayton, Dwight A [ORNL; Matlack, Katie [Georgia Institute of Technology; Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL); Light, Glenn [Southwest Research Institute, San Antonio

    2012-09-01

    The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

  13. A single step methane conversion into synthetic fuels using microplasma reactor

    NARCIS (Netherlands)

    Nozaki, Tomohiro; Agiral, A.; Gardeniers, Johannes G.E.; Yuzawa, Shuhei; Okazaki, Ken

    2011-01-01

    Direct conversion of natural gas into synthetic fuels such as methanol attracts keen attention because direct process can reduce capital and operating costs of high temperature, energy intensive, multi-step processes. We report a direct and selective synthesis of organic oxygenates such as methanol,

  14. Modeling and sizing of the heat exchangers of a new supercritical CO{sub 2} Brayton power cycle for energy conversion for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Serrano, I.P.; Cantizano, A.; Linares, J.I., E-mail: linares@upcomillas.es; Moratilla, B.Y.

    2014-10-15

    Highlights: •We propose a procedure to model the heat exchangers of a S-CO2 Brayton power cycle. •Discretization in sub-heat exchangers is performed due to complex behavior of CO{sub 2}. •Different correlations have been tested, verifying them with CFD when necessary. •Obtained sizes are agree with usual values of printed circuit heat exchangers. -- Abstract: TECNO{sub F}US is a research program financed by the Spanish Government to develop technologies related to a dual-coolant (He/Pb–Li) breeding blanket design concept including the auxiliary systems for a future power reactor (DEMO). One of the main issues of this program is the optimization of heat recovery from the reactor and its conversion into electrical power. This paper is focused on the methodology employed for the design and sizing of all the heat exchangers of the supercritical CO{sub 2} Brayton power cycle (S-CO2) proposed by the authors. Due to the large pressure difference between the fluids, and also to their compactness, Printed Circuit Heat Exchangers (PCHE) are suggested in literature for these type of cycles. Because of the complex behavior of CO{sub 2}, their design is performed by a numerical discretization into sub-heat exchangers, thus a higher precision is reached when the thermal properties of the fluids vary along the heat exchanger. Different empirical correlations for the pressure drop and the Nusselt number have been coupled and assessed. The design of the precooler (PC) and the low temperature recuperator (LTR) is also verified by simulations using CFD because of the near-critical behavior of CO{sub 2}. The size of all of the heat exchangers of the cycle have been assessed.

  15. Scaling of surface-plasma reactors with a significantly increased energy density for NO conversion.

    Science.gov (United States)

    Malik, Muhammad Arif; Xiao, Shu; Schoenbach, Karl H

    2012-03-30

    Comparative studies revealed that surface plasmas developing along a solid-gas interface are significantly more effective and energy efficient for remediation of toxic pollutants in air than conventional plasmas propagating in air. Scaling of the surface plasma reactors to large volumes by operating them in parallel suffers from a serious problem of adverse effects of the space charges generated at the dielectric surfaces of the neighboring discharge chambers. This study revealed that a conductive foil on the cathode potential placed between the dielectric plates as a shield not only decoupled the discharges, but also increased the electrical power deposited in the reactor by a factor of about forty over the electrical power level obtained without shielding and without loss of efficiency for NO removal. The shield had no negative effect on efficiency, which is verified by the fact that the energy costs for 50% NO removal were about 60 eV/molecule and the energy constant, k(E), was about 0.02 L/J in both the shielded and unshielded cases.

  16. A neutronic feasibility study for LEU conversion of the IR-8 research reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Deen, J. R.

    1998-10-22

    Equilibrium fuel cycle comparisons for the IR-8 research reactor were made for HEU(90%), HEU(36%), and LEU (19.75%) fuel assembly (FA) designs using three dimensional multi-group diffusion theory models benchmarked to detailed Monte Carlo models of the reactor. Comparisons were made of changes in reactivity, cycle length, average {sup 235}U discharge burnup, thermal neutron flux, and control rod worths for the 90% and 36% enriched IRT-3M fuel assembly and the 19.75% enriched IRT-4M fuel assembly with the same fuel management strategy. The results of these comparisons showed that a uranium density of 3.5 g/cm{sup 3} in the fuel meat would be required in the LEU IRT-4M fuel assembly to match the cycle length of the HEU(90%) IRT-3M FA and an LEU density of 3.7 g/cm{sup 3} is needed to match the cycle length of the HEU(36%) IRT-3M FA.

  17. Methane Conversion to C2 Hydrocarbons in Solid State Oxide Electrolyte Membrane Reactor

    Institute of Scientific and Technical Information of China (English)

    LI Jun; ZHAO Ling; ZHU Zhong-nan; XI Dan-li

    2005-01-01

    Provskite-type catalysts, Ln0.6 Sr0.4 FexCo1-x O3 (Ln = Nd,Pr, Gd, Sm, La, 0<x<1) and Ln0.8Na0.2CoO3(Ln= La,Gd, Sm) were synthesized, their catalytic properties in the oxidative coupling of methane (OCM) were examined in a fixed-bed reactor. The former group presented higher activity in the OCM, but the main product was carbon dioxide. While the later group showed lower activity but much higher selectivity to C2 hydrocarbons compared with the former. Electrochemical measurements were conducted in a solid oxide membrane reactor with La0.8 Na0.2CoO3 as catalyst. The results showed that methane was oxidized to carbon dioxide and ethane by two parallel reactions. Ethane was oxidized to ethene and carbon dioxide. A fraction of ethene was oxidized deeply to carbon dioxide. The total selectivity to C2 hydrocarbons exceeded 70%. Based on the experimental results, a kinetic model was suggested to describe the reaction results.

  18. Reactor Safety Research: Semiannual report, January-June 1986: Reactor Safety Research Program

    Energy Technology Data Exchange (ETDEWEB)

    1987-05-01

    Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the technology base supporting licensing decisions.

  19. Methane conversion using a dielectric barrier discharge reactor at atmospheric pressure for hydrogen production

    Science.gov (United States)

    Khadir, N.; Khodja, K.; Belasri, A.

    2017-09-01

    In the present paper, we carried out a theoretical study of dielectric barrier discharge (DBD) filled with pure methane gas. The homogeneous discharge model used in this work includes a plasma chemistry unit, an electrical circuit, and the Boltzmann equation. The model was applied to the case of a sinusoidal voltage at a period frequency of 50 kHz and under a gas pressure of 600 Torr. We investigated the temporal variation of electrical and kinetic discharge parameters such as plasma and dielectric voltages, the discharge current density, electric field, deposited power density, and the species concentration. We also checked the physical model validity by comparing its results with experimental work. According to the results discussed herein, the dielectric capacitance is the parameter that has the greatest effect on the methane conversion and H2/CH4 ratio. This work enriches the knowledge for the improvement of DBD for CH4 conversion and hydrogen production.

  20. 2011 Biomass Program Platform Peer Review. Thermochemical Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Grabowski, Paul E. [Office of Energy Efficiency and Renewable Energy (EERE), Washington, DC (United States)

    2012-02-01

    This document summarizes the recommendations and evaluations provided by an independent external panel of experts at the 2011 U.S. Department of Energy Biomass Program’s Thermochemical Conversion Platform Review meeting.

  1. 2011 Biomass Program Platform Peer Review: Biochemical Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Pezzullo, Leslie [Office of Energy Efficiency and Renewable Energy (EERE), Washington, DC (United States)

    2012-02-01

    This document summarizes the recommendations and evaluations provided by an independent external panel of experts at the 2011 U.S. Department of Energy Biomass Program’s Biochemical Conversion Platform Review meeting.

  2. Reactors for Catalytic Methanation in the Conversion of Biomass to Synthetic Natural Gas (SNG).

    Science.gov (United States)

    Schildhauer, Tilman J; Biollaz, Serge M A

    2015-01-01

    Production of Synthetic Natural Gas (SNG) from biomass is an important step to decouple the use of bioenergy from the biomass production with respect to both time and place. While anaerobic digestion of wet biomass is a state-of-the art process, wood gasification to producer gas followed by gas cleaning and methanation has only just entered the demonstration scale. Power-to-Gas applications using biogas from biomass fermentation or producer gas from wood gasification as carbon oxide source are under development. Due to the importance of the (catalytic) methanation step in the production of SNG from dry biomass or within Power-to-Gas applications, the specific challenges of this step and the developed reactor types are discussed in this review.

  3. Advanced reactors transition fiscal year 1995 multi-year program plan WBS 7.3

    Energy Technology Data Exchange (ETDEWEB)

    Loika, E.F.

    1994-09-22

    This document describes in detail the work to be accomplished in FY-1995 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 1995 Multi-Year Program Plan, DOE will provide authorization to perform the work outlined in the FY 1995 MYPP. Following direction given by the US Department of Energy (DOE) on December 15, 1993, Advanced Reactors Transition (ART), previously known as Advanced Reactors, will provide the planning and perform the necessary activities for placing the Fast Flux Test Facility (FFTF) in a radiologically and industrially safe shutdown condition. The DOE goal is to accomplish the shutdown in approximately five years. The Advanced Reactors Transition Multi-Year Program Plan, and the supporting documents; i.e., the FFTF Shutdown Program Plan and the FFTF Shutdown Project Resource Loaded Schedule (RLS), are defined for the life of the Program. During the transition period to achieve the Shutdown end-state, the facilities and systems will continue to be maintained in a safe and environmentally sound condition. Additionally, facilities that were associated with the Office of Nuclear Energy (NE) Programs, and are no longer required to support the Liquid Metal Reactor Program will be deactivated and transferred to an alternate sponsor or the Decontamination and Decommissioning (D and D) Program for final disposition, as appropriate.

  4. Analysis of LOCA Scenarios in the NIST Research Reactor Before and After Fuel Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Baek, J. S. [Brookhaven National Lab. (BNL), Upton, NY (United States); Cheng, L. Y. [Brookhaven National Lab. (BNL), Upton, NY (United States); Diamond, D. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2015-08-30

    An analysis has been done of hypothetical loss-of-coolant-accidents (LOCAs) in the research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The purpose of the analysis is to determine if the peak clad temperature remains below the Safety Limit, which is the blister temperature for the fuel. The configuration of the NBSR considered in the analysis is that projected for the future when changes will be made so that shutdown pumps do not operate when a LOCA signal is detected. The analysis was done for the present core with high-enriched uranium (HEU) fuel and with the proposed low-enriched uranium (LEU) fuel that would be used when the NBSR is converted from one to the other. The analysis consists of two parts. The first examines how the water would drain from the primary system following a break and the possibility for the loss of coolant from within the fuel element flow channels. This work is performed using the TRACE system thermal-hydraulic code. The second looks at the fuel clad temperature as a function of time given that the water may have drained from many of the flow channels and the water in the vessel is in a quasi-equilibrium state. The temperature behavior is investigated using the three-dimensional heat conduction code HEATING7.3. The results in all scenarios considered for both HEU and LEU fuel show that the peak clad temperature remains below the blister temperature.

  5. Plutonium Consumption Program, CANDU Reactor Project final report

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-31

    DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro`s Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel.

  6. U.S. Department of Energy Program of International Technical Cooperation for Research Reactor Utilization

    Energy Technology Data Exchange (ETDEWEB)

    Chong, D.; Manning, M.; Ellis, R.; Apt, K.; Flaim, S.; Sylvester, K.

    2004-10-03

    The U.S. Department of Energy, National Nuclear Security Administration (DOE/NNSA) has initiated collaborations with the national nuclear authorities of Egypt, Peru, and Romania for the purpose of advancing the commercial potential and utilization of their respective research reactors. Under its Office of International Safeguards ''Sister Laboratory'' program, DOE/NNSA has undertaken numerous technical collaborations over the past decade intended to promote peaceful applications of nuclear technology. Among these has been technical assistance in research reactor applications, such as neutron activation analysis, nuclear analysis, reactor physics, and medical radioisotope production. The current collaborations are intended to provide the subject countries with a methodology for greater commercialization of research reactor products and services. Our primary goal is the transfer of knowledge, both in administrative and technical issues, needed for the establishment of an effective business plan and utilization strategy for the continued operation of the countries' research reactors. Technical consultation, cooperation, and the information transfer provided are related to: identification, evaluation, and assessment of current research reactor capabilities for products and services; identification of opportunities for technical upgrades for new or expanded products and services; advice and consultation on research reactor upgrades and technical modifications; characterization of markets for reactor products and services; identification of competition and estimation of potential for market penetration; integration of technical constraints; estimation of cash flow streams; and case studies.

  7. Materials technology for an advanced space power nuclear reactor concept: Program summary

    Science.gov (United States)

    Gluyas, R. E.; Watson, G. K.

    1975-01-01

    The results of a materials technology program for a long-life (50,000 hr), high-temperature (950 C coolant outlet), lithium-cooled, nuclear space power reactor concept are reviewed and discussed. Fabrication methods and compatibility and property data were developed for candidate materials for fuel pins and, to a lesser extent, for potential control systems, reflectors, reactor vessel and piping, and other reactor structural materials. The effects of selected materials variables on fuel pin irradiation performance were determined. The most promising materials for fuel pins were found to be 85 percent dense uranium mononitride (UN) fuel clad with tungsten-lined T-111 (Ta-8W-2Hf).

  8. Reactor

    Science.gov (United States)

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  9. A computer program to determine the specific power of prismatic-core reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dobranich, D.

    1987-05-01

    A computer program has been developed to determine the maximum specific power for prismatic-core reactors as a function of maximum allowable fuel temperature, core pressure drop, and coolant velocity. The prismatic-core reactors consist of hexagonally shaped fuel elements grouped together to form a cylindrically shaped core. A gas coolant flows axially through circular channels within the elements, and the fuel is dispersed within the solid element material either as a composite or in the form of coated pellets. Different coolant, fuel, coating, and element materials can be selected to represent different prismatic-core concepts. The computer program allows the user to divide the core into any arbitrary number of axial levels to account for different axial power shapes. An option in the program allows the automatic determination of the core height that results in the maximum specific power. The results of parametric specific power calculations using this program are presented for various reactor concepts.

  10. Circulating and plateout activity program for gas-cooled reactors with arbitrary radioactive chains

    Energy Technology Data Exchange (ETDEWEB)

    Apperson, C.E. Jr.

    1978-03-01

    A time-dependent method for estimating the fuel body, circulating, plateout, and filter inventory of a high temperature gas-cooled reactor (HTGR) during normal operation is discussed. The primary coolant model accounts for the source, buildup, decay, and cleanup of isotopes that are gas borne inside the prestressed concrete reactor vessel (PCRV). This method has been implemented in the SUVIUS computer program that is described in detail.

  11. Reactors: A data-oriented synchronous/asynchronous programming model for distributed applications

    DEFF Research Database (Denmark)

    Field, John; Marinescu, Maria-Cristina; Stefansen, Christian Oskar Erik

    2009-01-01

    Our aim is to define the kernel of a simple and uniform programming model–the reactor model–which can serve as a foundation for building and evolving internet-scale programs. Such programs are characterized by collections of loosely-coupled distributed components that are assembled on the fly...... to produce a composite application. A reactor consists of two principal components: mutable state, in the form of a fixed collection of relations, and code, in the form of a fixed collection of rules in the style of Datalog. A reactor’s code is executed in response to an external stimulus, which takes...

  12. Technical support to the Nuclear Regulatory Commission for the boiling water reactor blowdown heat transfer program

    Energy Technology Data Exchange (ETDEWEB)

    Rice, R.E.

    1976-09-01

    Results are presented of studies conducted by Aerojet Nuclear Company (ANC) in FY 1975 to support the Nuclear Regulatory Commission (NRC) on the boiling water reactor blowdown heat transfer (BWR-BDHT) program. The support provided by ANC is that of an independent assessor of the program to ensure that the data obtained are adequate for verification of analytical models used for predicting reactor response to a postulated loss-of-coolant accident. The support included reviews of program plans, objectives, measurements, and actual data. Additional activity included analysis of experimental system performance and evaluation of the RELAP4 computer code as applied to the experiments.

  13. Advanced Reactors Transition program fiscal year 1998 multi-year work plan

    Energy Technology Data Exchange (ETDEWEB)

    Gantt, D.A.

    1997-09-25

    The mission of the Advanced Reactors Transition program is two-fold. First, the program is to maintain the Fast Flux Test Facility (FFTF) and the Fuels and Materials Examination Facility (FMEF) in Standby to support a possible future role in the tritium production strategy. Secondly, the program is to continue deactivation activities which do not conflict with the Standby directive. On-going deactivation activities include the processing of non-usable, irradiated, FFTF components for storage or disposal; deactivation of Nuclear Energy legacy test facilities; and deactivation of the Plutonium Recycle Test Reactor (PRTR) facility, 309 Building.

  14. Maintenance for power conversion system of gas turbine high temperature reactor (GTHTR300). Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Kosugiyama, Shinichi; Takada, Shoji; Katanishi, Shoji; Yan, Xing; Takizuka, Takakazu; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2002-11-01

    In order to be a suitable next generation nuclear power plant from reliable and economical points of view, it is necessary for GTHTR300 to have good maintenability and inspectability. Appropriate maintenance concept for the power conversion system of GTHTR300 consisting of a gas turbine, a compressor, a generator, a recuperator, a precooler and so on was studied based on results of the basic design of GTHTR300 in fiscal 2001. Considering degradation phenomena which could occur on each objective equipment, it is technically possible to reduce several maintenance items and extend maintenance interval for some equipment compared to those for existing LWR power plants and combined cycle fossil power plants. But owing to structural feature and installed location of each equipment, and fission product plate-out on each equipment, it became clear that some problems must be solved for making the maintenance works realistic and efficient. Solving the problems and confirming appropriateness of the proposed maintenance concept and plans will be done in coming detailing work of GTHTR300 design. (author)

  15. Capture programs, analysis, data graphication for the study of the thermometry of the TRIGA Mark III reactor core; Programas de captura, analisis y graficado de datos para el estudio de la termometria del nucleo del reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Paredes G, L.C

    1991-05-15

    This document covers the explanation of the capture programs, analysis and graphs of the data obtained during the measurement of the temperatures of the instrumented fuel element of the TRIGA Mark III reactor and of the coolant one near to this fuel, using the conversion card from Analogic to Digital of 'Data Translation', and using a signal conditioner for five temperature measurers with the help of thermo par type K, developed by the Simulation and Control of the nuclear systems management department, which gives a signal from 0 to 10 Vcd for an interval of temperature of 0 to 1000 C. (Author)

  16. Wind energy conversion system analysis model (WECSAM) computer program documentation

    Energy Technology Data Exchange (ETDEWEB)

    Downey, W T; Hendrick, P L

    1982-07-01

    Described is a computer-based wind energy conversion system analysis model (WECSAM) developed to predict the technical and economic performance of wind energy conversion systems (WECS). The model is written in CDC FORTRAN V. The version described accesses a data base containing wind resource data, application loads, WECS performance characteristics, utility rates, state taxes, and state subsidies for a six state region (Minnesota, Michigan, Wisconsin, Illinois, Ohio, and Indiana). The model is designed for analysis at the county level. The computer model includes a technical performance module and an economic evaluation module. The modules can be run separately or together. The model can be run for any single user-selected county within the region or looped automatically through all counties within the region. In addition, the model has a restart capability that allows the user to modify any data-base value written to a scratch file prior to the technical or economic evaluation. Thus, any user-supplied data for WECS performance, application load, utility rates, or wind resource may be entered into the scratch file to override the default data-base value. After the model and the inputs required from the user and derived from the data base are described, the model output and the various output options that can be exercised by the user are detailed. The general operation is set forth and suggestions are made for efficient modes of operation. Sample listings of various input, output, and data-base files are appended. (LEW)

  17. FFTF and Advanced Reactors Transition Program Resource Loaded Schedule

    Energy Technology Data Exchange (ETDEWEB)

    GANTT, D.A.

    2000-10-31

    This Resource Load Schedule (RLS) addresses two missions. The Advanced Reactors Transition (ART) mission, funded by DOE-EM, is to transition assigned, surplus facilities to a safe and compliant, low-cost, stable, deactivated condition (requiring minimal surveillance and maintenance) pending eventual reuse or D&D. Facilities to be transitioned include the 309 Building Plutonium Recycle Test Reactor (PRTR) and Nuclear Energy Legacy facilities. This mission is funded through the Environmental Management (EM) Project Baseline Summary (PBS) RL-TP11, ''Advanced Reactors Transition.'' The second mission, the Fast Flux Test Facility (FFTF) Project, is funded through budget requests submitted to the Office of Nuclear Energy, Science and Technology (DOE-NE). The FFTF Project mission is maintaining the FFTF, the Fuels and Materials Examination Facility (FMEF), and affiliated 400 Area buildings in a safe and compliant standby condition. This mission is to preserve the condition of the plant hardware, software, and personnel in a manner not to preclude a plant restart. This revision of the Resource Loaded Schedule (RLS) is based upon the technical scope in the latest revision of the following project and management plans: Fast Flux Test Facility Standby Plan (Reference 1); Hanford Site Sodium Management Plan (Reference 2); and 309 Building Transition Plan (Reference 4). The technical scope, cost, and schedule baseline is also in agreement with the concurrent revision to the ART Fiscal Year (FY) 2001 Multi-Year Work Plan (MYWP), which is available in an electronic version (only) on the Hanford Local Area Network, within the ''Hanford Data Integrator (HANDI)'' application.

  18. Investigation of the Performance of D2O-Cooled High-Conversion Reactors for Fuel Cycle Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Hikaru Hiruta; Gilles Youinou

    2013-09-01

    Th-U fueled cores are identified by evaluating the sensitivity coefficients of keff, mass balance, and void coefficient. The effect of advanced iron alloy cladding (i.e., FeCrAl) on the performance of Pu conversion in MOX fueled cores is studied instead of using standard stainless-steel cladding. Variations in clad thickness and coolant-to-fuel volume ratio are also exercised. The use of FeCrAl instead of SS as a cladding alloy reduces the required Pu enrichment and improves the Pu conversion rate primarily due to the absence of nickel in the cladding alloy that results in the reduction of the neutron absorption. Also the difference in void coefficients between SS and FeCrAl alloys is nearly 500 pcm over the entire burnup range. The report also shows sensitivity and uncertainty analyses in order to characterize D2O cooled HCPWRs from different aspects. The uncertainties of integral parameters (keff and void coefficient) for selected reactor cores are evaluated at different burnup points in order to find similarities and trends respect to D2O-HCPWR.

  19. Interactive programs with preschool children bring smiles and conversation to older adults: time-sampling study

    Science.gov (United States)

    2013-01-01

    Background Keeping older adults healthy and active is an emerging challenge of an aging society. Despite the importance of personal relationships to their health and well-being, changes in family structure have resulted in a lower frequency of intergenerational interactions. Limited studies have been conducted to compare different interaction style of intergenerational interaction. The present study aimed to compare the changes in visual attention, facial expression, engagement/behaviour, and intergenerational conversation in older adults brought about by a performance-based intergenerational (IG) program and a social-oriented IG program to determine a desirable interaction style for older adults. Methods The subjects of this study were 25 older adults who participated in intergenerational programs with preschool children aged 5 to 6 years at an adult day care centre in Tokyo. We used time sampling to perform a structured observation study. The 25 older participants of intergenerational programs were divided into two groups based on their interaction style: performance-based IG program (children sing songs and dance) and social-oriented IG program (older adults and children play games together). Based on the 5-minute video observation, we compared changes in visual attention, facial expression, engagement/behaviour, and intergenerational conversation between the performance-based and social-oriented IG programs. Results Constructive behaviour and intergenerational conversation were significantly higher in the social-oriented IG programming group than the performance-based IG programming group (pprogramming group than the performance-based IG programming (pprogramming group than the social-oriented IG programming group (pIntergenerational programs with preschool children brought smiles and conversation to older adults. The social-oriented IG program allowed older adults to play more roles than the performance-based IG program. The intergenerational programs provide

  20. Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced cha...

  1. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1982

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.; Sanders, J.P.

    1983-06-01

    During 1982 the High-Temperature Gas-Cooled Reactor (HTGR) Technology Program at Oak Ridge National Laboratory (ORNL) continued to develop experimental data required for the design and licensing of cogeneration HTGRs. The program involves fuels and materials development (including metals, graphite, ceramic, and concrete materials), HTGR chemistry studies, structural component development and testing, reactor physics and shielding studies, performance testing of the reactor core support structure, and HTGR application and evaluation studies.

  2. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1982

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.; Sanders, J.P.

    1983-06-01

    During 1982 the High-Temperature Gas-Cooled Reactor (HTGR) Technology Program at Oak Ridge National Laboratory (ORNL) continued to develop experimental data required for the design and licensing of cogeneration HTGRs. The program involves fuels and materials development (including metals, graphite, ceramic, and concrete materials), HTGR chemistry studies, structural component development and testing, reactor physics and shielding studies, performance testing of the reactor core support structure, and HTGR application and evaluation studies.

  3. Development status and potential program for development of proliferation-resistant molten-salt reactors

    Energy Technology Data Exchange (ETDEWEB)

    Engel, J.R.; Bauman, H.F.; Dearing, J.F.; Grimes, W.R.; McCoy, H.E. Jr.

    1979-03-01

    Preliminary studies of existing and conceptual molten-salt reactor (MSR) designs have led to the identification of conceptual systems that are technologically attractive when operated with denatured uranium as the principal fissile fuel. These denatured MSRs would also have favorable resource-utilization characteristics and substantial resistance to proliferation of weapons-usable nuclear materials. The report presents a summary of the current status of technology and a discussion of the major technical areas of a possible base program to develop commercial denatured MSRs. The general areas treated are (1) reactor design and development, (2) safety and safety related technology, (3) fuel-coolant behavior and fuel processing, and (4) reactor materials. A substantial development effort could lead to authorization for construction of a molten-salt test reactor about 5 years after the start of the program and operation of the unit about 10 years later. A prototype commercial denatured MSR could be expected to begin operating 25 years from the start of the program. The postulated base program would extend over 32 years and would cost about $700 million (1978 dollars, unescalated). Additional costs to construct the MSTR, $600 million, and the prototype commercial plant, $1470 million, would bring the total program cost to about $2.8 billion. Additional allowances probably should be made to cover contingencies and incidental technology areas not explicitly treated in this preliminary review.

  4. The effects of China’s Sloping Land Conversion Program on agricultural households

    DEFF Research Database (Denmark)

    Liu, Zhen; Henningsen, Arne

    In the late 1990s, China aimed to mitigate environmental degradation from agricultural production activities by introducing the world’s largest ’Payments for Environmental Services’ (PES) program ― the Sloping Land Conversion Program (SLCP). In order to analyze its effects on agricultural...... compensation payment rates has had some notable, but generally small effects....

  5. Annual report in compliance with the reactor sharing program, September 1, 1994--August 31, 1995

    Energy Technology Data Exchange (ETDEWEB)

    Karam, R.A.

    1997-04-01

    This report contains information with regard to facilities utilization, descriptions (brief), personnel, organization, and programs of the Neely Nuclear Research Center (NNRC) at the Georgia Institute of Technology. The NNRC has two major facilities: the Georgia Tech Research Reactor and the Hot Cell Laboratory. This report of NNRC utilization is prepared in compliance with the contract requirements between the U.S. Department of Energy and the Georgia Institute of Technology. The NNRC is a participant in the University Reactor Sharing Program; as such, it makes available its 5 MW research reactor, its Co-60 irradiation facility and its activation analysis laboratory to large numbers of students and faculty from many universities and colleges.

  6. TRAC-PF1: an advanced best-estimate computer program for pressurized water reactor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Liles, D.R.; Mahaffy, J.H.

    1984-02-01

    The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light water reactors. The TRAC-PF1 program provides this capability for pressurized water reactors and for many thermal-hydraulic experimental facilities. The code features either a one-dimensional or a three-dimensional treatment of the pressure vessel and its associated internals; a two-phase, two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field; flow-regime-dependent constitutive equation treatment; optional reflood tracking capability for both bottom flood and falling-film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. This report describes the thermal-hydraulic models and the numerical solution methods used in the code. Detailed programming and user information also are provided.

  7. Athabasca University: Conversion from Traditional Distance Education to Online Courses, Programs and Services

    Directory of Open Access Journals (Sweden)

    Alan Davis

    2001-01-01

    Full Text Available In its 30 years of operation, Athabasca University has witnessed the full impact of the growth of online distance education. Its conversion from mixed media course production and telephone/ mail tutoring to a variety of electronic information and communication technologies has been heterogeneous across disciplines and programs. Undergraduate programs in business, computing, and some social science programs have largely led the conversion, and all graduate programs have, since their inception, employed various features of online delivery. The parallel conversion of student services has been equally important to the effectiveness of these processes. The implications of this approach for the quality of offerings, support systems, costing, and the primary mandate of the University (which is to remove barriers, not create them are discussed.

  8. Gas reactor international cooperative program. HTR-synfuel application assessment

    Energy Technology Data Exchange (ETDEWEB)

    1979-09-01

    This study assesses the technical, environmental and economic factors affecting the application of the High Temperature Gas-Cooled Thermal Reactor (HTR) to: synthetic fuel production; and displacement of fossil fuels in other industrial and chemical processes. Synthetic fuel application considered include coal gasification, direct coal liquefaction, oil shale processing, and the upgrading of syncrude to motor fuel. A wide range of other industrial heat applications was also considered, with emphasis on the use of the closed-loop thermochemical energy pipeline to supply heat to dispersed industrial users. In this application syngas (H/sub 2/ +CO/sub 2/) is produced at the central station HTR by steam reforming and the gas is piped to individual methanators where typically 1000/sup 0/F steam is generated at the industrial user sites. The products of methanation (CH/sub 4/ + H/sub 2/O) are piped back to the reformer at the central station HTR.

  9. The BASIC Instructional Program Student Manual: Mainsail Conversion,

    Science.gov (United States)

    1978-04-01

    34 that teaches elementary programming in the BASIC language. This manual is the student’s main source of information about the BIP system and the BASIC...of different situations, in that they allow the pro- grammer to write the sequence only once and yet have it accessible from many different parts of

  10. Impact of the High Flux Isotope Reactor HEU to LEU Fuel Conversion on Cold Source Nuclear Heat Generation Rates

    Energy Technology Data Exchange (ETDEWEB)

    Chandler, David [ORNL

    2014-03-01

    Under the sponsorship of the US Department of Energy National Nuclear Security Administration, staff members at the Oak Ridge National Laboratory have been conducting studies to determine whether the High Flux Isotope Reactor (HFIR) can be converted from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. As part of these ongoing studies, an assessment of the impact that the HEU to LEU fuel conversion has on the nuclear heat generation rates in regions of the HFIR cold source system and its moderator vessel was performed and is documented in this report. Silicon production rates in the cold source aluminum regions and few-group neutron fluxes in the cold source moderator were also estimated. Neutronics calculations were performed with the Monte Carlo N-Particle code to determine the nuclear heat generation rates in regions of the HFIR cold source and its vessel for the HEU core operating at a full reactor power (FP) of 85 MW(t) and the reference LEU core operating at an FP of 100 MW(t). Calculations were performed with beginning-of-cycle (BOC) and end-of-cycle (EOC) conditions to bound typical irradiation conditions. Average specific BOC heat generation rates of 12.76 and 12.92 W/g, respectively, were calculated for the hemispherical region of the cold source liquid hydrogen (LH2) for the HEU and LEU cores, and EOC heat generation rates of 13.25 and 12.86 W/g, respectively, were calculated for the HEU and LEU cores. Thus, the greatest heat generation rates were calculated for the EOC HEU core, and it is concluded that the conversion from HEU to LEU fuel and the resulting increase of FP from 85 MW to 100 MW will not impact the ability of the heat removal equipment to remove the heat deposited in the cold source system. Silicon production rates in the cold source aluminum regions are estimated to be about 12.0% greater at BOC and 2.7% greater at EOC for the LEU core in comparison to the HEU core. Silicon is aluminum s major transmutation product and

  11. Gas-cooled fast reactor program. Progress report, January 1, 1980-June 30, 1981

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.

    1981-09-01

    Since the national Gas-Cooled Fast Breeder Reactor Program has been terminated, this document is the last progress report until reinstatement. It is divided into three sections: Core Flow Test Loop, GCFR shielding and physics, and GCFR pressure vessel and closure studies. (DLC)

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

    Energy Technology Data Exchange (ETDEWEB)

    Hulvey, R.K.

    1995-09-21

    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

  13. 10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

    Science.gov (United States)

    2010-01-01

    ... within the physical constraints of the system, the neutron spectrum, temperature history, and maximum... light water nuclear power reactors which result from exposure of these materials to neutron irradiation and the thermal environment. Under the program, fracture toughness test data are obtained...

  14. Determination of the Arrhenius Activation Energy Using a Temperature-Programmed Flow Reactor.

    Science.gov (United States)

    Chan, Kit-ha C.; Tse, R. S.

    1984-01-01

    Describes a novel method for the determination of the Arrhenius activation energy, without prejudging the validity of the Arrhenius equation or the concept of activation energy. The method involves use of a temperature-programed flow reactor connected to a concentration detector. (JN)

  15. Participation in the United States Department of Energy Reactor Sharing Program

    Energy Technology Data Exchange (ETDEWEB)

    Mulder, R.U.; Benneche, P.E.; Hosticka, B.

    1992-05-01

    The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics (to become the Department of Mechanical, Aerospace and Nuclear Engineering on July 1, 1992). As such, it is effectively used to support educational programs in engineering and science at the University of Virginia as well as those at other area colleges and universities. The expansion of support to educational programs in the mid-east region is a major objective. To assist in meeting this objective, the University of Virginia has been supported under the US Department of Energy (DOE) Reactor Sharing Program since 1978. Due to the success of the program, this proposal requests continued DOE support through August 1993.

  16. Review and analysis of the 1980-1989 biomass thermochemical conversion program

    Energy Technology Data Exchange (ETDEWEB)

    Stevens, D.J.

    1994-09-01

    In the period between 1980 and 1989, the U.S. Department of Energy (DOE) sponsored research and development projects through its Biomass Thermochemical Conversion (BTC) Program. Thermochemical conversion technologies use elevated temperatures to convert biomass into more useful forms of energy such as fuel gases or transportation fuels. The BTC Program included a wide range of biomass conversion projects in the areas of gasification, pyrolysis, liquefaction, and combustion. This work formed the basis of the present DOE research and development efforts on advanced liquid fuel and power generation systems. At the beginning of Fiscal Year 1989, the management of the BTC Program was transferred from Pacific Northwest Laboratory (PNL) to National Renewable Energy Laboratory (NREL, formerly Solar Energy Research Institute). This document presents a summary of the research which was performed under the BTC Program during the 1981-1989 time frame. The document consists of an analysis of the research projects which were funded by the BTC Program and a bibliography of published documents. This work will help ensure that information from PNL`s BTC Program is available to those interested in biomass conversion technologies. The background of the BTC Program is discussed in the first chapter of this report. In addition, a brief summary of other related biomass research and development programs funded by the U.S. Department of Energy and others is presented with references where additional information can be found. The remaining chapters of the report present a detailed summary of the research projects which were funded by the BTC Program. The progress which was made on each project is summarized, the overall impact on biomass conversion is discussed, and selected references are provided.

  17. Fast reactor safety program. Progress report, January-March 1980

    Energy Technology Data Exchange (ETDEWEB)

    None

    1980-05-01

    The goal of the DOE LMFBR Safety Program is to provide a technology base fully responsive to safety considerations in the design, evaluation, licensing, and economic optimization of LMFBRs for electrical power generation. A strategy is presented that divides safety technology development into seven program elements, which have been used as the basis for the Work Breakdown Structure (WBS) for the Program. These elements include four lines of assurance (LOAs) involving core-related safety considerations, an element supporting non-core-related plant safety considerations, a safety R and D integration element, and an element for the development of test facilities and equipment to be used in Program experiments: LOA-1 (prevent accidents); LOA-2 (limit core damage); LOA-3 (maintain containment integrity); LOA-4 (attenuate radiological consequences); plant considerations; R and D integration; and facility development.

  18. SAFSIM: A computer program for engineering simulations of space reactor system performance

    Energy Technology Data Exchange (ETDEWEB)

    Dobranich, D.

    1992-07-01

    SAFSIM (System Analysis Flow SIMulator) is a FORTRAN computer program that provides engineering simulations of user-specified flow networks at the system level. It includes fluid mechanics, heat transfer, and reactor dynamics capabilities. SAFSIM provides sufficient versatility to allow the simulation of almost any flow system, from a backyard sprinkler system to a clustered nuclear reactor propulsion system. In addition to versatility, speed and robustness are primary goals of SAFSIM. The current capabilities of SAFSIM are summarized, and some illustrative example results are presented.

  19. Effect of the variation of the level of lactose conversion in an immobilized lactase reactor upon operating costs for the production of Baker's yeast from hydrolyzed permeate obtained from the ultrafiltration of cottage cheese whey

    Energy Technology Data Exchange (ETDEWEB)

    Scott, T.C.; Hill, C.G. Jr.; Amundson, C.H.

    1987-01-01

    Operating costs for the production of Baker's yeast from hydrolyzed permeate from the ultrafiltration of cottage cheese whey were calculated as a function of the level of lactose conversion in the immobilized lactase reactor. These costs were calculated for the case of 90% conversion of lactose in the reactor and compared to those which result when running the reactor at lower conversions with recycle of unreacted lactose. Total operating costs were estimated by combining individual operating costs for the immobilized enzyme reactor, costs associated with processing a lactose recycle stream, and energy costs associated with cooling the reactor feed stream and sterilizing the hydrolysate stream. It was determined that operating costs are minimized at about 9.9 cents per pound of lactose when the reactor is run at approximately 72% conversion. This represents a savings of 2.4 cents per pound of lactose over the case of a once-through 90% conversion of lactose in the reactor. 8 refs., 4 figs., 9 tabs.

  20. Army Gas-Cooled Reactor Systems Program. Operation of ML-1 reactor skid in GCRE: safety evaluation report

    Energy Technology Data Exchange (ETDEWEB)

    None

    1964-10-01

    The operation of the ML-1 reactor skid in the modified GCRE facility, utilizing the GCRE reactor coolant circulating and heat removal systems, is described. An evaluation of the safety considerations associated with this mode of operation indicates that the consequences of the maximum credible accident are less severe than those previously approved for operation of the ML-1 reactor at the ML-1 test site or for operation of the GCRE-I reactor in the GCRE facility.

  1. Overview of the US Department of Energy Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    K. A. McCarthy; D. L. Williams; R. Reister

    2012-05-01

    The US Department of Energy Light Water Reactor Sustainability Program is focused on the long-term operation of US commercial power plants. It encompasses two facets of long-term operation: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the nuclear industry that support implementation of performance improvement technologies. An important aspect of the Light Water Reactor Sustainability Program is partnering with industry and the Nuclear Regulatory Commission to support and conduct the long-term research needed to inform major component refurbishment and replacement strategies, performance enhancements, plant license extensions, and age-related regulatory oversight decisions. The Department of Energy research, development, and demonstration role focuses on aging phenomena and issues that require long-term research and/or unique Department of Energy laboratory expertise and facilities and are applicable to all operating reactors. This paper gives an overview of the Department of Energy Light Water Reactor Sustainability Program, including vision, goals, and major deliverables.

  2. High-Power Microwave Transmission and Mode Conversion Program

    Energy Technology Data Exchange (ETDEWEB)

    Vernon, Ronald J. [Univ. of Wisconsin, Madison, WI (United States)

    2015-08-14

    This is a final technical report for a long term project to develop improved designs and design tools for the microwave hardware and components associated with the DOE Plasma Fusion Program. We have developed basic theory, software, fabrication techniques, and low-power measurement techniques for the design of microwave hardware associated gyrotrons, microwave mode converters and high-power microwave transmission lines. Specifically, in this report we discuss our work on designing quasi-optical mode converters for single and multiple frequencies, a new method for the analysis of perturbed-wall waveguide mode converters, perturbed-wall launcher design for TE0n mode gyrotrons, quasi-optical traveling-wave resonator design for high-power testing of microwave components, and possible improvements to the HSX microwave transmission line.

  3. SEPARATION OF HYDROGEN AND CARBON DIOXIDE USING A NOVEL MEMBRANE REACTOR IN ADVANCED FOSSIL ENERGY CONVERSION PROCESS

    Energy Technology Data Exchange (ETDEWEB)

    Shamsuddin Illias

    2002-06-10

    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 this new development. We designed and built a membrane reactor to study the reforming reaction. A two-dimensional pseudo-homogeneous reactor model was developed to study the performance of the membrane reactor parametrically. The important results are presented in this report.

  4. Updated Generation IV Reactors Integrated Materials Technology Program Plan, Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Halsey, William [Lawrence Livermore National Laboratory (LLNL); Hayner, George [Idaho National Laboratory (INL); Katoh, Yutai [ORNL; Klett, James William [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Stoller, Roger E [ORNL; Wilson, Dane F [ORNL

    2005-12-01

    The Department of Energy's (DOE's) Generation IV Nuclear Energy Systems Program will address the research and development (R&D) necessary to support next-generation nuclear energy systems. Such R&D will be guided by the technology roadmap developed for the Generation IV International Forum (GIF) over two years with the participation of over 100 experts from the GIF countries. The roadmap evaluated over 100 future systems proposed by researchers around the world. The scope of the R&D described in the roadmap covers the six most promising Generation IV systems. The effort ended in December 2002 with the issue of the final Generation IV Technology Roadmap [1.1]. The six most promising systems identified for next generation nuclear energy are described within the roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor - SCWR and the Very High Temperature Reactor - VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor - GFR, the Lead-cooled Fast Reactor - LFR, and the Sodium-cooled Fast Reactor - SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides, and may provide an alternative to accelerator-driven systems. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of the structural materials needed to ensure their safe and reliable operation. Accordingly, DOE has identified materials as one of the focus areas for Gen IV technology development.

  5. Participation in the US Department of Energy reactor sharing program. Final progress report, October 1996--September 1997

    Energy Technology Data Exchange (ETDEWEB)

    Mulder, R.U.; Benneche, P.E.; Hosticka, B.

    1998-04-01

    The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would not be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA from use at their institutions. These areas are discussed in this report.

  6. Participation in the US Department of Energy Reactor Sharing Program. Annual report, September 30, 1993--September 29, 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-03-01

    The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would not be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA for use at their institutions. These areas are discussed further in the report.

  7. University of Florida--US Department of Energy 1994-1995 reactor sharing program

    Energy Technology Data Exchange (ETDEWEB)

    Vernetson, W.G.

    1996-06-01

    The grant support of $24,250 (1994-95?) was well used by the University of Florida as host institution to support various educational institutions in the use of UFTR Reactor. All users and uses were screened to assure the usage was for educational institutions eligible for participation in the Reactor Sharing Program; where research activities were involved, care was taken to assure the research was not funded by grants for contract funding from outside sources. Over 12 years, the program has been a key catalyst for renewing utilization of UFTR both by external users around the State of Florida and the Southeast and by various faculty members within the University of Florida. Tables provide basic information about the 1994-95 program and utilization of UFTR.

  8. Biological CO2 conversion to acetate in subsurface coal-sand formation using a high-pressure reactor system

    Science.gov (United States)

    Ohtomo, Y.; Ijiri, A.; Ikegawa, Y.; Tsutsumi, M.; Imachi, H.; Uramoto, G.; Hoshino, T.; Morono, Y.; Tanikawa, W.; Hirose, T.; Inagaki, F.

    2013-12-01

    The geological CO2 sequestration into subsurface unmineable oil/gas fields and coal formations has been considered as one of the possible ways to reduce dispersal of anthropogenic greenhouse gasses into the atmosphere. However, feasibility of CO2 injection largely depends on a variety of geological and economical settings, and its ecological consequences have remained largely unpredictable. To address these issues, we developed a new flow-through-type CO2 injection system designated as the 'geobio-reactor system' to examine possible geophysical, geochemical and microbiological impact caused by CO2 injection under in-situ pressure (0-100 MPa) and temperature (0-70°C) conditions. In this study, we investigated Eocene bituminous coal-sandstones in the northwestern Pacific coast, Hokkaido, Japan, using the geobio-reactor system. Anaerobic artificial fluid and CO2 (flow rate: 0.002 and 0.00001 mL/min, respectively) were continuously supplemented into the coal-sand column under the pore pressure of 40 MPa (confined pressure: 41 MPa) at 40°C for 56 days. Molecular analysis of bacterial 16S rRNA genes showed that predominant bacterial components were physically dispersed from coal to sand as the intact form during experiment. Cultivation experiments from sub-sampling fluids indicated that some terrestrial microbes could preserve their survival in subsurface condition. Molecular analysis of archaeal 16S rRNA genes also showed that no methanogens were activated during experiment. We also anaerobically incubated the coal sample using conventional batch-type cultivation technique with a medium for methanogens. After one year of the batch incubation at 20°C, methane could be detected from the cultures except for the acetate-fed culture. The sequence of archaeal 16S rRNA genes via PCR amplification obtained from the H2 plus formate-fed culture was affiliated with a hydrogenotrophic methanogen within the genus Methanobacterium, whereas the methanol plus trimethylamine culture

  9. Aging management program of the reactor building concrete at Point Lepreau Generating Station

    Science.gov (United States)

    Aldea, C.-M.; Shenton, B.; Demerchant, M. M.; Gendron, T.

    2011-04-01

    In order for New Brunswick Power Nuclear (NBPN) to control the risks of degradation of the concrete reactor building at the Point Lepreau Generating Station (PLGS) the development of an aging management plan (AMP) was initiated. The intention of this plan was to determine the requirements for specific structural components of concrete of the reactor building that require regular inspection and maintenance to ensure the safe and reliable operation of the plant. The document is currently in draft form and presents an integrated methodology for the application of an AMP for the concrete of the reactor building. The current AMP addresses the reactor building structure and various components, such as joint sealant and liners that are integral to the structure. It does not include internal components housed within the structure. This paper provides background information regarding the document developed and the strategy developed to manage potential degradation of the concrete of the reactor building, as well as specific programs and preventive and corrective maintenance activities initiated.

  10. Aging management program of the reactor building concrete at Point Lepreau Generating Station

    Directory of Open Access Journals (Sweden)

    Gendron T.

    2011-04-01

    Full Text Available In order for New Brunswick Power Nuclear (NBPN to control the risks of degradation of the concrete reactor building at the Point Lepreau Generating Station (PLGS the development of an aging management plan (AMP was initiated. The intention of this plan was to determine the requirements for specific structural components of concrete of the reactor building that require regular inspection and maintenance to ensure the safe and reliable operation of the plant. The document is currently in draft form and presents an integrated methodology for the application of an AMP for the concrete of the reactor building. The current AMP addresses the reactor building structure and various components, such as joint sealant and liners that are integral to the structure. It does not include internal components housed within the structure. This paper provides background information regarding the document developed and the strategy developed to manage potential degradation of the concrete of the reactor building, as well as specific programs and preventive and corrective maintenance activities initiated.

  11. Coal conversion rate in 1t/d PSU liquefaction reactor; 1t/d PSU ekika hannoto ni okeru sekitan tenka sokudo no kento

    Energy Technology Data Exchange (ETDEWEB)

    Ikeda, K.; Imada, K. [Nippon Steel Corp., Tokyo (Japan); Nogami, Y.; Inokuchi, K. [Mitsui SRC Development Co. Ltd., Tokyo (Japan)

    1996-10-28

    To investigate the coal liquefaction characteristics, coal slurry samples were taken from the outlets of the reactors and slurry preheater of NEDOL process 1 t/d process supporting unit (PSU), and were analyzed. Tanito Harum coal was used for liquefaction, and the slurry was prepared with recycle solvent. Liquefaction was performed using synthetic iron sulfide catalyst at reaction temperatures, 450 and 465{degree}C. Solubility of various solid samples was examined against n-hexane, toluene, and tetrahydrofuran (THF). When considering the decrease of IMO (THF-insoluble and ash) as a characteristic of coal conversion reaction, around 20% at the outlet of the slurry preheater, around 70% within the first reactor, and several percents within the successive second and third reactors were converted against supplied coal. Increase of reaction temperature led to the increase of evaporation of oil fraction, which resulted in the decrease of actual slurry flow rate and in the increase of residence time. Thus, the conversion of coal was accelerated by the synergetic effect of temperature and time. Reaction rate constant of the coal liquefaction was around 2{times}10{sup -1} [min{sup -1}], which increased slightly with increasing the reaction temperature from 450 to 465{degree}C. 3 refs., 5 figs., 1 tab.

  12. Development of a conversion program to make SIMIND-generated SPECT data interfile acceptable to Xeleris.

    Science.gov (United States)

    Pandey, Anil K; Kumar, Jay P; Aheer, Deepak; Sharma, Anshul; Hasan, Binish; Patel, Chetan; Bal, Chandra S; Kumar, Rakesh

    2015-12-01

    The Monte Carlo code SIMIND is used in nuclear medicine for research purposes, and also for testing the validity of various applications. Conversion of a SIMIND-generated interfile (header and image data file) is required to process the simulated image data on a Xeleris workstation. Currently there is no conversion program provided with SIMIND to convert its interfile, which is acceptable in any nuclear medicine workstation. Manual editing of the header file is possible with any standard text editor but it is time-consuming (requiring ∼1 h) and stressful. To alleviate this, we have developed a conversion program using FreeMat V4.0 (an open source software similar to MATLAB from MathWorks). It takes an average of 0.04404 s to convert a SIMIND SPECT data interfile and make it acceptable to Xeleris.

  13. The current state of the Russian reduced enrichment research reactors program

    Energy Technology Data Exchange (ETDEWEB)

    Aden, V.G.; Kartashov, E.F.; Lukichev, V.A. [and others

    1997-08-01

    During the last year after the 16-th International Conference on Reducing Fuel Enrichment in Research Reactors held in October, 1993 in Oarai, Japan, the conclusive stage of the Program on reducing fuel enrichment (to 20% in U-235) in research reactors was finally made up in Russia. The Program was started late in 70th and the first stage of the Program was completed by 1986 which allowed to reduce fuel enrichment from 80-90% to 36%. The completion of the Program current stage, which is counted for 5-6 years, will exclude the use of the fuel enriched by more than 20% from RF to other countries such as: Poland, Czeck Republick, Hungary, Roumania, Bulgaria, Libya, Viet-Nam, North Korea, Egypt, Latvia, Ukraine, Uzbekistan and Kazakhstan. In 1994 the Program, approved by RF Minatom authorities, has received the status of an inter-branch program since it was admitted by the RF Ministry for Science and Technical Policy. The Head of RF Minatom central administrative division N.I.Ermakov was nominated as the Head of the Russian Program, V.G.Aden, RDIPE Deputy Director, was nominated as the scientific leader. The Program was submitted to the Commission for Scientific, Technical and Economical Cooperation between USA and Russia headed by Vice-President A. Gore and Prime Minister V. Chemomyrdin and was given support also.

  14. Payment for Environmental Services: The Sloping Land Conversion Program in Ningxia Autonomous Region of China

    NARCIS (Netherlands)

    Zhang, L.; Tu, Q.; Mol, A.P.J.

    2008-01-01

    China's Sloping Land Conversion Program has been implemented since 2002. It aims to achieve goals of ecological recovery and poverty alleviation, by retiring steeply sloping land from crop production and freeing surplus agricultural labor for off-farm activities. Given the huge investment that has b

  15. Initial Neutronics Analyses for HEU to LEU Fuel Conversion of the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Kontogeorgakos, D. [Argonne National Lab. (ANL), Argonne, IL (United States); Derstine, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Bauer, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Stevens, J. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2013-06-01

    The purpose of the TREAT reactor is to generate large transient neutron pulses in test samples without over-heating the core to simulate fuel assembly accident conditions. The power transients in the present HEU core are inherently self-limiting such that the core prevents itself from overheating even in the event of a reactivity insertion accident. The objective of this study was to support the assessment of the feasibility of the TREAT core conversion based on the present reactor performance metrics and the technical specifications of the HEU core. The LEU fuel assembly studied had the same overall design, materials (UO2 particles finely dispersed in graphite) and impurities content as the HEU fuel assembly. The Monte Carlo N–Particle code (MCNP) and the point kinetics code TREKIN were used in the analyses.

  16. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Science.gov (United States)

    2013-10-24

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors AGENCY... Cooling Systems for New Boiling-Water Reactors.'' This RG describes testing methods the NRC staff considers acceptable for demonstrating the operability of emergency core cooling systems (ECCSs) for boiling...

  17. U.S. Department of Energy University Reactor Instrumentation Program Final Report for 1992-94 Grant for the University of Florida Training Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vernetson, William G.

    1999-04-01

    Overall, the instrumentation obtained under the first year 1992-93 University Reactor Instrumentation Program grant assured that the goals of the program were well understood and met as well as possible at the level of support provided for the University of Florida Training Reactor facility. Though the initial grant support of $21,000 provided toward the purchase of $23,865 of proposed instrumentation certainly did not meet many of the facility's needs, the instrumentation items obtained and implemented did meet some critical needs and hence the goals of the Program to support modernization and improvement of reactor facilities such as the UFTR within the academic community. Similarly, the instrumentation obtained under the second year 1993-94 University Reactor Instrumentation Program grant again met some of the critical needs for instrumentation support at the UFTR facility. Again, though the grant support of $32,799 for proposed instrumentation at the same cost projection does not need all of the facility's needs, it does assure continued facility viability and improvement in operations. Certainly, reduction of forced unavailability of the reactor is the most obvious achievement of the University Reactor Instrumentation Program to date at the UFTR. Nevertheless, the ability to close out several expressed-inspection concerns of the Nuclear Regulatory Commission with acquisition of the low level survey meter and the area radiation monitoring system is also very important. Most importantly, with modest cost sharing the facility has been able to continue and even accelerate the improvement and modernization of a facility, especially in the Neutron Activation Analysis Laboratory, that is used by nearly every post-secondary school in the State of Florida and several in other states, by dozens of departments within the University of Florida, and by several dozen high schools around the State of Florida on a regular basis. Better, more reliable service to such

  18. International academic program in technologies of light-water nuclear reactors. Phases of development and implementation

    Science.gov (United States)

    Geraskin, N. I.; Glebov, V. B.

    2017-01-01

    The results of implementation of European educational projects CORONA and CORONA II dedicated to preserving and further developing nuclear knowledge and competencies in the area of technologies of light-water nuclear reactors are analyzed. Present article addresses issues of design and implementation of the program for specialized training in the branch of technologies of light-water nuclear reactors. The systematic approach has been used to construct the program for students of nuclear specialties, which corresponding to IAEA standards and commonly accepted nuclear principles recognized in the European Union. Possibilities of further development of the international cooperation between countries and educational institutions are analyzed. Special attention is paid to e-learning/distance training, nuclear knowledge preservation and interaction with European Nuclear Education Network.

  19. Light Water Reactor Sustainability Program: Digital Architecture Project Plan

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, Ken [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    There are many technologies available to the nuclear power industry to improve efficiency in plant work activities. These range from new control room technologies to those for mobile field workers. They can make a positive impact on a wide range of performance objectives – increase in productivity, human error reduction, validation of results, accurate transfer of data, and elimination of repetitive tasks. It is expected that the industry will more and more turn to these technologies to achieve these operational efficiencies to lower costs. At the same time, this will help utilities manage a looming staffing problem as the inevitable retirement wave of the more seasoned workers affects both staffing levels and knowledge retention. A barrier to this wide-scale implementation of new technologies for operational efficiency is the lack of a comprehensive digital architecture that can support the real-time information exchanges needed to achieve the desired operational efficiencies. This project will define an advanced digital architecture that will accommodate the entire range of system, process, and plant worker activity to enable the highest degree of integration, thereby creating maximum efficiency and productivity. This pilot project will consider a range of open standards that are suitable for the various data and communication requirements of a seamless digital environment. It will map these standards into an overall architecture to support the II&C developments of this research program.

  20. Thermal-hydraulic Fortran program for steady-state calculations of plate-type fuel research reactors

    Directory of Open Access Journals (Sweden)

    Khedr Ahmed

    2008-01-01

    Full Text Available The safety assessment of research and power reactors is a continuous process covering their lifespan and requiring verified and validated codes. Power reactor codes all over the world are well established and qualified against real measuring data and qualified experimental facilities. These codes are usually sophisticated, require special skills and consume a lot of running time. On the other hand, most research reactor codes still require much more data for validation and qualification. It is, therefore, of benefit to any regulatory body to develop its own codes for the review and assessment of research reactors. The present paper introduces a simple, one-dimensional Fortran program called THDSN for steady-state thermal-hydraulic calculations of plate-type fuel research reactors. Besides calculating the fuel and coolant temperature distributions and pressure gradients in an average and hot channel, the program calculates the safety limits and margins against the critical phenomena encountered in research reactors, such as the onset of nucleate boiling, critical heat flux and flow instability. Well known thermal-hydraulic correlations for calculating the safety parameters and several formulas for the heat transfer coefficient have been used. The THDSN program was verified by comparing its results for 2 and 10 MW benchmark reactors with those published in IAEA publications and a good agreement was found. Also, the results of the program are compared with those published for other programs, such as the PARET and TERMIC.

  1. REACTOR AND SHIELD PHYSICS. Comprehensive Technical Report, General Electric Direct-Air-Cycle, Aircraft Nuclear Propulsion Program.

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, W.E.; Simpson, J.D.

    1962-01-01

    This volume is one of twenty-one summarizing the Aircraft Nuclear Propulsion Program of the General Electric Company. This volume describes the experimental and theoretical work accomplished in the areas of reactor and shield physics.

  2. Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, July 1, 1979-September 30, 1979

    Energy Technology Data Exchange (ETDEWEB)

    1980-03-07

    The results of work performed from July 1, 1979 through September 30, 1979 on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. 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 status of the data management system is presented. In addition, the progress in the screening test program is described.

  3. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, January 1, 1980-March 31, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-25

    Results are presented of work performed on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Included are 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, including screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850, and 950/sup 0/C.

  4. Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, April 1, 1979-June 30, 1979

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-25

    The results are presented of work performed on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. 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 status of the data management system is presented. In addition, the progress in the screening test program is described.

  5. Strengthening the fission reactor nuclear science and engineering program at UCLA. Final technical report

    Energy Technology Data Exchange (ETDEWEB)

    Okrent, D.

    1997-06-23

    This is the final report on DOE Award No. DE-FG03-92ER75838 A000, a three year matching grant program with Pacific Gas and Electric Company (PG and E) to support strengthening of the fission reactor nuclear science and engineering program at UCLA. The program began on September 30, 1992. The program has enabled UCLA to use its strong existing background to train students in technological problems which simultaneously are of interest to the industry and of specific interest to PG and E. The program included undergraduate scholarships, graduate traineeships and distinguished lecturers. Four topics were selected for research the first year, with the benefit of active collaboration with personnel from PG and E. These topics remained the same during the second year of this program. During the third year, two topics ended with the departure o the students involved (reflux cooling in a PWR during a shutdown and erosion/corrosion of carbon steel piping). Two new topics (long-term risk and fuel relocation within the reactor vessel) were added; hence, the topics during the third year award were the following: reflux condensation and the effect of non-condensable gases; erosion/corrosion of carbon steel piping; use of artificial intelligence in severe accident diagnosis for PWRs (diagnosis of plant status during a PWR station blackout scenario); the influence on risk of organization and management quality; considerations of long term risk from the disposal of hazardous wastes; and a probabilistic treatment of fuel motion and fuel relocation within the reactor vessel during a severe core damage accident.

  6. Joint Assessment of ETRR-2 Research Reactor Operations Program, Capabilities, and Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Bissani, M; O' Kelly, D S

    2006-05-08

    A joint assessment meeting was conducted at the Egyptian Atomic Energy Agency (EAEA) followed by a tour of Egyptian Second Research Reactor (ETRR-2) on March 22 and 23, 2006. The purpose of the visit was to evaluate the capabilities of the new research reactor and its operations under Action Sheet 4 between the U.S. DOE and the EAEA, ''Research Reactor Operation'', and Action Sheet 6, ''Technical assistance in The Production of Radioisotopes''. Preliminary Recommendations of the joint assessment are as follows: (1) ETRR-2 utilization should be increased by encouraging frequent and sustained operations. This can be accomplished in part by (a) Improving the supply-chain management for fresh reactor fuel and alleviating the perception that the existing fuel inventory should be conserved due to unreliable fuel supply; and (b) Promulgating a policy for sample irradiation priority that encourages the use of the reactor and does not leave the decision of when to operate entirely at the discretion of reactor operations staff. (2) Each experimental facility in operation or built for a single purpose should be reevaluated to focus on those that most meet the goals of the EAEA strategic business plan. Temporary or long-term elimination of some experimental programs might be necessary to provide more focused utilization. There may be instances of emerging reactor applications for which no experimental facility is yet designed or envisioned. In some cases, an experimental facility may have a more beneficial use than the purpose for which it was originally designed. For example, (a) An effective Boron Neutron Capture Therapy (BNCT) program requires nearby high quality medical facilities. These facilities are not available and are unlikely to be constructed near the Inshas site. Further, the BNCT facility is not correctly designed for advanced research and therapy programs using epithermal neutrons. (b) The ETRR-2 is frequently operated to

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

    Energy Technology Data Exchange (ETDEWEB)

    Pytel, K.; Mieleszczenko, W.; Lechniak, J.; Moldysz, A.; Andrzejewski, K.; Kulikowska, T.; Marcinkowska, A.; Garner, P. L.; Hanan, N. A.; Nuclear Engineering Division; Institute of Atomic Energy (Poland)

    2010-03-01

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

  8. Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Van den Branden, G. [SCK CEN (Belgium); Kalcheva, S. [SCK CEN (Belgium); Sikik, E. [SCK CEN (Belgium); Koonen, E. [SCK CEN (Belgium)

    2015-12-01

    BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water (Figure 1). The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cm2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident.

  9. Final report for U.S. Department of Energy Grant DE-FG02-95NE38118-5 University Reactor Sharing Program [Purdue University

    Energy Technology Data Exchange (ETDEWEB)

    Bean, R.S.

    2001-06-01

    Under the Reactor Sharing Program, a total of 350 high school students participated in the neutron activation experiment and 484 high school and university students and members of the general public participated in reactor tours.

  10. Acetate conversion in anaerobic biogas reactors: Traditional and molecular tools for studying this important group of anaerobic microorganisms

    DEFF Research Database (Denmark)

    Schmidt, Jens Ejbye; Mladenovska, Zuzana; Lange, Marianne;

    2000-01-01

    Different methods were applied to study the role of aceticlastic methanogens in biogas reactors treating solid waste and wastewater. We used traditional microbiological methods, immunological and 16S rRNA ribosomal probes for detection of the methanogens. Using this approach we identified the met...

  11. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, September 23, 1976--December 31, 1976

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    This report presents the results of work performed from September 23, 1976 through December 31, 1976 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the affect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed.

  12. Recommended methods for evaluating the benefits of ECUT Program outputs. [Energy Conversion and Utilization

    Energy Technology Data Exchange (ETDEWEB)

    Levine, L.O.; Winter, C.

    1986-03-01

    This study was conducted to define and develop techniques that could be used to assess the complete spectrum of positive effects resulting from the Energy Conversion and Utilization Technologies (ECUT) Program activities. These techniques could then be applied to measure the benefits from past ECUT outputs. In addition, the impact of future ECUT outputs could be assessed as part of an ongoing monitoring process, after sufficient time has elapsed to allow their impacts to develop.

  13. BREEDER: a microcomputer program for financial analysis of a large-scale prototype breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Giese, R.F.

    1984-04-01

    This report describes a microcomputer-based, single-project financial analysis program: BREEDER. BREEDER is a user-friendly model designed to facilitate frequent and rapid analyses of the financial implications associated with alternative design and financing strategies for electric generating plants and large-scale prototype breeder (LSPB) reactors in particular. The model has proved to be a useful tool in establishing cost goals for LSPB reactors. The program is available on floppy disks for use on an IBM personal computer (or IBM look-a-like) running under PC-DOS or a Kaypro II transportable computer running under CP/M (and many other CP/M machines). The report documents version 1.5 of BREEDER and contains a user's guide. The report also includes a general overview of BREEDER, a summary of hardware requirements, a definition of all required program inputs, a description of all algorithms used in performing the construction-period and operation-period analyses, and a summary of all available reports. The appendixes contain a complete source-code listing, a cross-reference table, a sample interactive session, several sample runs, and additional documentation of the net-equity program option.

  14. Practicing End-of-Life Conversations: Physician Communication Training Program in Palliative Care.

    Science.gov (United States)

    Rucker, Bronwyn; Browning, David M

    2015-01-01

    A Physician Communication Training Program (PCTP) utilizing scripts based on actual family conferences with patients, families, and the health care team was developed at one medical center in the Northeast. The program was designed, adapted, and directed by a palliative care social worker. The primary goal of the program is to help residents and attending physicians build better communication skills in establishing goals of care and in end-of-life planning. The scripts focus on improving physicians' basic skills in conducting family meetings, discussing advance directives, prognosis, brain death, and withdrawal of life support. Excerpts from the scripts utilized in the program are included. Feedback from participants has been positive, with all respondents indicating improvement in their capacity to take part in these challenging conversations.

  15. Light Water Reactor Sustainability Program Status of Silicon Carbide Joining Technology Development

    Energy Technology Data Exchange (ETDEWEB)

    Shannon M. Bragg-Sitton

    2013-09-01

    Advanced, accident tolerant nuclear fuel systems are currently being investigated for potential application in currently operating light water reactors (LWR) or in reactors that have attained design certification. Evaluation of potential options for accident tolerant nuclear fuel systems point to the potential benefits of silicon carbide (SiC) relative to Zr-based alloys, including increased corrosion resistance, reduced oxidation and heat of oxidation, and reduced hydrogen generation under steam attack (off-normal conditions). If demonstrated to be applicable in the intended LWR environment, SiC could be used in nuclear fuel cladding or other in-core structural components. Achieving a SiC-SiC joint that resists corrosion with hot, flowing water, is stable under irradiation and retains hermeticity is a significant challenge. This report summarizes the current status of SiC-SiC joint development work supported by the Department of Energy Light Water Reactor Sustainability Program. Significant progress has been made toward SiC-SiC joint development for nuclear service, but additional development and testing work (including irradiation testing) is still required to present a candidate joint for use in nuclear fuel cladding.

  16. PR-EDB: Power Reactor Embrittlement Data Base, Version 2. Revision 2, Program description

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-01-01

    Investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes Standard Review Plans (SRP`s) and Guides for license renewal can be greatly expedited by the use of a well-designed computerized data base. Also, such a data base is essential for the validation of embrittlement prediction models by researchers. The Power Reactor Embrittlement Data Base (PR-EDB) is such a comprehensive collection of data for US commercial nuclear reactors. The current version of the PR-EDB contains the Charpy test data that were irradiated in 252 capsules of 96 reactors and consists of 207 data points for heat-affected-zone (HAZ) materials (98 different HAZ), 227 data points for weld materials (105 different welds), 524 data points for base materials (136 different base materials), including 297 plate data points (85 different plates), 119 forging data points (31) different forging), and 108 correlation monitor materials data points (3 different plates). The data files are given in dBASE format and can be accessed with any computer using the DOS operating system. ``User-friendly`` utility programs are used to retrieve and select specific data, manipulate data, display data to the screen or printer, and to fit and plot Charpy impact data. The results of several studies investigated are presented in Appendix D.

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

    Energy Technology Data Exchange (ETDEWEB)

    David Petti; Hans Gougar; Gary Bell

    2005-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    David Petti; Hans Gougar; Gary Bell

    2005-05-01

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

  19. Research and development program for PWR safety at the CEA reactor thermal hydraulics laboratories

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, M. [CEA, Grenoble (France)

    1995-04-15

    Since the start of the French electronuclear program, the three partners Fermate, EDF and Cea (DRN and IPSN) have devoted considerable effort to research and development for safety issues. In particular an important program on thermal hydraulics was initiated at the beginning of the seventies. It is illustrated by the development of the CATHARE thermalhydraulic safety code which includes physical models derived from a large experimental support program and the construction of the BETHSY integral facility which is aimed to assess both the CATHARE code and the physical relevance of the accident management procedures to be applied on reactors. The state of the art on this program is described with particular emphasis on the capabilities and the assessment of the last version of CATHARE and the lessons drawn from 50 BETHSY tests performed so far. The future plans for safety research cover the following strategy: - to solve the few problems identified on present computing tools and extend the assessment - to solve the few problems identified on present computing tools and extend the assessment - to perform safety studies on the basis of plant operation feedback - to contribute to treating the safety issues related to the future reactors and in particular the case of severe accidents which have to be taken into account from the design stage. The program on severe accidents is aimed to support the design studies performed by the industrial partners and to provide computing tools which model the various phases of severe accidents and will be validated on experiments performed with real and simulating materials. The main lines of the program are: - the development of the TOLBIAC 3D code for the thermal hydraulics of core melt pools, which will be validated against the Bali experiment presently under construction - the Sultan experiment, to study the capability of cooling by external flooding of the reactor vessel - the development of the MC-3D code for core melt

  20. Advanced reactors and novel reactions for the conversion of triglyceride based oils into high quality renewable transportation fuels

    Science.gov (United States)

    Linnen, Michael James

    Sustainable energy continues to grow more important to all societies, leading to the research and development of a variety of alternative and renewable energy technologies. Of these, renewable liquid transportation fuels may be the most visible to consumers, and this visibility is further magnified by the long-term trend of increasingly expensive petroleum fuels that the public consumes. While first-generation biofuels such as biodiesel and fuel ethanol have been integrated into the existing fuel infrastructures of several countries, the chemical differences between them and their petroleum counterparts reduce their effectiveness. This gives rise to the development and commercialization of second generation biofuels, many of which are intended to have equivalent properties to those of their petroleum counterparts. In this dissertation, the primary reactions for a second-generation biofuel process, known herein as the University of North Dakota noncatalytic cracking process (NCP), have been studied at the fundamental level and improved. The NCP is capable of producing renewable fuels and chemicals that are virtually the same as their petroleum counterparts in performance and quality (i.e., petroleum-equivalent). In addition, a novel analytical method, FIMSDIST was developed which, within certain limitations, can increase the elution capabilities of GC analysis and decrease sample processing times compared to other high resolution methods. These advances are particularly useful for studies of highly heterogeneous fuel and/or organic chemical intermediates, such as those studied for the NCP. However the data from FIMSDIST must be supplemented with data from other methods such as for certain carboxylic acid, to provide accurate, comprehensive results, From a series of TAG cracking experiments that were performed, it was found that coke formation during cracking is most likely the result of excessive temperature and/or residence time in a cracking reactor. Based on this

  1. Nuclear Systems Enhanced Performance Program, Maintenance Cycle Extension in Advanced Light Water Reactor Design

    Energy Technology Data Exchange (ETDEWEB)

    Professor Neill Todreas

    2001-10-01

    A renewed interest in new nuclear power generation in the US has spurred interest in developing advanced reactors with features which will address the public's concerns regarding nuclear generation. However, it is economic performance which will dictate whether any new orders for these plants will materialize. Economic performance is, to a great extent, improved by maximizing the time that the plant is on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Indeed, the strategy for the advanced light water reactor plant IRIS (International Reactor, Innovative and Secure) is to utilize an eight year operating cycle. This report describes a formalized strategy to address, during the design phase, the maintenance-related barriers to an extended operating cycle. The top-level objective of this investigation was to develop a methodology for injecting component and system maintainability issues into the reactor plant design process to overcome these barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the IRIS design. The first step in meeting the top-level objective was to determine the types of operating cycle length barriers that the IRIS design team is likely to face. Evaluation of previously identified regulatory and investment protection surveillance program barriers preventing a candidate operating PWR from achieving an extended (48 month) cycle was conducted in the context of the IRIS design. From this analysis, 54 known IRIS operating cycle length barriers were identified. The resolution methodology was applied to each of these barriers to generate design solution alternatives for consideration in the IRIS design. The methodology developed has been demonstrated to narrow the design space to feasible design solutions which enable a desired operating cycle length, yet is general enough to have broad applicability. Feedback from the IRIS design team

  2. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, July 1--September 30, 1978

    Energy Technology Data Exchange (ETDEWEB)

    1978-11-24

    Results of work performed from July 1, 1978 through September 30, 1978 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. Candidate alloys were evaluated for Very High Temperature Reactor Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the affect of simulated reactor primary coolant (Helium containing small amounts of various other gases), the high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. The activities associated with the characterization of the materials for the screening test program are reported, i.e., test specimen preparation, information from the materials characterization tests performed by General Electric, and the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment. The status of the data management system is presented.

  3. Preliminary analysis of Th-U conversion performance in a ZrH-moderated molten salt reactor%氢化锆慢化熔盐堆钍铀转换性能初步分析

    Institute of Scientific and Technical Information of China (English)

    吴攀; 蔡翔舟; 余呈刚; 陈金根; 徐刚

    2016-01-01

    中子能谱对钍基燃料在熔盐堆中的利用效率及温度反馈系数等安全问题有较大影响,所以对熔盐堆新型慢化剂的研究具有重要意义。本工作基于 SCALE6计算程序,对不同几何栅元结构的氢化锆栅元组件在熔盐堆的物理性能进行了研究,分别计算了中子能谱、钍铀转换比、233U 浓度、总温度反馈系数以及燃耗等中子物理参量。结果表明,减小六边形栅元对边距或者增加熔盐占栅元体积比可以增加钍铀转换比和改善温度反应性系数;当加入的氢化锆慢化剂体积份额为0.1时就可以将熔盐堆233U初始浓度降低到2.5×10−2以内;氢化锆慢化熔盐堆在超热谱条件下,其233U初装载量和超铀核素产量较小,同时堆芯较为紧凑。%Background: The neutron spectrum plays an important role in thorium-based fuel utilization efficiency and temperature feedback coefficient concerning reactor operation safety, so it is very important to study the new moderator material used in molten salt reactor (MSR).Purpose: This study aims to analyze thethorium-uranium conversion performance of a ZrH-moderated molten salt reactor and analyze the feasibility of ZrH as moderator in molten salt reactor.Methods: SCALE program is used to calculate neutron spectrum, thorium uranium conversion ratio,233U concentration, total temperature feedback coefficient and burnup calculation with different lattice parameters.Results: The thorium uranium conversion ratio and total temperature feedback coefficient can be improved significantly by reducing lattice size or increasing salt volume ratio; the initial233U concentration for start reactor can be easily controlled under 2.5×10−2 when the volume share of added ZrH is 0.1.Conclusion:Compared to the graphite-moderated MSR, ZrH-moderated MSR reduces initial233U inventory and transuraniums (TRUs) production, and makes its core more compact.

  4. Application of the successive linear programming technique to the optimum design of a high flux reactor using LEU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mo, S.C.

    1991-01-01

    The successive linear programming technique is applied to obtain the optimum thermal flux in the reflector region of a high flux reactor using LEU fuel. The design variables are the reactor power, core radius and coolant channel thickness. The constraints are the cycle length, average heat flux and peak/average power density ratio. The characteristics of the optimum solutions with various constraints are discussed.

  5. Application of the successive linear programming technique to the optimum design of a high flux reactor using LEU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mo, S.C.

    1991-12-31

    The successive linear programming technique is applied to obtain the optimum thermal flux in the reflector region of a high flux reactor using LEU fuel. The design variables are the reactor power, core radius and coolant channel thickness. The constraints are the cycle length, average heat flux and peak/average power density ratio. The characteristics of the optimum solutions with various constraints are discussed.

  6. Modeling of packed bed membrane reactors: impact of oxygen distribution on conversion and selectivity in partial oxidation systems

    NARCIS (Netherlands)

    Kürten, Ulrich

    2003-01-01

    In this thesis the focus is on the effect of the oxygen distribution on the performance of the PBMR in terms of conversion and selectivities. A detailed description of the oxygen transport through the membrane is not considered in this work. Here, the membrane is considered as an ideal distributor w

  7. OSMOSE program : statistical review of oscillation measurements in the MINERVE reactor R1-UO2 configuration.

    Energy Technology Data Exchange (ETDEWEB)

    Stoven, G.; Klann, R.; Zhong, Z.; Nuclear Engineering Division

    2007-08-28

    The OSMOSE program is a collaboration on reactor physics experiments between the United States Department of Energy and the France Commissariat Energie Atomique. At the working level, it is a collaborative effort between the Argonne National Laboratory and the CEA Cadarache Research Center. The objective of this program is to measure very accurate integral reaction rates in representative spectra for the actinides important to future nuclear system designs, and to provide the experimental data for improving the basic nuclear data files. The main outcome of the OSMOSE measurement program will be an experimental database of reactivity-worth measurements in different neutron spectra for the heavy nuclides. This database can then be used as a benchmark to verify and validate reactor analysis codes. The OSMOSE program (Oscillation in Minerve of isotopes in Eupraxic Spectra) aims at improving neutronic predictions of advanced nuclear fuels through oscillation measurements in the MINERVE facility on samples containing the following separated actinides: {sup 232}Th, {sup 233}U, {sup 234}U, {sup 235}U, {sup 236}U, {sup 238}U, {sup 237}Np, {sup 238}Pu, {sup 239}Pu, {sup 240}Pu, {sup 241}Pu, {sup 242}Pu, {sup 241}Am, {sup 243}Am, {sup 244}Cm, and {sup 245}Cm. The first part of this report provides an overview of the experimental protocol and the typical processing of a series of experimental results which is currently performed at CEA-Cadarache. In the second part of the report, improvements to this technique are presented, as well as the program that was created to process oscillation measurement results from the MINERVE facility in the future.

  8. Conversion of activated-sludge reactors to microbial fuel cells for wastewater treatment coupled to electricity generation.

    Science.gov (United States)

    Yoshizawa, Tomoya; Miyahara, Morio; Kouzuma, Atsushi; Watanabe, Kazuya

    2014-11-01

    Wastewater can be treated in microbial fuel cells (MFCs) with the aid of microbes that oxidize organic compounds using anodes as electron acceptors. Previous studies have suggested the utility of cassette-electrode (CE) MFCs for wastewater treatment, in which rice paddy-field soil was used as the inoculum. The present study attempted to convert an activated-sludge (AS) reactor to CE-MFC and use aerobic sludge in the tank as the source of microbes. We used laboratory-scale (1 L in capacity) reactors that were initially operated in an AS mode to treat synthetic wastewater, containing starch, yeast extract, peptone, plant oil, and detergents. After the organics removal became stable, the aeration was terminated, and CEs were inserted to initiate an MFC-mode operation. It was demonstrated that the MFC-mode operation treated the wastewater at similar efficiencies to those observed in the AS-mode operation with COD-removal efficiencies of 75-80%, maximum power densities of 150-200 mW m(-2) and Coulombic efficiencies of 20-30%. These values were similar to those of CE-MFC inoculated with the soil. Anode microbial communities were analyzed by pyrotag sequencing of 16S rRNA gene PCR amplicons. Comparative analyses revealed that anode communities enriched from the aerobic sludge were largely different from those from the soil, suggesting that similar reactor performances can be supported by different community structures. The study demonstrates that it is possible to construct wastewater-treatment MFCs by inserting CEs into water-treatment tanks.

  9. Standard guide for establishing a quality assurance program for uranium conversion facilities

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2004-01-01

    1.1 This guide provides guidance and recommended practices for establishing a comprehensive quality assurance program for uranium conversion facilities. 1.2 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate health and safety practices and determine the applicability of regulatory limitations prior to use. 1.3 The basic elements of a quality assurance program appear in the following order: FUNCTION SECTION Organization 5 Quality Assurance Program 6 Design Control 7 Instructions, Procedures & Drawings 8 Document Control 9 Procurement 10 Identification and Traceability 11 Processes 12 Inspection 13 Control of Measuring and Test Equipment 14 Handling, Storage and Shipping 15 Inspection, Test and Operating Status 16 Control of Nonconforming Items 17 Corrective Actions 18 Quality Assurance Records 19 Audits 20 TABLE 1 NQA-1 Basic Requirements Relat...

  10. SOLID STATE ENERGY CONVERSION ALLIANCE (SECA) SOLID OXIDE FUEL CELL PROGRAM

    Energy Technology Data Exchange (ETDEWEB)

    Unknown

    2003-06-01

    This report summarizes the progress made during the September 2001-March 2002 reporting period under Cooperative Agreement DE-FC26-01NT41245 for the U. S. Department of Energy, National Energy Technology Laboratory (DOE/NETL) entitled ''Solid State Energy Conversion Alliance (SECA) Solid Oxide Fuel Cell Program''. The program focuses on the development of a low-cost, high-performance 3-to-10-kW solid oxide fuel cell (SOFC) system suitable for a broad spectrum of power-generation applications. The overall objective of the program is to demonstrate a modular SOFC system that can be configured to create highly efficient, cost-competitive, and environmentally benign power plants tailored to specific markets. When fully developed, the system will meet the efficiency, performance, life, and cost goals for future commercial power plants.

  11. RISMC advanced safety analysis working plan: FY2015 - FY2019. Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo H; Smith, Curtis L

    2014-09-01

    In this report, the Advanced Safety Analysis Program (ASAP) objectives and value proposition is described. ASAP focuses on modernization of nuclear power safety analysis (tools, methods and data); implementing state-of-the-art modeling techniques (which include, for example, enabling incorporation of more detailed physics as they become available); taking advantage of modern computing hardware; and combining probabilistic and mechanistic analyses to enable a risk informed safety analysis process. The modernized tools will maintain the current high level of safety in our nuclear power plant fleet, while providing an improved understanding of safety margins and the critical parameters that affect them. Thus, the set of tools will provide information to inform decisions on plant modifications, refurbishments, and surveillance programs, while improving economics. The set of tools will also benefit the design of new reactors, enhancing safety per unit cost of a nuclear plant. As part of the discussion, we have identified three sets of stakeholders, the nuclear industry, the Department of Energy (DOE), and associated oversight organizations. These three groups would benefit from ASAP in different ways. For example, within the DOE complex, the possible applications that are seen include the safety of experimental reactors, facility life extension, safety-by-design in future generation advanced reactors, and managing security for the storage of nuclear material. This report provides information in five areas: (1) A value proposition (“why is this important?”) that will make the case for stakeholder’s use of the ASAP research and development (R&D) products; (2) An identification of likely end users and pathway to adoption of enhanced tools by the end-users; (3) A proposed set of practical and achievable “use case” demonstrations; (4) A proposed plan to address ASAP verification and validation (V&V) needs; and (5) A proposed schedule for the multi-year ASAP.

  12. Conversion of the CALAP (Computer Aided Landform Analysis Program) Program from FORTRAN to DUCK.

    Science.gov (United States)

    1986-09-01

    t 10 "We will assume you have a 1:50,000 scale topographic" t 5 "map sheet/sheets containing a rectangular geographic " t ( 5 "area in which you wish...34I *- I,] : -g DETA. L.14 2-SEP-19e3 11:09 Page 1 * P5Sytm Kenneth Hayjes Smart SsesTcnlg PROGRAM DETA(5) DETAILED ANALYSIS SEGEMENT FOR PROGRAM

  13. Biocatalysis conversion of methanol to methane in an upflow anaerobic sludge blanket (UASB) reactor: Long-term performance and inherent deficiencies.

    Science.gov (United States)

    Lu, Xueqin; Zhen, Guangyin; Chen, Mo; Kubota, Kengo; Li, Yu-You

    2015-12-01

    Long-term performance of methanol biocatalysis conversion in a lab-scale UASB reactor was evaluated. Properties of granules were traced to examine the impact of methanol on granulation. Methanolic wastewater could be stably treated during initial 240d with the highest biogas production rate of 18.6 ± 5.7 L/Ld at OLR 48 g-COD/Ld. However, the reactor subsequently showed severe granule disintegration, inducing granule washout and process upsets. Some steps (e.g. increasing influent Ca(2+) concentration, etc.) were taken to prevent rising dispersion, but no clear improvement was observed. Further characterizations in granules revealed that several biotic/abiotic factors all caused the dispersion: (1) depletion of extracellular polymeric substances (EPS) and imbalance of protein/polysaccharide ratio in EPS; (2) restricted formation of hard core and weak Ca-EPS bridge effect due to insufficient calcium supply; and (3) simplification of species with the methanol acclimation. More efforts are required to solve the technical deficiencies observed in methanolic wastewater treatment.

  14. Conversion of olive pomace oil to cocoa butter-like fat in a packed-bed enzyme reactor.

    Science.gov (United States)

    Ciftçi, Ozan Nazim; Fadiloğlu, Sibel; Göğüş, Fahrettin

    2009-01-01

    Refined olive pomace oil (ROPO) was utilized as a source oil for production of cocoa butter-like fat. Immobilized sn-1,3 specific lipase catalyzed acidolysis of ROPO with palmitic (PA) and stearic (SA) acids was performed in a laboratory scale packed-bed reactor. Effect of reactor conditions on product formation was studied at various substrate mole ratios (ROPO:PA:SA; 1:1:1, 1:1:3, 1:3:3, 1:2:6), enzyme loads (10%, 20%, 40%), substrate flow rates (1.5, 4.5, 7.5, 15 ml/min) and solvent amounts (150, 400 ml). The highest yield (10.9% POP, 19.7% POS and 11.2% SOS) was obtained at 40% enzyme load, 1:2:6 substrate mole ratio, 45 degrees C, 7.5 ml/min substrate flow rate, 150 ml solvent and 3h reaction time. The melting profile and SFC of the product were comparable to those of CB. Polarized light microscope (PLM) images showed no drastic changes in polymorphic behavior between CB and product.

  15. Modeling of packed bed membrane reactors: impact of oxygen distribution on conversion and selectivity in partial oxidation systems

    OpenAIRE

    Kürten, Ulrich

    2003-01-01

    In this thesis the focus is on the effect of the oxygen distribution on the performance of the PBMR in terms of conversion and selectivities. A detailed description of the oxygen transport through the membrane is not considered in this work. Here, the membrane is considered as an ideal distributor with a fixed membrane flow, independent from the reaction conditions inside the packed bed. This assumption enables an independent discussion of the influence of the oxygen distribution on the perfo...

  16. Light Water Reactor Sustainability Program: Risk-Informed Safety Margins Characterization (RISMC) Pathway Technical Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Martineau, Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Szilard, Ronaldo [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). As the current Light Water Reactor (LWR) NPPs age beyond 60 years, there are possibilities for increased frequency of Systems, Structures, and Components (SSCs) degradations or failures that initiate safety-significant events, reduce existing accident mitigation capabilities, or create new failure modes. Plant designers commonly “over-design” portions of NPPs and provide robustness in the form of redundant and diverse engineered safety features to ensure that, even in the case of well-beyond design basis scenarios, public health and safety will be protected with a very high degree of assurance. This form of defense-in-depth is a reasoned response to uncertainties and is often referred to generically as “safety margin.” Historically, specific safety margin provisions have been formulated, primarily based on “engineering judgment.”

  17. Thermal Analysis for Mobile Reactor

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>Mobile reactor design in the paper is consisted of two grades of thermal electric conversion. The first grade is the thermionic conversion inside the core and the second grade is thermocouple conversion

  18. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, April 1--June 30, 1978

    Energy Technology Data Exchange (ETDEWEB)

    1978-08-31

    The objectives of the program are to evaluate candidate alloys for Very High Temperature Reactor Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the affect of simulated reactor primary coolant (Helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in the report includes the activities associated with the procurement of the materials for the screening test program, information from vendor certification for the materials receiver, and preliminary information from the materials characterization tests performed by General Electric. The construction status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment are discussed. The status of the data management system is also reviewed.

  19. Optimal conditions and operational parameters for conversion of Robusta coffee residues in a continuous stirred tank reactor

    Energy Technology Data Exchange (ETDEWEB)

    Msambichaka, B.L.; Kivaisi, A.K.; Rubindamayugi, M.S.T. [Univ. of Dar es Salaam, Applied Microbiology Unit (Tanzania, United Republic of)

    1997-12-31

    This experiment studied the possibility of optimizing anaerobic degradation, developing microbial adaptation and establishing long term process stability in a Continuous Stirred Tank Reactor (CSTR) running on Robusta coffee hulls as feed substrate. Decrease in lag phase and increase in methane production rate in batch culture experiment conducted before and after process stabilization of each operational phase in the CSTR clearly suggested that microbial adaptation to increasing coffee percentage composition was attained. Through gradual increase of coffee percentage composition, from 10% coffee, 2% VS, 20 days HRT and a 1 g VS/1/day loading rate to 80% coffee, 4.5% VS, 12 days HRT and a loading rate of 3 g VS/1/day the CSTR system was optimized at a maximum methane yield of 535 ml/g VS. Again it was possible to attain long term process stability at the above mentioned optimal operational parameters for a further 3 month period. (au)

  20. Mathematical programming models for the economic design and assessment of wind energy conversion systems

    Science.gov (United States)

    Reinert, K. A.

    The use of linear decision rules (LDR) and chance constrained programming (CCP) to optimize the performance of wind energy conversion clusters coupled to storage systems is described. Storage is modelled by LDR and output by CCP. The linear allocation rule and linear release rule prescribe the size and optimize a storage facility with a bypass. Chance constraints are introduced to explicitly treat reliability in terms of an appropriate value from an inverse cumulative distribution function. Details of deterministic programming structure and a sample problem involving a 500 kW and a 1.5 MW WECS are provided, considering an installed cost of $1/kW. Four demand patterns and three levels of reliability are analyzed for optimizing the generator choice and the storage configuration for base load and peak operating conditions. Deficiencies in ability to predict reliability and to account for serial correlations are noted in the model, which is concluded useful for narrowing WECS design options.

  1. A Computer Program for Modeling the Conversion of Organic Waste to Energy

    Directory of Open Access Journals (Sweden)

    Pragasen Pillay

    2011-11-01

    Full Text Available This paper presents a tool for the analysis of conversion of organic waste into energy. The tool is a program that uses waste characterization parameters and mass flow rates at each stage of the waste treatment process to predict the given products. The specific waste treatment process analysed in this paper is anaerobic digestion. The different waste treatment stages of the anaerobic digestion process are: conditioning of input waste, secondary treatment, drying of sludge, conditioning of digestate, treatment of digestate, storage of liquid and solid effluent, disposal of liquid and solid effluents, purification, utilization and storage of combustible gas. The program uses mass balance equations to compute the amount of CH4, NH3, CO2 and H2S produced from anaerobic digestion of organic waste, and hence the energy available. Case studies are also presented.

  2. Evaluating China's Slope Land Conversion Program as sustainable management in Tianquan and Wuqi Counties.

    Science.gov (United States)

    Bullock, Allison; King, Brian

    2011-08-01

    Increased soil erosion on sloped land has become a significant environmental concern in China that has been attributed to human activities such as deforestation, over-cultivation, and over-grazing of livestock. In order to reduce soil erosion on sloped lands, the Chinese government has responded by implementing large-scale, ecological rehabilitation programs, including the "Grain for Green" reforestation project. This program involves financial incentives to transition farmers into other economic activities with the goal of reducing ecological pressures and degradation. Because of the scope and potential impacts from these programs, detailed research is needed to understand their social and ecological effects. This paper reports on research conducted in Tianquan County, Sichuan Province, and Wuqi County, Shaanxi Province, that evaluates the effects of the program upon local economies and household livelihood systems. The paper argues that the successful conversion of farmland under "Grain for Green" depends upon local government involvement, local economic development, and funding for local projects. Without economic development within rural economies, we conclude that farmers will remain dependent upon continued subsidy assistance to meet the policy's ambitious environmental restrictions, thereby undermining the program's long-term sustainability. Copyright © 2011 Elsevier Ltd. All rights reserved.

  3. Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 3. Appendix A. Equipment list

    Energy Technology Data Exchange (ETDEWEB)

    1979-11-01

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. Volume 1 reports the overall plant and reactor system and was prepared by the General Electric Company. Core scoping studies were performed which evaluated the effects of annular and cylindrical core configurations, radial blanket zones, burnup, and ball heavy metal loadings. The reactor system, including the PCRV, was investigated for both the annular and cylindrical core configurations. Volume 3 is an Appendix containing the equipment list for the plant and was also prepared by United Engineers and Constructors, Inc. It tabulates the major components of the plant and describes each in terms of quantity, type, orientation, etc., to provide a basis for cost estimation.

  4. AURORA: A FORTRAN program for modeling well stirred plasma and thermal reactors with gas and surface reactions

    Energy Technology Data Exchange (ETDEWEB)

    Meeks, E.; Grcar, J.F.; Kee, R.J. [Sandia National Labs., Livermore, CA (United States). Thermal and Plasma Processes Dept.; Moffat, H.K. [Sandia National Labs., Albuquerque, NM (United States). Surface Processing Sciences Dept.

    1996-02-01

    The AURORA Software is a FORTRAN computer program that predicts the steady-state or time-averaged properties of a well mixed or perfectly stirred reactor for plasma or thermal chemistry systems. The software was based on the previously released software, SURFACE PSR which was written for application to thermal CVD reactor systems. AURORA allows modeling of non-thermal, plasma reactors with the determination of ion and electron concentrations and the electron temperature, in addition to the neutral radical species concentrations. Well stirred reactors are characterized by a reactor volume, residence time or mass flow rate, heat loss or gas temperature, surface area, surface temperature, the incoming temperature and mixture composition, as well as the power deposited into the plasma for non-thermal systems. The model described here accounts for finite-rate elementary chemical reactions both in the gas phase and on the surface. The governing equations are a system of nonlinear algebraic relations. The program solves these equations using a hybrid Newton/time-integration method embodied by the software package TWOPNT. The program runs in conjunction with the new CHEMKIN-III and SURFACE CHEMKIN-III packages, which handle the chemical reaction mechanisms for thermal and non-thermal systems. CHEMKIN-III allows for specification of electron-impact reactions, excitation losses, and elastic-collision losses for electrons.

  5. Guide for licensing evaluations using CRAC2: A computer program for calculating reactor accident consequences

    Energy Technology Data Exchange (ETDEWEB)

    White, J.E.; Roussin, R.W.; Gilpin, H.

    1988-12-01

    A version of the CRAC2 computer code applicable for use in analyses of consequences and risks of reactor accidents in case work for environmental statements has been implemented for use on the Nuclear Regulatory Commission Data General MV/8000 computer system. Input preparation is facilitated through the use of an interactive computer program which operates on an IBM personal computer. The resulting CRAC2 input deck is transmitted to the MV/8000 by using an error-free file transfer mechanism. To facilitate the use of CRAC2 at NRC, relevant background material on input requirements and model descriptions has been extracted from four reports - ''Calculations of Reactor Accident Consequences,'' Version 2, NUREG/CR-2326 (SAND81-1994) and ''CRAC2 Model Descriptions,'' NUREG/CR-2552 (SAND82-0342), ''CRAC Calculations for Accident Sections of Environmental Statements, '' NUREG/CR-2901 (SAND82-1693), and ''Sensitivity and Uncertainty Studies of the CRAC2 Computer Code,'' NUREG/CR-4038 (ORNL-6114). When this background information is combined with instructions on the input processor, this report provides a self-contained guide for preparing CRAC2 input data with a specific orientation toward applications on the MV/8000. 8 refs., 11 figs., 10 tabs.

  6. Solid State Energy Conversion Alliance (SECA) Solid Oxide Fuel Cell Program

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen Minh

    2006-07-31

    This report summarizes the work performed for Phase I (October 2001 - August 2006) under Cooperative Agreement DE-FC26-01NT41245 for the U. S. Department of Energy, National Energy Technology Laboratory (DOE/NETL) entitled 'Solid State Energy Conversion Alliance (SECA) Solid Oxide Fuel Cell Program'. The program focuses on the development of a low-cost, high-performance 3-to-10-kW solid oxide fuel cell (SOFC) system suitable for a broad spectrum of power-generation applications. During Phase I of the program significant progress has been made in the area of SOFC technology. A high-efficiency low-cost system was designed and supporting technology developed such as fuel processing, controls, thermal management, and power electronics. Phase I culminated in the successful demonstration of a prototype system that achieved a peak efficiency of 41%, a high-volume cost of $724/kW, a peak power of 5.4 kW, and a degradation rate of 1.8% per 500 hours. . An improved prototype system was designed, assembled, and delivered to DOE/NETL at the end of the program. This prototype achieved an extraordinary peak efficiency of 49.6%.

  7. High-Temperature Gas-Cooled Reactor Technology Development Program: Annual progress report for period ending December 31, 1987

    Energy Technology Data Exchange (ETDEWEB)

    Jones, J.E.,Jr.; Kasten, P.R.; Rittenhouse, P.L.; Sanders, J.P.

    1989-03-01

    The High-Temperature Gas-Cooled Reactor (HTGR) Program being carried out under the US Department of Energy (DOE) continues to emphasize the development of modular high-temperature gas-cooled reactors (MHTGRs) possessing a high degree of inherent safety. The emphasis at this time is to develop the preliminary design of the reference MHTGR and to develop the associated technology base and licensing infrastructure in support of future reactor deployment. A longer-term objective is to realize the full high-temperature potential of HTGRs in gas turbine and high-temperature, process-heat applications. This document summarizes the activities of the HTGR Technology Development Program for the period ending December 31, 1987.

  8. Light Water Reactor Sustainability Program A Reference Plan for Control Room Modernization: Planning and Analysis Phase

    Energy Technology Data Exchange (ETDEWEB)

    Jacques Hugo; Ronald Boring; Lew Hanes; Kenneth Thomas

    2013-09-01

    The U.S. Department of Energy’s Light Water Reactor Sustainability (LWRS) program is collaborating with a U.S. nuclear utility to bring about a systematic fleet-wide control room modernization. To facilitate this upgrade, a new distributed control system (DCS) is being introduced into the control rooms of these plants. The DCS will upgrade the legacy plant process computer and emergency response facility information system. In addition, the DCS will replace an existing analog turbine control system with a display-based system. With technology upgrades comes the opportunity to improve the overall human-system interaction between the operators and the control room. To optimize operator performance, the LWRS Control Room Modernization research team followed a human-centered approach published by the U.S. Nuclear Regulatory Commission. NUREG-0711, Rev. 3, Human Factors Engineering Program Review Model (O’Hara et al., 2012), prescribes four phases for human factors engineering. This report provides examples of the first phase, Planning and Analysis. The three elements of Planning and Analysis in NUREG-0711 that are most crucial to initiating control room upgrades are: • Operating Experience Review: Identifies opportunities for improvement in the existing system and provides lessons learned from implemented systems. • Function Analysis and Allocation: Identifies which functions at the plant may be optimally handled by the DCS vs. the operators. • Task Analysis: Identifies how tasks might be optimized for the operators. Each of these elements is covered in a separate chapter. Examples are drawn from workshops with reactor operators that were conducted at the LWRS Human System Simulation Laboratory HSSL and at the respective plants. The findings in this report represent generalized accounts of more detailed proprietary reports produced for the utility for each plant. The goal of this LWRS report is to disseminate the technique and provide examples sufficient to

  9. Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 2. Conceptual balance of plant design

    Energy Technology Data Exchange (ETDEWEB)

    1979-11-01

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. This volume describes the conceptual balance-of-plant (BOP) design and was prepared by United Engineers and Constructors, Inc. of Philadelphia, Pennsylvania. The major emphasis of the BOP study was a preliminary design of an overall plant to provide a basis for future studies.

  10. Reactor safety issues resolved by the 2D/3D program

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    The 2D/3D Program studied multidimensional thermal-hydraulics in a PWR core and primary system during the end-of-blowdown and post-blowdown phases of a large-break LOCA (LBLOCA), and during selected small-break LOCA (SBLOCA) transients. The program included tests at the Cylindrical Core Test Facility (CCTF), the Slab Core Test Facility (SCTF), and the Upper Plenum Test Facility (UPTF), and computer analyses using TRAC. Tests at CCTF investigated core thermal-hydraulics and overall system behavior while tests at SCTF concentrated on multidimensional core thermal-hydraulics. The UPTF tests investigated two-phase flow behavior in the downcomer, upper plenum, tie plate region, and primary loops. TRAC analyses evaluated thermal-hydraulic behavior throughout the primary system in tests as well as in PWRs. This report summarizes the test and analysis results in each of the main areas where improved information was obtained in the 2D/3D Program. The discussion is organized in terms of the reactor safety issues investigated. This report was prepared in a coordination among US, Germany and Japan. US and Germany have published the report as NUREG/IA-0127 and GRS-101 respectively. (author).

  11. Research and Development of Multiphysics Models in Support of the Conversion of the High Flux Isotope Reactor to Low Enriched Uranium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bodey, Isaac T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Curtis, Franklin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Arimilli, Rao V. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ekici, Kivanc [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Freels, James D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-11-01

    ABSTRACT The findings presented in this report are results of a five year effort lead by the RRD Division of the ORNL, which is focused on research and development toward the conversion of the High Flux Isotope Reactor (HFIR) fuel from high-enriched uranium (HEU) to low-enriched uranium (LEU). This report focuses on the tasks accomplished by the University of Tennessee Knoxville (UTK) team from the Department of Mechanical, Aerospace, and Biomedical Engineering (MABE) that provided expert support in multiphysics modeling of complex problems associated with the LEU conversion of the HFIR reactor. The COMSOL software was used as the main computational modeling tool, whereas Solidworks was also used in support of computer-aided-design (CAD) modeling of the proposed LEU fuel design. The UTK research has been governed by a statement of work (SOW), which was updated annually to clearly define the specific tasks reported herein. Ph.D. student Isaac T. Bodey has focused on heat transfer and fluid flow modeling issues and has been aided by his major professor Dr. Rao V. Arimilli. Ph.D. student Franklin G. Curtis has been focusing on modeling the fluid-structure interaction (FSI) phenomena caused by the mechanical forces acting on the fuel plates, which in turn affect the fluid flow in between the fuel plates, and ultimately the heat transfer, is also affected by the FSI changes. Franklin Curtis has been aided by his major professor Dr. Kivanc Ekici. M.Sc. student Adam R. Travis has focused two major areas of research: (1) on accurate CAD modeling of the proposed LEU plate design, and (2) reduction of the model complexity and dimensionality through interdimensional coupling of the fluid flow and heat transfer for the HFIR plate geometry. Adam Travis is also aided by his major professor, Dr. Kivanc Ekici. We must note that the UTK team, and particularly the graduate students, have been in very close collaboration with Dr. James D. Freels (ORNL technical monitor and mentor

  12. Research and Development of Multiphysics Models in Support of the Conversion of the High Flux Isotope Reactor to Low Enriched Uranium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bodey, Isaac T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Curtis, Franklin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Arimilli, Rao V. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ekici, Kivanc [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Freels, James D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-11-01

    The findings presented in this report are results of a five year effort led by the RRD Division of the ORNL, which is focused on research and development toward the conversion of the High Flux Isotope Reactor (HFIR) fuel from high-enriched uranium (HEU) to low-enriched uranium (LEU). This report focuses on the tasks accomplished by the University of Tennessee Knoxville (UTK) team from the Department of Mechanical, Aerospace, and Biomedical Engineering (MABE) that provided expert support in multiphysics modeling of complex problems associated with the LEU conversion of the HFIR reactor. The COMSOL software was used as the main computational modeling tool, whereas Solidworks was also used in support of computer-aided-design (CAD) modeling of the proposed LEU fuel design. The UTK research has been governed by a statement of work (SOW), which was updated annually to clearly define the specific tasks reported herein. Ph.D. student Isaac T. Bodey has focused on heat transfer and fluid flow modeling issues and has been aided by his major professor Dr. Rao V. Arimilli. Ph.D. student Franklin G. Curtis has been focusing on modeling the fluid-structure interaction (FSI) phenomena caused by the mechanical forces acting on the fuel plates, which in turn affect the fluid flow in between the fuel plates, and ultimately the heat transfer, is also affected by the FSI changes. Franklin Curtis has been aided by his major professor Dr. Kivanc Ekici. M.Sc. student Adam R. Travis has focused two major areas of research: (1) on accurate CAD modeling of the proposed LEU plate design, and (2) reduction of the model complexity and dimensionality through interdimensional coupling of the fluid flow and heat transfer for the HFIR plate geometry. Adam Travis is also aided by his major professor, Dr. Kivanc Ekici. We must note that the UTK team, and particularly the graduate students, have been in very close collaboration with Dr. James D. Freels (ORNL technical monitor and mentor) and have

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

    Energy Technology Data Exchange (ETDEWEB)

    David Petti

    2014-06-01

    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 UO

  14. Light Water Reactor Sustainability Program: Analysis of Pressurized Water Reactor Station Blackout caused by external flooding using the RISMC toolkit

    Energy Technology Data Exchange (ETDEWEB)

    Mandelli, Diego; Smith, Curtis; Prescott, Steven; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua; Kinoshita, Robert

    2014-08-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impacts of these two factors on the safety of the plant, the Risk Informed Safety Margin Characterization project aims to provide insights to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This paper focuses on the impacts of power uprate on the safety margin of a boiling water reactor for a flooding induced station black-out event. Analysis is performed by using a combination of thermal-hydraulic codes and a stochastic analysis tool currently under development at the Idaho National Laboratory, i.e. RAVEN. We employed both classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. Results obtained give a detailed investigation of the issues associated with a plant power uprate including the effects of station black-out accident scenarios. We were able to quantify how the timing of specific events was impacted by a higher nominal reactor core power. Such safety insights can provide useful information to the decision makers to perform risk informed margins management.

  15. Participation in the United States Department of Energy Reactor Sharing Program. Annual report, August 31, 1991--August 29, 1992

    Energy Technology Data Exchange (ETDEWEB)

    Mulder, R.U.; Benneche, P.E.; Hosticka, B.

    1992-05-01

    The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics (to become the Department of Mechanical, Aerospace and Nuclear Engineering on July 1, 1992). As such, it is effectively used to support educational programs in engineering and science at the University of Virginia as well as those at other area colleges and universities. The expansion of support to educational programs in the mid-east region is a major objective. To assist in meeting this objective, the University of Virginia has been supported under the US Department of Energy (DOE) Reactor Sharing Program since 1978. Due to the success of the program, this proposal requests continued DOE support through August 1993.

  16. Planning Document for an NBSR Conversion Safety Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    Diamond D. J.; Baek J.; Hanson, A.L.; Cheng, L-Y.; Brown, N.; Cuadra, A.

    2013-09-25

    The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the National Bureau of Standards Reactor (NBSR). The NBSR is a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in an aluminum alloy, and the development of the fabrication techniques. This report is a planning document for the conversion Safety Analysis Report (SAR) that would be submitted to, and approved by, the Nuclear Regulatory Commission (NRC) before the reactor could be converted.This report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis herein is on the SAR chapters that require significant changes as a result of conversion, primarily Chapter 4, Reactor Description, and Chapter 13, Safety Analysis. The document provides information on the proposed design for the LEU fuel elements and identifies what information is still missing. This document is intended to assist ongoing fuel development efforts, and to provide a platform for the development of the final conversion SAR. This report contributes directly to the reactor conversion pillar of the GTRI program, but also acts as a boundary condition for the fuel development and fuel fabrication pillars.

  17. SOLID STATE ENERGY CONVERSION ALLIANCE (SECA) SOLID OXIDE FUEL CELL PROGRAM

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen Minh; Jim Powers

    2003-10-01

    This report summarizes the work performed for April 2003--September 2003 reporting period under Cooperative Agreement DE-FC26-01NT41245 for the U.S. Department of Energy, National Energy Technology Laboratory (DOE/NETL) entitled ''Solid State Energy Conversion Alliance (SECA) Solid oxide Fuel Cell Program''. During this reporting period, the conceptual system design activity was completed. The system design, including strategies for startup, normal operation and shutdown, was defined. Sealant and stack materials for the solid oxide fuel cell (SOFC) stack were identified which are capable of meeting the thermal cycling and degradation requirements. A cell module was tested which achieved a stable performance of 0.238 W/cm{sup 2} at 95% fuel utilization. The external fuel processor design was completed and fabrication begun. Several other advances were made on various aspects of the SOFC system, which are detailed in this report.

  18. DOE-NE Light Water Reactor Sustainability Program and EPRI Long-Term Operations Program. Joint Research and Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    Williams, Don

    2014-04-01

    Nuclear power has contributed almost 20% of the total amount of electricity generated in the United States over the past two decades. High capacity factors and low operating costs make nuclear power plants (NPPs) some of the most economical power generators available. Further, nuclear power remains the single largest contributor (nearly 70%) of non-greenhouse gas-emitting electric power generation in the United States. Even when major refurbishments are performed to extend operating life, these plants continue to represent cost-effective, low-carbon assets to the nation’s electrical generation capability. By the end of 2014, about one-third of the existing domestic fleet will have passed their 40th anniversary of power operations, and about one-half of the fleet will reach the same 40-year mark within this decade. Recognizing the challenges associated with pursuing extended service life of commercial nuclear power plants, the U.S. Department of Energy’s (DOE) Office of Nuclear Energy (NE) and the Electric Power Research Institute (EPRI) have established separate but complementary research and development programs (DOE-NE’s Light Water Reactor Sustainability [LWRS] Program and EPRI’s Long-Term Operations [LTO] Program) to address these challenges. To ensure that a proper linkage is maintained between the programs, DOE-NE and EPRI executed a memorandum of understanding in late 2010 to “establish guiding principles under which research activities (between LWRS and LTO) could be coordinated to the benefit of both parties.” This document represents the third annual revision to the initial version (March 2011) of the plan as called for in the memorandum of understanding.

  19. Light Water Reactor Sustainability Research and Development Program Plan -- Fiscal Year 2009–2013

    Energy Technology Data Exchange (ETDEWEB)

    Idaho National Laboratory

    2009-12-01

    Nuclear power has reliably and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. By the year 2030, domestic demand for electrical energy is expected to grow to levels of 16 to 36% higher than 2007 levels. At the same time, most currently operating nuclear power plants will begin reaching the end of their 60-year operating licenses. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary this year. U.S. regulators have begun considering extended operations of nuclear power plants and the research needed to support long-term operations. The Light Water Reactor Sustainability (LWRS) Research and Development (R&D) Program, developed and sponsored by the Department of Energy, is performed in close collaboration with industry R&D programs. The purpose of the LWRS R&D Program is to provide technical foundations for licensing and managing long-term, safe and economical operation of the current operating nuclear power plants. The LWRS R&D Program vision is captured in the following statements: Existing operating nuclear power plants will continue to safely provide clean and economic electricity well beyond their first license- extension period, significantly contributing to reduction of United States and global carbon emissions, enhancement of national energy security, and protection of the environment. There is a comprehensive technical basis for licensing and managing the long-term, safe, economical operation of nuclear power plants. Sustaining the existing operating U.S. fleet also will improve its international engagement

  20. Light Water Reactor Sustainability Program: Digital Technology Business Case Methodology Guide

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, Ken [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lawrie, Sean [ScottMadden, Inc., Raleigh, NC (United States); Hart, Adam [ScottMadden, Inc., Raleigh, NC (United States); Vlahoplus, Chris [ScottMadden, Inc., Raleigh, NC (United States)

    2014-09-01

    The Department of Energy’s (DOE’s) Light Water Reactor Sustainability Program aims to develop and deploy technologies that will make the existing U.S. nuclear fleet more efficient and competitive. The program has developed a standard methodology for determining the impact of new technologies in order to assist nuclear power plant (NPP) operators in building sound business cases. The Advanced Instrumentation, Information, and Control (II&C) Systems Technologies Pathway is part of the DOE’s Light Water Reactor Sustainability (LWRS) Program. It conducts targeted research and development (R&D) to address aging and reliability concerns with the legacy instrumentation and control and related information systems of the U.S. operating light water reactor (LWR) fleet. This work involves two major goals: (1) to ensure that legacy analog II&C systems are not life-limiting issues for the LWR fleet and (2) to implement digital II&C technology in a manner that enables broad innovation and business improvement in the NPP operating model. Resolving long-term operational concerns with the II&C systems contributes to the long-term sustainability of the LWR fleet, which is vital to the nation’s energy and environmental security. The II&C Pathway is conducting a series of pilot projects that enable the development and deployment of new II&C technologies in existing nuclear plants. Through the LWRS program, individual utilities and plants are able to participate in these projects or otherwise leverage the results of projects conducted at demonstration plants. Performance advantages of the new pilot project technologies are widely acknowledged, but it has proven difficult for utilities to derive business cases for justifying investment in these new capabilities. Lack of a business case is often cited by utilities as a barrier to pursuing wide-scale application of digital technologies to nuclear plant work activities. The decision to move forward with funding usually hinges on

  1. Coal conversion and biomass conversion: Volume 1: Final report on USAID (Agency for International Development)/GOI (Government of India) Alternate Energy Resources and Development Program

    Energy Technology Data Exchange (ETDEWEB)

    Kulkarni, A.; Saluja, J.

    1987-06-30

    The United States Agency for International Development (AID), in joint collaboration with the Government of India (GOI), supported a research and development program in Alternate Energy Resources during the period March 1983 to June 1987. The primary emphasis of this program was to develop new and advanced coal and biomass conversion technologies for the efficient utilization of coal and biomass feedstocks in India. This final ''summary'' report is divided into two volumes. This Report, Volume I, covers the program overview and coal projects and Volume II summarizes the accomplishments of the biomass projects. The six projects selected in the area of coal were: Evaluation of the Freeboard Performance in a Fluidized-Bed Combustor; Scale-up of AFBC boilers; Rheology, Stability and Combustion of Coal-Water Slurries; Beneficiation of Fine Coal in Dense Medium Cyclones; Hot Gas Cleanup and Separation; and Cold Gas Cleanup and Separation.

  2. Refined finite element modelling for the vibration analysis of large rotating machines: Application to the gas turbine modular helium reactor power conversion unit

    Science.gov (United States)

    Combescure, D.; Lazarus, A.

    2008-12-01

    This paper is aimed at presenting refined finite element modelling used for dynamic analysis of large rotating machines. The first part shows an equivalence between several levels of modelling: firstly, models made of beam elements and rigid disc with gyroscopic coupling representing the position of the rotating shaft in an inertial frame; secondly full three-dimensional (3D) or 3D shell models of the rotor and the blades represented in the rotating frame and finally two-dimensional (2D) Fourier model for both rotor and stator. Simple cases are studied to better understand the results given by analysis performed using a rotating frame and the equivalence with the standard calculations with beam elements. Complete analysis of rotating machines can be performed with models in the frames best adapted for each part of the structure. The effects of several defects are analysed and compared with this approach. In the last part of the paper, the modelling approach is applied to the analysis of the large rotating shaft part of the power conversion unit of the GT-MHR nuclear reactor.

  3. Development of a crack growth analysis is program for reactor head penetration

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sung Yull; Choi, Kwang Hee; Park, Jeong Il [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Kang, Young Hwan; Park, Sung Ho; Kim, Il; Kim, Young Jong; Yoo, Young Joon; Yoo, Wan; Maeng, Wan Young; Choi, Suk Nam; Kim, Kee Suk; Yoon, Sung Won; Kim, Jee Ho; Park, Myung Kyu [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-12-31

    Crack growth analysis program for Reactor Head Penetration is being developed for applying to plants such as, Kori 1, Kori 2, Kori 3,4 YoungKwang 1,2 and Uljin 1,2 (1) Stress Evaluation - The stress analysis is required to evaluate the structure integrity for the RVH penetration tubes. The RVH penetration tubes are geometrically non-symmetry except center one. Thus, 3D finite element analysis should be employed for the stress analysis. The magnitude and distribution of residual stress resulted from welding can be determined analytically by simulation welding procedure. (2) Flaw Evaluation - There are two objectives of the penetration tube flaw evaluation to predict the time required for a crack to propagate to the acceptance criteria. The first objective is to perform the parametric evaluation for a postulated crack. The second objective is to develop the flaw evaluation program for the crack detected during the inspection. (3) Characterization of Material Properties of Alloy 600 - These study is to provide data which similarly represent the properties of PWR power plants in Korea. The data is used for analyzing of the stress distribution around penetration tubes. And the PWSCC data will be used for the crack growth rate of the penetration tubes. (author). 92 refs., 121 figs.

  4. Synthesis and in-Situ Conversion of Phosgene in a Microchannel Reactor%微通道内光气的合成以及原位转化

    Institute of Scientific and Technical Information of China (English)

    胡万鹏; 周兴贵

    2014-01-01

    Phosgene is an important but high toxic chemical intermediate, and its safe preparation is of scientific and industrial importance. In a microchannel reactor, synthesis of phosgene was first conducted using anhydrous iron(III) chloride as catalyst and carbon tetrachloride and hydrogen peroxide as reactants, and then its in-situ conversion was achieved by reacting phosgene with ammonia water.The factors affecting the yield of phosgene such as reaction temperature, concentration of H2O2, concentration of catalyst and feeding rate were investigated. The phosgene yield reaches 93% under the optimum experimental condition. The results show the feasibility of the on-site synthesis and in-situ conversion of phosgene in a microchannel reactor, which could be an efficient method to eliminate the potential risks in production, storage, transport and utilization of phosgene. This method would be also applicable for organic compounds phosgenation reaction in a microchannel reactor.%光气(COCl2)是一种重要的但又剧毒的化工中间体,其安全生产方法学研究具有重要的科学意义和工业价值。以 FeCl3为催化剂、以四氯化碳(CCl4)与过氧化氢(H2O2)为反应原料,以微通道为反应器,以氨气和光气反应生成尿素为探针反应,实现了在微通道反应器内原位生产和利用光气。通过考察反应温度、H2O2浓度、FeCl3浓度以及进料流速对光气收率的影响,确定了反应的最佳条件(温度55.0℃,CCl4浓度0.104 mol/L,FeCl3质量分数1.0%,H2O2质量分数15.0%,总流率12.0 mL/min),在该条件下,光气的收率可达93%。本研究揭示了微通道反应器内原位合成和利用光气的可行性,为更多有机化合物的光气化提供了有益参考,而且该方法有望有效消除光气在存储、运输和使用过程中因泄露带来的潜在危险。

  5. Turning Content into Conversation: How The GLOBE Program is Growing its Brand Online

    Science.gov (United States)

    Zwerin, R.; Randolph, J. G.; Andersen, T.; Mackaro, J.; Malmberg, J.; Tessendorf, S. A.; Wegner, K.

    2012-12-01

    Social Media is now a ubiquitous way for individuals, corporations, governments and communities to communicate. However, the same does not hold quite as true for the science community as many science educators, thought leaders and science programs are either reluctant or unable to build and cultivate a meaningful social media strategy. This presentation will show how The GLOBE Program uses social media to disseminate messages, build a meaningful and engaged following and grow a brand on an international scale using a proprietary Inside-Out strategy that leverages social media platforms such as Facebook, LinkedIn, Twitter, YouTube and Blogs to significantly increase influencers on a worldwide scale. In addition, this poster presentation will be interactive, so viewers will be able to touch and feel the social experience. Moreover, GLOBE representatives will be on hand to talk viewers through how they can implement a social media strategy that will allow them to turn their content into meaningful conversation. About The GLOBE Program: GLOBE is a science and education program that connects a network of students, teachers and scientists from around the world to better understand, sustain and improve Earth's environment at local, regional and global scales. By engaging students in hands-on learning of Earth system science, GLOBE is an innovative way for teachers to get students of all ages excited about scientific discovery locally and globally. To date, more than 23 million measurements have been contributed to the GLOBE database, creating meaningful, standardized, global research-quality data sets that can be used in support of student and professional scientific research. Since beginning operations in 1995, over 58,000 trained teachers and 1.5 million students in 112 countries have participated in GLOBE. For more information or to become involved, visit www.globe.gov.

  6. Gas-cooled reactor programs: high-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1981

    Energy Technology Data Exchange (ETDEWEB)

    1982-06-01

    Information is presented concerning HTGR chemistry; fueled graphite development; irradiation services for General Atomic Company; prestressed concrete pressure vessel development; HTGR structural materials; graphite development; high-temperature reactor physics studies; shielding studies; component flow test loop studies; core support performance test; and application and project assessments.

  7. Light Water Reactor Sustainability Program Grizzly Year-End Progress Report

    Energy Technology Data Exchange (ETDEWEB)

    Benjamin Spencer; Yongfeng Zhang; Pritam Chakraborty; S. Bulent Biner; Marie Backman; Brian Wirth; Stephen Novascone; Jason Hales

    2013-09-01

    The Grizzly software application is being developed under the Light Water Reactor Sustainability (LWRS) program to address aging and material degradation issues that could potentially become an obstacle to life extension of nuclear power plants beyond 60 years of operation. Grizzly is based on INL’s MOOSE multiphysics simulation environment, and can simultaneously solve a variety of tightly coupled physics equations, and is thus a very powerful and flexible tool with a wide range of potential applications. Grizzly, the development of which was begun during fiscal year (FY) 2012, is intended to address degradation in a variety of critical structures. The reactor pressure vessel (RPV) was chosen for an initial application of this software. Because it fulfills the critical roles of housing the reactor core and providing a barrier to the release of coolant, the RPV is clearly one of the most safety-critical components of a nuclear power plant. In addition, because of its cost, size and location in the plant, replacement of this component would be prohibitively expensive, so failure of the RPV to meet acceptance criteria would likely result in the shutting down of a nuclear power plant. The current practice used to perform engineering evaluations of the susceptibility of RPVs to fracture is to use the ASME Master Fracture Toughness Curve (ASME Code Case N-631 Section III). This is used in conjunction with empirically based models that describe the evolution of this curve due to embrittlement in terms of a transition temperature shift. These models are based on an extensive database of surveillance coupons that have been irradiated in operating nuclear power plants, but this data is limited to the lifetime of the current reactor fleet. This is an important limitation when considering life extension beyond 60 years. The currently available data cannot be extrapolated with confidence further out in time because there is a potential for additional damage mechanisms (i

  8. Solid State Reactor Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Mays, G.T.

    2004-03-10

    The Solid State Reactor (SSR) is an advanced reactor concept designed to take advantage of Oak Ridge National Laboratory's (ORNL's) recently developed graphite foam that has enhanced heat transfer characteristics and excellent high-temperature mechanical properties, to provide an inherently safe, self-regulated, source of heat for power and other potential applications. This work was funded by the U.S. Department of Energy's Nuclear Energy Research Initiative (NERI) program (Project No. 99-064) from August 1999 through September 30, 2002. The initial concept of utilizing the graphite foam as a basis for developing an advanced reactor concept envisioned that a suite of reactor configurations and power levels could be developed for several different applications. The initial focus was looking at the reactor as a heat source that was scalable, independent of any heat removal/power conversion process. These applications might include conventional power generation, isotope production and destruction (actinides), and hydrogen production. Having conducted the initial research on the graphite foam and having performed the scoping parametric analyses from neutronics and thermal-hydraulic perspectives, it was necessary to focus on a particular application that would (1) demonstrate the viability of the overall concept and (2) require a reasonably structured design analysis process that would synthesize those important parameters that influence the concept the most as part of a feasible, working reactor system. Thus, the application targeted for this concept was supplying power for remote/harsh environments and a design that was easily deployable, simplistic from an operational standpoint, and utilized the new graphite foam. Specifically, a 500-kW(t) reactor concept was pursued that is naturally load following, inherently safe, optimized via neutronic studies to achieve near-zero reactivity change with burnup, and proliferation resistant. These four major areas

  9. Fast Spectrum Molten Salt Reactor Options

    Energy Technology Data Exchange (ETDEWEB)

    Gehin, Jess C [ORNL; Holcomb, David Eugene [ORNL; Flanagan, George F [ORNL; Patton, Bruce W [ORNL; Howard, Rob L [ORNL; Harrison, Thomas J [ORNL

    2011-07-01

    During 2010, fast-spectrum molten-salt reactors (FS-MSRs) were selected as a transformational reactor concept for light-water reactor (LWR)-derived heavy actinide disposition by the Department of Energy-Nuclear Energy Advanced Reactor Concepts (ARC) program and were the subject of a preliminary scoping investigation. Much of the reactor description information presented in this report derives from the preliminary studies performed for the ARC project. This report, however, has a somewhat broader scope-providing a conceptual overview of the characteristics and design options for FS-MSRs. It does not present in-depth evaluation of any FS-MSR particular characteristic, but instead provides an overview of all of the major reactor system technologies and characteristics, including the technology developments since the end of major molten salt reactor (MSR) development efforts in the 1970s. This report first presents a historical overview of the FS-MSR technology and describes the innovative characteristics of an FS-MSR. Next, it provides an overview of possible reactor configurations. The following design features/options and performance considerations are described including: (1) reactor salt options-both chloride and fluoride salts; (2) the impact of changing the carrier salt and actinide concentration on conversion ratio; (3) the conversion ratio; (4) an overview of the fuel salt chemical processing; (5) potential power cycles and hydrogen production options; and (6) overview of the performance characteristics of FS-MSRs, including general comparative metrics with LWRs. The conceptual-level evaluation includes resource sustainability, proliferation resistance, economics, and safety. The report concludes with a description of the work necessary to begin more detailed evaluation of FS-MSRs as a realistic reactor and fuel cycle option.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-08-01

    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.

  11. Light Water Reactor Sustainability Program: Analysis of Pressurized Water Reactor Station Blackout Caused by External Flooding Using the RISMC Toolkit

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mandelli, Diego [Idaho National Lab. (INL), Idaho Falls, ID (United States); Prescott, Steven [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cogliati, Joshua [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinoshita, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-08-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impact of these factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the application of a RISMC detailed demonstration case study for an emergent issue using the RAVEN and RELAP-7 tools. This case study looks at the impact of a couple of challenges to a hypothetical pressurized water reactor, including: (1) a power uprate, (2) a potential loss of off-site power followed by the possible loss of all diesel generators (i.e., a station black-out event), (3) and earthquake induces station-blackout, and (4) a potential earthquake induced tsunami flood. The analysis is performed by using a set of codes: a thermal-hydraulic code (RELAP-7), a flooding simulation tool (NEUTRINO) and a stochastic analysis tool (RAVEN) – these are currently under development at the Idaho National Laboratory.

  12. Fuel handling accident analysis for the University of Missouri Research Reactor's High Enriched Uranium to Low Enriched Uranium fuel conversion initiative

    Science.gov (United States)

    Rickman, Benjamin

    In accordance with the 1986 amendment concerning licenses for research and test reactors, the MU Research Reactor (MURR) is planning to convert from using High-Enriched Uranium (HEU) fuel to the use of Low-Enriched Uranium (LEU) fuel. Since the approval of a new LEU fuel that could meet the MURR's performance demands, the next phase of action for the fuel conversion process is to create a new Safety Analysis Report (SAR) with respect to the LEU fuel. A component of the SAR includes the Maximum Hypothetical Accident (MHA) and accidents that qualify under the class of Fuel Handling Accidents (FHA). In this work, the dose to occupational staff at the MURR is calculated for the FHAs. The radionuclide inventory for the proposed LEU fuel was calculated using the ORIGEN2 point-depletion code linked to the MURR neutron spectrum. The MURR spectrum was generated from a Monte Carlo Neutron transPort (MCNP) simulation. The coupling of these codes create MONTEBURNS, a time-dependent burnup code. The release fraction from each FHA within this analysis was established by the methodology of the 2006 HEU SAR, which was accepted by the NRC. The actual dose methodology was not recorded in the HEU SAR, so a conservative path was chosen. In compliance to NUREG 1537, when new methodology is used in a HEU to LEU analysis, it is necessary to re-evaluate the HEU accident. The Total Effective Dose Equivalent (TEDE) values were calculated in addition to the whole body dose and thyroid dose to operation personnel. The LEU FHA occupational TEDE dose was 349 mrem which is under the NRC regulatory occupational dose limit of 5 rem TEDE, and under the LEU MHA limit of 403 mrem. The re-evaluated HEU FHA occupational TEDE dose was 235 mrem, which is above the HEU MHA TEDE dose of 132 mrem. Since the new methodology produces a dose that is larger than the HEU MHA, we can safely assume that it is more conservative than the previous, unspecified dose.

  13. Influence of the flux axial form on the conversion rate and duration of cycle between recharging for ThPu and U{sub nat} fuels in CANDU reactors; Influence de la forme axiale du flux sur le taux de conversion et la duree du cycle entre rechargements pour du combustible ThPu et U{sub nat} dans les reacteurs CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Chambon, Richard [Laboratoire de Physique Subatomique et de Cosmologie, Universite Joseph Fourier / CNRS-IN2P3, 53 Avenue des Martyrs, F-38026 Grenoble (France)

    2007-01-15

    To face the increasing world power demand the world nuclear sector must be continuously updated and developed as well. Thus reactors of new types are introduced and advanced fuel cycles are proposed. The technological and economic feasibility and the transition of the present power park to a renewed park require thorough studies and scenarios, which are highly dependent on the reactor performances. The conversion rate and cycle span between recharging are important parameters in the scenarios studies. In this frame, we have studied the utilisation of thorium in the CANDU type reactors and particularly the influence of axial form of the flux, i.e. of the recharging mode, on the conversion rate and duration of the cycle between recharging. The results show that up to a first approximation the axial form of the flux resulting from the neutron transport calculations for assessing the conversion rate is not necessary to be taken into account. However the time span between recharging differs up to several percents if the axial form of the flux is taken into consideration in transport calculations. Thus if the burnup or the recharging frequency are parameters which influence significantly the deployment scenarios of a nuclear park an approach more refined than a simple transport evolution in a typical cell/assembly is recommended. Finally, the results of this study are not more general than for the assumed conditions but they give a thorough calculation method valid for any recharging/fuel combination in a CANDU type reactor.

  14. Summary of NR Program Prometheus Efforts

    Energy Technology Data Exchange (ETDEWEB)

    J Ashcroft; C Eshelman

    2006-02-08

    The Naval Reactors Program led work on the development of a reactor plant system for the Prometheus space reactor program. The work centered on a 200 kWe electric reactor plant with a 15-20 year mission applicable to nuclear electric propulsion (NEP). After a review of all reactor and energy conversion alternatives, a direct gas Brayton reactor plant was selected for further development. The work performed subsequent to this selection included preliminary nuclear reactor and reactor plant design, development of instrumentation and control techniques, modeling reactor plant operational features, development and testing of core and plant material options, and development of an overall project plan. Prior to restructuring of the program, substantial progress had been made on defining reference plant operating conditions, defining reactor mechanical, thermal and nuclear performance, understanding the capabilities and uncertainties provided by material alternatives, and planning non-nuclear and nuclear system testing. The mission requirements for the envisioned NEP missions cannot be accommodated with existing reactor technologies. Therefore concurrent design, development and testing would be needed to deliver a functional reactor system. Fuel and material performance beyond the current state of the art is needed. There is very little national infrastructure available for fast reactor nuclear testing and associated materials development and testing. Surface mission requirements may be different enough to warrant different reactor design approaches and development of a generic multi-purpose reactor requires substantial sacrifice in performance capability for each mission.

  15. Summary of NR Program Prometheus Efforts

    Science.gov (United States)

    Ashcroft, John; Eshelman, Curtis

    2007-01-01

    The Naval Reactors Program led work on the development of a reactor plant system for the Prometheus space reactor program. The work centered on a 200 kWe electric reactor plant with a 15-20 year mission applicable to nuclear electric propulsion (NEP). After a review of all reactor and energy conversion alternatives, a direct gas Brayton reactor plant was selected for further development. The work performed subsequent to this selection included preliminary nuclear reactor and reactor plant design, development of instrumentation and control techniques, modeling reactor plant operational features, development and testing of core and plant material options, and development of an overall project plan. Prior to restructuring of the program, substantial progress had been made on defining reference plant operating conditions, defining reactor mechanical, thermal and nuclear performance, understanding the capabilities and uncertainties provided by material alternatives, and planning non-nuclear and nuclear system testing. The mission requirements for the envisioned NEP missions cannot be accommodated with existing reactor technologies. Therefore concurrent design, development and testing would be needed to deliver a functional reactor system. Fuel and material performance beyond the current state of the art is needed. There is very little national infrastructure available for fast reactor nuclear testing and associated materials development and testing. Surface mission requirements may be different enough to warrant different reactor design approaches and development of a generic multi-purpose reactor requires substantial sacrifice in performance capability for each mission.

  16. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, July 1, 1980-September 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-12-12

    Objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. 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: screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850, 950 and 1050/sup 0/C. Initiation of controlled purity helium creep-rupture testing in the intensive screening test program is discussed. In addition, the results of 1000-hour exposures at 750 and 850/sup 0/C on several experimental alloys are discussed.

  17. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, April 1, 1980-June 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-11-14

    Objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. 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 and metallographic analysis for materials thermally exposed or tested at 750, 850 and 950/sup 0/C. The initiation of air creep-rupture testing in the intensive screening test program is discussed. In addition, the status of the data management system is described.

  18. Development plan for the External Hazards Experimental Group. Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Coleman, Justin Leigh [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis Lee [Idaho National Lab. (INL), Idaho Falls, ID (United States); Burns, Douglas Edward [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kammerer, Annie [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    This report describes the development plan for a new multi-partner External Hazards Experimental Group (EHEG) coordinated by Idaho National Laboratory (INL) within the Risk-Informed Safety Margin Characterization (RISMC) technical pathway of the Light Water Reactor Sustainability Program. Currently, there is limited data available for development and validation of the tools and methods being developed in the RISMC Toolkit. The EHEG is being developed to obtain high-quality, small- and large-scale experimental data validation of RISMC tools and methods in a timely and cost-effective way. The group of universities and national laboratories that will eventually form the EHEG (which is ultimately expected to include both the initial participants and other universities and national laboratories that have been identified) have the expertise and experimental capabilities needed to both obtain and compile existing data archives and perform additional seismic and flooding experiments. The data developed by EHEG will be stored in databases for use within RISMC. These databases will be used to validate the advanced external hazard tools and methods.

  19. Reactor safety research programs. Quarterly progress report, January 1--March 31, 1978

    Energy Technology Data Exchange (ETDEWEB)

    Romano, A. J. [comp.

    1978-04-01

    Progress is summarized in the following areas: (1) gas reactor safety evaluation, (2) THOR code development, (3) foreign code review, (4) SSC code development, (5) LMFBR and LWR safety experiments, (6) fast reactor safety code validation, (7) stress corrosion cracking of PWR steam generator tubing, and (8) technical coordination of structural integrity.

  20. Household-level effects of China’s Sloping Land Conversion Program under price and policy shifts

    NARCIS (Netherlands)

    Komarek, A.M.; Shi Xiaoping,; Heerink, N.B.M.

    2014-01-01

    This study examined how agricultural households involved in China's Sloping Land Conversion Program (SLCP) could respond to expected changes in environmental and livestock policies and changing commodity prices. We calibrated a farm household model using 2009 survey data collected in northeast Gansu

  1. Feasibility of sustaining the upland conversion program by establishing a domestic carbon market

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    To sustain the upland conversion program (UCP) in China after the government compensation expires, we suggest an establishment of a domestic carbon market where forest carbon from the UCP can be traded. Taking southwest China's Yunnan Province as an example, we explored the feasibility of switching the UCP to a carbon offset project. The breakeven carbon price which is equivalent to the opportunity cost of agricultural cultivation was estimated and then compared with the carbon price in the  international market. We found that it is feasible to change the UCP to a carbon offset project if the duration is longer than 10 years at a discount rate of 7%, and if the recent bid price (147.2 Yuan-t1) for Chinese carbon offset project prevails. The feasibility is better for converted land with lower productivity when the project duration is given. For a given site index, the feasibility is lower as pro-ject length is reduced. The results of sensitivity analysis show that the feasibility will be enhanced as the discount rate increases;however, the changes in the price of agricultural products and the amount of sequestered carbon have insignificant effects on the choice of sites and project duration.

  2. DESAIN TERAS PLTN JENIS PEBBLE BED MODULAR REACTOR (PBMR) MENGGUNAKAN PAKET PROGRAM MCNP-5 PADA KONDISI BEGINNING OF LIFE

    OpenAIRE

    Ralind Re Marla; Yohannes Sardjono; Supardi Supardi

    2015-01-01

    Telah dilakukan desain teras Pembangkit Listrik Tenaga Nuklir (PLTN) untuk jenis Pebble Bed Modular Reactor (PBMR) dengan daya 70 MWe untuk keperluan proses smelter pada keadaan beginning of life (BOL). Analisis ini bertujuan untuk mengetahui persen pengkayaan, distribusi suhu dan nilai keselamatan dengan koefisien reaktivitas teras yang negatif pada reaktor jenis PBMR apabila daya reaktor 70 MWe. Analisis menggunakan program Monte Carlo N-Particle-5 (MCNP5) dan dari hasil analisis ini dihara...

  3. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

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

  4. Coaching Conversations in Early Childhood Programs: The Contributions of Coach and Coachee

    Science.gov (United States)

    Jayaraman, Gayatri; Marvin, Christine; Knoche, Lisa; Bainter, Sue

    2015-01-01

    Studies to date have linked early childhood (EC) coaching to child, family, and teacher outcomes but have not investigated "what" is happening in a coaching conversation. This exploratory study specifically unpacks nuances associated with the coaching conversation process and associations between the EC coaches' behaviors and coachees'…

  5. Reactor Materials Program electrochemical potential measurements by ORNL with unirradiated and irradiated stainless steel specimens

    Energy Technology Data Exchange (ETDEWEB)

    Baumann, E.W.; Caskey, G.R. Jr.

    1993-07-01

    Effect of irradiation of stainless steel on electrochemical potential (ECP) was investigated by measurements in dilute HNO{sub 3} and H{sub 2}O{sub 2} solutions, conditions simulating reactor moderator. The electrodes were made from unirradiated/irradiated, unsensitized/sensitized specimens from R-reactor piping. Results were inconclusive because of budgetary restrictions. The dose rate may have been too small to produce a significant radiolytic effect. Neither the earlier CERT corrosion susceptibility tests nor the present ECP measurements showed a pronounced effect of irradiation on susceptibility of the stainless steel to IGSCC; this is confirmed by the absence in the stainless steel of the SRS reactor tanks (except for the C Reactor tank knuckle area).

  6. Molten-salt reactor program. Semiannual progress report for period ending February 29, 1976

    Energy Technology Data Exchange (ETDEWEB)

    McNeese, L.E.

    1976-08-01

    Separate abstracts and indexing were prepared for sections dealing with MSBR design and development; chemistry of fuel-salt and coolant-salt systems and analytical methods; materials development; fuel processing for molten-salt reactors; and salt production. (DG)

  7. Thermophilic (55°C) conversion of methanol in methanogenic-UASB reactors: influence of sulphate on methanol degradation and competition

    NARCIS (Netherlands)

    Paulo, P.L.; Vallero, M.V.G.; Trevino, R.H.M.; Lettinga, G.; Lens, P.N.L.

    2004-01-01

    Two upflow sludge bed reactors (UASB) were operated for 80 days at 55 degreesC with methanol as the substrate with an organic loading rate (OLR) of about 20 g COD l(-1) per day and a hydraulic retention time (HRT) of 10 h. One UASB was operated without sulphate addition (control reactor-R1) whereas

  8. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

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

  9. Multi-objective Programming and Social Welfare Analysis of Rural-urban Land Conversion Decision-making

    Institute of Scientific and Technical Information of China (English)

    Song Min; Zhang Anlu

    2008-01-01

    Rural-urban land conversion is an inevitable phenomenon in urbanization and industrialization. And the decision-making issue about this conversion is multi-objective because the social decision maker (the whole of central government and local authority) has to integrate the requirements of different interest groups (rural collective economic organizations, peasants, urban land users and the ones affected indirectly) and harmonize the sub-objects (economic, social and ecological outcomes) of this land allocation process. This paper established a multi-objective programming model for rural-urban land conversion decision-making and made some social welfare analysis correspondingly. Result shows that the general object of rural-urban land conversion decision-making is to reach the optimal level of social welfare in a certain state of resources allocation, while the preference of social decision makers and the value judgment of interest groups are two crucial factors which determine the realization of the rural-urban land conversion decision-making objects.

  10. PLUG: A FORTRAN program for the analysis of PLUG flow reactors with gas-phase and surface chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Larson, R.S.

    1996-01-01

    This manual describes the structure and usage of the computer program PLUG, which simulates the behavior of plug flow chemical reactors. More specifically, the code is designed to model the non-dispersive one-dimensional flow of a chemically reacting ideal gas mixture in a conduit of essentially arbitrary geometry. The code makes use of the CHEMKIN and SURFACE CHEMKIN software packages to handle gas-phase and heterogeneous kinetics as well as thermodynamic properties. In addition, the standard implicit code DASSL is used to solve the set of differential/algebraic equations describing the reactor. These equations are briefly discussed here, after which the procedures for running PLUG are described in some detail. Input and output files for a sample problem involving chemical vapor deposition are given.

  11. (I) A Declarative Framework for ERP Systems(II) Reactors: A Data-Driven Programming Model for Distributed Applications

    DEFF Research Database (Denmark)

    Stefansen, Christian Oskar Erik

    This dissertation is a collection of six adapted research papers pertaining to two areas of research. (I) A Declarative Framework for ERP Systems: • POETS: Process-Oriented Event-driven Transaction Systems. The paper describes an ontological analysis of a small segment of the enterprise domain....... • Using Soft Constraints to Guide Users in Flexible Business Process Management Systems. The paper shows how the inability of a process language to express soft constraints—constraints that can be violated occasionally, but are closely monitored—leads to a loss of intentional information in process...... on the idea of soft constraints the paper explains the design, semantics, and use of a language for allocating work in business processes. The language lets process designers express both hard constraints and soft constraints. (II) The Reactors programming model: • Reactors: A Data-Oriented Synchronous...

  12. (I) A Declarative Framework for ERP Systems(II) Reactors: A Data-Driven Programming Model for Distributed Applications

    DEFF Research Database (Denmark)

    Stefansen, Christian Oskar Erik

    on the idea of soft constraints the paper explains the design, semantics, and use of a language for allocating work in business processes. The language lets process designers express both hard constraints and soft constraints. (II) The Reactors programming model: • Reactors: A Data-Oriented Synchronous......, namely the general ledger and accounts receivable. The result is an event-based approach to designing ERP systems and an abstract-level sketch of the architecture. • Compositional Specification of Commercial Contracts. The paper describes the design, multiple semantics, and use of a domain......-specific language (DSL) for modeling commercial contracts. • SMAWL: A SMAll Workflow Language Based on CCS. The paper shows how workflow patterns can be encoded in CCS and proceeds to design a macro language, SMAWL, for workflows based on those patterns. The semantics of SMAWL is defined via translation to CCS...

  13. Fischer-Tropsch Synthesis: Influence of CO Conversion on Selectivities H2/CO Usage Ratios and Catalyst Stability for a 0.27 percent Ru 25 percent Co/Al2O3 using a Slurry Phase Reactor

    Energy Technology Data Exchange (ETDEWEB)

    W Ma; G Jacobs; Y Ji; T Bhatelia; D Bukur; S Khalid; B Davis

    2011-12-31

    The effect of CO conversion on hydrocarbon selectivities (i.e., CH{sub 4}, C{sub 5+}, olefin and paraffin), H{sub 2}/CO usage ratios, CO{sub 2} selectivity, and catalyst stability over a wide range of CO conversion (12-94%) on 0.27%Ru-25%Co/Al{sub 2}O{sub 3} catalyst was studied under the conditions of 220 C, 1.5 MPa, H{sub 2}/CO feed ratio of 2.1 and gas space velocities of 0.3-15 NL/g-cat/h in a 1-L continuously stirred tank reactor (CSTR). Catalyst samples were withdrawn from the CSTR at different CO conversion levels, and Co phases (Co, CoO) in the slurry samples were characterized by XANES, and in the case of the fresh catalysts, EXAFS as well. Ru was responsible for increasing the extent of Co reduction, thus boosting the active site density. At 1%Ru loading, EXAFS indicates that coordination of Ru at the atomic level was virtually solely with Co. It was found that the selectivities to CH{sub 4}, C{sub 5+}, and CO{sub 2} on the Co catalyst are functions of CO conversion. At high CO conversions, i.e. above 80%, CH{sub 4} selectivity experienced a change in the trend, and began to increase, and CO{sub 2} selectivity experienced a rapid increase. H{sub 2}/CO usage ratio and olefin content were found to decrease with increasing CO conversion in the range of 12-94%. The observed results are consistent with water reoxidation of Co during FTS at high conversion. XANES spectroscopy of used catalyst samples displayed spectra consistent with the presence of more CoO at higher CO conversion levels.

  14. Investigation of neutronic and safety parameters variation in 5 MW research reactor due to U{sub 3}O{sub 8}Al fuel conversion to ThO{sub 2} + U{sub 3}O{sub 8}Al

    Energy Technology Data Exchange (ETDEWEB)

    Gholamzadeh, Zohreh; Alipoor, Zahra; Mirvakili, Seyed Mohammad [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of). Reactor Research School; Feghhi, Seyed Amir Hossein [Shahid Beheshti Univ., Tehran (Iran, Islamic Republic of). Dept. of Radiation Application; Vahedi, M.; Bagheri, H. [Univ. of Science and Technology, Tehran (Iran, Islamic Republic of). Energy Dept.; Tenreiro, Claudio [Talca Univ. (Chile). Energy Dept.

    2017-05-15

    Thorium-based fuels could comprise several advantages and are being investigated as a potentially competitive option with uranium-based fuels for research or power reactors. The present study investigates computationally the application of two different thorium-based fuels in a research reactor. Void and temperature reactivity coefficients, safety factor, power peaking factor, neutron generation time, effective delayed neutron fraction and {sup 135}Xe worth parameter were investigated for the fuel conversions. The results showed both the investigated fuels would not significantly disturb neutronic and safety parameters of the modeled core in comparison with its routine fuel loading. However, 235-enriched thorium based fuel concluded in noticeably reduction of High Level Waste (HLW) but 233-enriched type could be taken in attention because of its longer fuel cycle (∝15 %) and integrated neutron flux (∝23 %).

  15. Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

    Energy Technology Data Exchange (ETDEWEB)

    Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J.; Slater, C.O.

    1982-01-01

    During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor.

  16. Gas-Cooled Thermal Reactor Program. Semiannual technical progress report, April 1, 1983-September 30, 1983

    Energy Technology Data Exchange (ETDEWEB)

    1983-12-01

    An assessment of the HTGR opportunities from the year 2000 through 2045 was the principal activity on the Market Definition Task (WBS 03). Within the Plant Technology (WBS 13) task, there were activities to develop analytical methods for investigation of Coolant Transport Behavior and to define methods and criteria for High Temperature Structural Engineering design. The activities in support of the HTGR-SC/C Lead Plant (WBS 30 and 31) were the participation in the Lead Plant System Engineering (LPSE) effort and the plant simulation task. The efforts on the Advanced HTGR systems was performed under the Modular Reactor Systems (MRS) (WBS 41) to study the potential for multiple small reactors to provide lower costs, improved safety, and higher availability than the large monolithic core reactors.

  17. Policies and practices pertaining to the selection, qualification requirements, and training programs for nuclear-reactor operating personnel at the Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Culbert, W.H.

    1985-10-01

    This document describes the policies and practices of the Oak Ridge National Laboratory (ORNL) regarding the selection of and training requirements for reactor operating personnel at the Laboratory's nuclear-reactor facilities. The training programs, both for initial certification and for requalification, are described and provide the guidelines for ensuring that ORNL's research reactors are operated in a safe and reliable manner by qualified personnel. This document gives an overview of the reactor facilities and addresses the various qualifications, training, testing, and requalification requirements stipulated in DOE Order 5480.1A, Chapter VI (Safety of DOE-Owned Reactors); it is intended to be in compliance with this DOE Order, as applicable to ORNL facilities. Included also are examples of the documentation maintained amenable for audit.

  18. Reactor mass flow data base prepared for the nonproliferation alternative systems assessment program

    Energy Technology Data Exchange (ETDEWEB)

    Primm III, R.T.C

    1981-02-01

    This report presents charge and discharge mass flow data for reactors judged to have received sufficient technical development to enable them to be demonstrated or commercially available by the year 2000. Brief descriptions of the reactors and fuel cycles evaluated are presented. A discussion of the neutronics methods used to produce the mass flow data is provided. Detailed charge and discharge fuel isotopics are presented. U/sub 3/O/sub 8/, separative work, and fissile material requirements are computed and provided for each fuel cycle.

  19. Considerations on a critical experiment program for a large fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1980-09-15

    The design studies for large LMFBR in Japan are being continued by PNC and Electric Companies. Both of them have adopted loop-type and 1000 MWe class reactor as the reference design, but main parameters of the core have not yet been fixed. The main core parameters of the present design are shown. Comparing the geometrical properties with those in ZPPR-9 and 10, some of the degrees of mockup are not satisfactory. In addition, there is another difference between a reactor and a mockup critical assembly. Therefore extrapolation is important to apply the results of JUPITER to the core design.

  20. LOCA simulation in the national research universal reactor program: postirradiation examination results for the third materials experiment (MT-3)

    Energy Technology Data Exchange (ETDEWEB)

    Rausch, W.N.

    1984-04-01

    A series of in-reactor experiments were conducted using full-length 32-rod pressurized water reactor (PWR) fuel bundles as part of the Loss-of-Coolant Accident (LOCA) Simulation Program. The third materials experiment (MT-3) was the sixth in the series of thermal-hydraulic and materials deformation/rutpure experiments conducted in the National Research Universal (NRU) reactor, Chalk River, Ontario, Canada. The main objective of the experiment was to evaluate ballooning and rupture during active two-phase cooling in the temperature range from 1400 to 1500/sup 0/F (1030 to 1090 K). The 12 test rods in the center of the 32-rod bundle were initially pressurized to 550 psi (3.8 MPa) to insure rupture in the correct temperature range. All 12 of the rods ruptured, with an average peak bundle strain of approx. 55%. The UKAEA also funded destructive postirradiation examination (PIE) of several of the ruptured rods from the MT-3 experiment. This report describes the work performed and presents the PIE results. Information obtained during the PIE included cladding thickness measurements metallography, and particle size analysis of the cracked and broken fuel pellets.

  1. Development of core design and analysis technology for integral reactor; development of coolant activity and dose evaluation program

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Chang Sun; Kim, Byeong Soo; Go, Hyun Seok; Lee, Young Wook; Jang, Mee [Seoul National University, Seoul (Korea)

    2002-03-01

    SMART, small- medium-sized integral reactor, is different from the customary electricity-generation PWR in design concepts and structures. The conventional coolant activity evaluation codes used in customary PWRs cannot be applied to SMART. In this study, SAEP(Specific Activity Evaluation Program) is developed that can be applied to both customary PWR and advanced reactor such as SMART. SAEP uses three methods(SAEP Ver.02, Ver.05, Ver.06) to evaluate coolant activity. They solve inhomogeneous linearly-coupled differential equations generated by considering nuclear system as N sub-components. Coolant activities of customary PWR are evaluated by use of SAEP. The results show good agreement with FSAR data. SAEP is used to evaluate coolant activities for SMART and the results are proposed in this study. These results show that SAEP is able to perform coolant activity evaluation for both customary PWR and advanced reactor such as SMART. In addition, with respect to radiation shielding optimization, conventional optimization methods and their characteristics related to radiation shielding are reviewed and analyzed. Strategies for proper usage of conventional methods are proposed to agree with the shielding design cases. 30 refs., 25 figs., 6 tabs. (Author)

  2. Development of core design and analysis technology for integral reactor; development of coolant activity and dose evaluation program

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Chang Sun; Kim, Byeong Soo; Go, Hyun Seok; Lee, Young Wook; Jang, Mee [Seoul National University, Seoul (Korea)

    2002-03-01

    SMART, small- medium-sized integral reactor, is different from the customary electricity-generation PWR in design concepts and structures. The conventional coolant activity evaluation codes used in customary PWRs cannot be applied to SMART. In this study, SAEP(Specific Activity Evaluation Program) is developed that can be applied to both customary PWR and advanced reactor such as SMART. SAEP uses three methods(SAEP Ver.02, Ver.05, Ver.06) to evaluate coolant activity. They solve inhomogeneous linearly-coupled differential equations generated by considering nuclear system as N sub-components. Coolant activities of customary PWR are evaluated by use of SAEP. The results show good agreement with FSAR data. SAEP is used to evaluate coolant activities for SMART and the results are proposed in this study. These results show that SAEP is able to perform coolant activity evaluation for both customary PWR and advanced reactor such as SMART. In addition, with respect to radiation shielding optimization, conventional optimization methods and their characteristics related to radiation shielding are reviewed and analyzed. Strategies for proper usage of conventional methods are proposed to agree with the shielding design cases. 30 refs., 25 figs., 6 tabs. (Author)

  3. Tokamak Fusion Test Reactor. Final conceptual design report. [Overall cost and scheduling program

    Energy Technology Data Exchange (ETDEWEB)

    1976-02-01

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

  4. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, October 1, 1979-December 31, 1979

    Energy Technology Data Exchange (ETDEWEB)

    1980-04-18

    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.

  5. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, January 1, 1978--March 31, 1978

    Energy Technology Data Exchange (ETDEWEB)

    1978-06-26

    The activities associated with the procurement of the materials for the screening test program, information from vendor certification for the materials received, and preliminary information from the materials characterization tests performed by GE are reported. The construction status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment are discussed. The final recommended impurity levels for the screening phase helium are presented and the rational behind this gas chemistry is discussed. The status of the data management system is presented.

  6. AN ANALYSIS OF STUDENTS’ SPEAKING ABILITY IN ENGLISH CONVERSATION CLUB (ECC PROGRAM AT THE 3rd SEMESTER

    Directory of Open Access Journals (Sweden)

    M. Fadhly Farhy Abbas

    2017-07-01

    Full Text Available This research is based on students’ speaking ability who had followed the EnglishConversation Club (ECC program especially for the third semester of English Department. Thepurpose of this study was to analyze the students’ speaking ability at the 3rd semester in the EnglishConversation Club FKIP UNILAK Pekanbaru. The type of the research was mixed method in typeof explanatory design. The number of participant was 53 students. The researcher used twoinstruments, those were test and interview. In analyzing the data, it used in descriptive statistics.The result of the analysis showed that the average score of 3rd semester students’ speaking abilitywas 45.42. It can be concluded that the students’ speaking ability was categorized into failed . Thescore of Standard Deviation was 7.02, Variance was 49.30, and Range was 36 points . It meansthat the students’ speaking ability was homogeneous. According to the Z-Score, it can be seen that49.06% students’ ability was higher than average and 50.94% students ability was below theaverage. In conclusion, the students’ ability in learning speaking English was failed, it had beenaffected by some factors, those were lack of vocabulary, grammar and motivation. It was supportedby the interview, eventhough the students’ perception to English Conversation Club (ECCprogram was positive, but in fact, the students’ frequency to speak English was seldom, they wereless practice speaking English everyday.Keywords : Speaking , English Conversation Club (ECC

  7. DESAIN TERAS DAN BAHAN BAKAR PLTN JENIS PEBBLE BED MODULAR REACTOR (PBMR DENGAN MENGGUNAKAN PROGRAM SRAC

    Directory of Open Access Journals (Sweden)

    Sungkowo Wahyu Santoso

    2015-03-01

    Full Text Available Analisis desain down scale teras dan bahan bakar PBMR-HTR dengan menggunakan program SRAC bertujuan mengetahui pengaruh variasi pengayaan U235, burnable poison, laju aliran pendingin dan suhu pendingin masuk terhadap kekritisan teras serta aspek-aspek keselamatan reaktor nuklir dengan parameter nilai keff dan koefisien reaktivitas suhu bahan bakar, moderator dan pendingin. Teras PBMR-HTR berbentuk silinder finite dengan lubang ditengahnya yang berisi 334.000 bahan bakar pebble bed. Bahan bakar berupa UO2, moderator grafit dan pendingin helium. Model desain down scale dilakukan pada ½ teras yang mewakili keseluruhan teras. Penelitian dilakukan dengan memvariasikan pengayaan bahan bakar sebesar 8%, 8,5%, 9%, 9,5% dan 10% sementara variasi konsentrasi burnable poison sebesar 5 ppm, 7 ppm, 9 ppm, 11 ppm, dan 15 ppm. Variasi laju aliran pendingin sebesar 60%, 80%, 100%, 120%, dan 140% sementara variasi suhu masukan pendingin sebesar 673,15K; 723,15K; 773,15K; 823,15K dan 873,15K. Pada penelitian ini keff pada BOL tanpa Gd2O3 sebesar 1.026213 dan EOL sebesar 0.995865 dengan excess reactivity sebesar 2,5 % dengan pengkayaan U235 9%. Sementara keff pada BOL dengan menggunakan Gd2O3 sebesar 1.0069680 dan EOL sebesar 0.9961928 dengan excess reactivity sebesar 0.69 % dengan konsentrasi Gd2O3 7 ppm. Koefisien reaktivitas suhu bahan bakar,moderator dan pendingin berturut-turut sebesar -9,074583E-05/K, -2,971833E-05/K dan 1,120700E-05/K. Koefisien reaktivitas bernilai negatif menunjukkan karakteristik keselamatan melekat (inherent safety telah terpenuhi. Peningkatan suhu masukan dan penurunan laju aliran pendingin berkontribusi menurunkan nilai keff teras sehingga koefisien reaktivitas bernilai negatif. Kata kunci : PBMR-HTR, kritikalitas, reaktivitas, down scale, burnable poison Core and fuel down scale analysis on PBMR-HTR using SRAC program aims to identify the influence of U235 enrichment, burnable poison, coolant flow rate and coolant temperature

  8. State and Countermeasures for Farmland Conversion to forestland Program in Fugong County%福贡县山城绿化发展现状与对策

    Institute of Scientific and Technical Information of China (English)

    黄泽胜

    2012-01-01

    文章分析了福贡县城郊山地绿化现状。为增加绿量,从政策、技术层面提出有关对策和措施。%Farmland conversion to forestland program acts as an effective means to prevent soil and water loss, protect ecological environment, promote the local economic development and help farmers off poverty and into well - off life. The paper analyzed the implementation state, benefits achieved and existing difficulties facing the farmland conversion to forestland program. And then specific measures were proposed to further strengthen the implementation of the farm- land conversion to forestland program.

  9. Experimental Engineering Section semiannual progress report (excluding reactor programs), March 1, 1975--August 31, 1975

    Energy Technology Data Exchange (ETDEWEB)

    Scott, C.D.; Burtis, C.A.; Groenier, W.S.; Pitt, W.W.; Ryon, A.D.; Watson, C.D.; Watson, J.S. (comps.)

    1976-09-01

    Sections are included on centrifugal analyzer development, advanced analytical techniques, bioengineering research and development, environmental studies, chemical engineering research, controlled thermonuclear processing, and coal conversion process development. A separate abstract was prepared for each section. (JRD)

  10. Status of French reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A. [Commissariat a l`Energie Atomique, Saclay (France)

    1997-08-01

    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.

  11. QUARTERLY PROGRESS REPORT JANUARY, FEBRUARY, MARCH, 1967 REACTOR FUELS AND MATERIALS DEVELOPMENT PROGRAMS FOR FUELS AND MATERIALS BRANCH OF USAEC DEVISION OF REACTOR DEVELOPMENT AND TECHNOLOGY

    Energy Technology Data Exchange (ETDEWEB)

    Albaugh, F. W.; Bush, S. H.; Cadwell, J. J.; de Halas, D. R.; Worlton, D. C.

    1967-06-01

    Work is reported in the areas of: fast fuels oxides and nitrides; nuclear ceramics; nuclear graphite; basic swelling studies; irradiation damage to reactor metals; ATR gas loop operation and maintenance; metallic fuels; nondestructive testing research; and fast reactor dosimetry and damage analysis.

  12. The HOR HEU/LEU core conversion

    Energy Technology Data Exchange (ETDEWEB)

    Gibcus, H.P.M.; Vries, J.W. de [Department of HOR-Development, IRI, Delft University of Technology, Delft (Netherlands); Leege, P.F.A. de [Department of Reactor Physics, IRI, Delft University of Technology, Delft (Netherlands)

    1999-07-01

    On March 16, 1998 the first two LEU (low enriched uranium) fuel elements were introduced in the HOR reactor of the Interfaculty Reactor Institute (IRI). At the moment the core consists of 5 LEU and 19 HEU (highly enriched uranium) fuel elements. It is anticipated that after about 13 core reload operations the HOR will be fully converted from HEU to LEU to a so called LEU compact core consisting of about 21 LEU fuel assemblies and as many beryllium reflector elements. The HOR HEU/LEU core conversion program, progress of the step-by-step transition phase, including comparisons of calculations and measurements, as well as the impact on the utilization of the reactor are discussed. (author)

  13. Light water reactor fuel element suitable for thorium employment in a discrete seed and blanket configuration with the aim to attain conversion ratios above the range of one

    Energy Technology Data Exchange (ETDEWEB)

    Hrovat, M.F.; Grosse, K.H.; Seemann, R. [ALD Vacuum Technologies GmbH, Hanau (Germany)

    2008-07-01

    The thorium resources in the world are relatively large. According to the IAEA-NEA-publication ''Red Book'' they amount to 4.5 10E6 metric tons and are about 4 times greater than the resources of Uranium. The fuel element described in this paper could be used in light water reactor (LWR) preferably in pressurized water reactor (PWR). The seed (feed) rods contain uranium 235 as fissionable material and the blanket (breed) rods contain thorium and uranium. The thorium in the blanket rods is converted to fissionable U-233 by irradiation with thermal neutrons. The U-233 produced is a valuable fissionable material and is characterized by high revalues, where t is defined as the number of fission neutrons per absorption in fissile materials. By optimized configuration and loading of the seed- and blanket rods the thorium is converted to U-233 and the U-238 is converted to fissionable Plutonium isotopes. Consequently more fissionable material is generated than is used. The fuel cycle is also flexible. Thus U-235, Pu-239 or weapons-grade Plutonium can be used.Based on knowledge obtained in the development of fuel elements for material test reactors (MTR), high temperature reactors (HTR) and light water reactors (LWR), a new design of fuel element suitable for thorium employment in PWR is described.

  14. Purchase of Microwave Reactors for Implementation of Small-scale Microwave-accelerated Organic Chemistry Laboratory Program in Undergraduate Curriculum and Synthetic Chemistry Research at HU

    Science.gov (United States)

    2015-05-16

    Microwave-accelerated Organic Chemistry Laboratory Program in Undergraduate Curriculum and Synthetic Chemistry Research at HU The views, opinions and/or...Research Triangle Park, NC 27709-2211 Microwave laboratory facility, Undergraduate organic chemistry laboratory, Implementation/Development of...Purchase of Microwave Reactors for Implementation of Small-scale Microwave-accelerated Organic Chemistry Laboratory Program in Undergraduate

  15. LIGHT WATER REACTOR SUSTAINABILITY PROGRAM ADVANCED INSTRUMENTATION, INFORMATION, AND CONTROL SYSTEMS TECHNOLOGIES TECHNICAL PROGRAM PLAN FOR 2013

    Energy Technology Data Exchange (ETDEWEB)

    Hallbert, Bruce; Thomas, Ken

    2014-07-01

    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.

  16. Light Water Reactor Sustainability Program Advanced Instrumentation, Information, and Control Systems Technologies Technical Program Plan for FY 2016

    Energy Technology Data Exchange (ETDEWEB)

    Hallbert, Bruce Perry [Idaho National Laboratory; Thomas, Kenneth David [Idaho National Laboratory

    2015-10-01

    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.

  17. Light Water Reactor Sustainability Program Advanced Instrumentation, Information, and Control Systems Technologies Technical Program Plan for 2013

    Energy Technology Data Exchange (ETDEWEB)

    Hallbert, Bruce [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Thomas, Ken [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2014-09-01

    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.

  18. DESAIN TERAS PLTN JENIS PEBBLE BED MODULAR REACTOR (PBMR MENGGUNAKAN PAKET PROGRAM MCNP-5 PADA KONDISI BEGINNING OF LIFE

    Directory of Open Access Journals (Sweden)

    Ralind Re Marla

    2015-03-01

    Full Text Available Telah dilakukan desain teras Pembangkit Listrik Tenaga Nuklir (PLTN untuk jenis Pebble Bed Modular Reactor (PBMR dengan daya 70 MWe untuk keperluan proses smelter pada keadaan beginning of life (BOL. Analisis ini bertujuan untuk mengetahui persen pengkayaan, distribusi suhu dan nilai keselamatan dengan koefisien reaktivitas teras yang negatif pada reaktor jenis PBMR apabila daya reaktor 70 MWe. Analisis menggunakan program Monte Carlo N-Particle-5 (MCNP5 dan dari hasil analisis ini diharapkan dapat memenuhi syarat dalam mendukung program percepatan pembangunan kelistrikan batubara 10.000 MWe khususnya untuk proses smelter, yang tersebar merata di wilayah Indonesia. Hasil penelitian menunjukkan bahwa, faktor perlipatan efektif (k-eff Reaktor jenis PBMR daya 70 MWe mengalami kondisi kritis pada pengkayaan 5,626 % dengan nilai faktor perlipatan efektif 1,00031±0,00087 dan nilai koefisien reaktivitas suhu pada -10,0006 pcm/K. Dari hasil analisis daat disimpulkan bahwa reaktor jenis PBMR daya 70 MWe adalah aman.   ABSTRACT The core design of Nuclear Power Plant for Pebble Bed Modular Reactor (PBMR type with 70 MWe capacity power in Beginning of Life (BOL has been performed. The aim of this analysis, to know percent enrichment, temperature distribution and safety value by negative temperature coefficient at type PBMR if reactor power become lower equal to 70 MWe. This analysis was expected become one part of overview project development the power plant with 10.000 MWe of total capacity, spread evenly in territory of Indonesia especially to support of smelter industries. The results showed that, effective multiplication factor (keff with power 70 MWe critical condition at enrichment 5,626 %is 1,00031±0,00087, based on enrichment result, a value of the temperature coefficient reactivity is - 10,0006 pcm/K. Based on the results of these studies, it can beconcluded that the PBMR 70 MWe design is theoritically safe.

  19. FASTER Test Reactor Preconceptual Design Report

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-31

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

  20. FASTER test reactor preconceptual design report summary

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-02-29

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

  1. Assumptions and Criteria for Performing a Feasability Study of the Conversion of the High Flux Isotope Reactor Core to Use Low-Enriched Uranium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Primm, R.T., III; Ellis, R.J.; Gehin, J.C.; Moses, D.L.; Binder, J.L.; Xoubi, N. (U. of Cincinnati)

    2006-02-01

    A computational study will be initiated during fiscal year 2006 to examine the feasibility of converting the High Flux Isotope Reactor from highly enriched uranium fuel to low-enriched uranium. The study will be limited to steady-state, nominal operation, reactor physics and thermal-hydraulic analyses of a uranium-molybdenum alloy that would be substituted for the current fuel powder--U{sub 3}O{sub 8} mixed with aluminum. The purposes of this document are to (1) define the scope of studies to be conducted, (2) define the methodologies to be used to conduct the studies, (3) define the assumptions that serve as input to the methodologies, (4) provide an efficient means for communication with the Department of Energy and American research reactor operators, and (5) expedite review and commentary by those parties.

  2. Shippingport operations with the Light Water Breeder Reactor core. (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Budd, W.A. (ed.)

    1986-03-01

    This report describes the operation of the Shippingport Atomic Power Station during the LWBR (Light Water Breeder Reactor) Core lifetime. It also summarizes the plant-oriented operations during the period preceding LWBR startup, which include the defueling of The Pressurized Water Reactor Core 2 (PWR-2) and the installation of the LWBR Core, and the operations associated with the defueling of LWBR. The intent of this report is to examine LWBR experience in retrospect and present pertinent and significant aspects of LWBR operations that relate primarily to the nuclear portion of the Station. The nonnuclear portion of the Station is discussed only as it relates to overall plant operation or to unusual problems which result from the use of conventional equipment in radioactive environments. 30 refs., 69 figs., 27 tabs.

  3. Fiscal year 1999 multi-year work plan, advanced reactors transition program

    Energy Technology Data Exchange (ETDEWEB)

    Gantt, D.A.

    1998-09-17

    The Advanced Reactors Transition (ART) has two missions. One, funded by DOE-EM is to transition assigned, surplus facilities to a safe and compliant, low-cost stable, deactivated condition (requiring minimal surveillance and maintenance) pending eventual reuse or D and D. Facilities to be transitioned include the 309 Building/Plutonium Recycle Test Reactor (PRTR) and Nuclear Energy (NE) Legacy Facilities. The second mission, funded by DOE-NE, is to maintain the Fast Flux Test Facility (FFTF) and affiliated 400 Area buildings in a safe and compliant standby condition. The condition of the plant hardware, software and personnel is to be preserved in a manner not to preclude a plant restart.

  4. High temperature catalytic membrane reactors

    Energy Technology Data Exchange (ETDEWEB)

    1990-03-01

    Current state-of-the-art inorganic oxide membranes offer the potential of being modified to yield catalytic properties. The resulting modules may be configured to simultaneously induce catalytic reactions with product concentration and separation in a single processing step. Processes utilizing such catalytically active membrane reactors have the potential for dramatically increasing yield reactions which are currently limited by either thermodynamic equilibria, product inhibition, or kinetic selectivity. Examples of commercial interest include hydrogenation, dehydrogenation, partial and selective oxidation, hydrations, hydrocarbon cracking, olefin metathesis, hydroformylation, and olefin polymerization. A large portion of the most significant reactions fall into the category of high temperature, gas phase chemical and petrochemical processes. Microporous oxide membranes are well suited for these applications. A program is proposed to investigate selected model reactions of commercial interest (i.e. dehydrogenation of ethylbenzene to styrene and dehydrogenation of butane to butadiene) using a high temperature catalytic membrane reactor. Membranes will be developed, reaction dynamics characterized, and production processes developed, culminating in laboratory-scale demonstration of technical and economic feasibility. As a result, the anticipated increased yield per reactor pass economic incentives are envisioned. First, a large decrease in the temperature required to obtain high yield should be possible because of the reduced driving force requirement. Significantly higher conversion per pass implies a reduced recycle ratio, as well as reduced reactor size. Both factors result in reduced capital costs, as well as savings in cost of reactants and energy.

  5. A Monte Carlo program to calculate the exposure rate from airborne radioactive gases inside a nuclear reactor containment building.

    Science.gov (United States)

    Sherbini, S; Tamasanis, D; Sykes, J; Porter, S W

    1986-12-01

    A program was developed to calculate the exposure rate resulting from airborne gases inside a reactor containment building. The calculations were performed at the location of a wall-mounted area radiation monitor. The program uses Monte Carlo techniques and accounts for both the direct and scattered components of the radiation field at the detector. The scattered component was found to contribute about 30% of the total exposure rate at 50 keV and dropped to about 7% at 2000 keV. The results of the calculations were normalized to unit activity per unit volume of air in the containment. This allows the exposure rate readings of the area monitor to be used to estimate the airborne activity in containment in the early phases of an accident. Such estimates, coupled with containment leak rates, provide a method to obtain a release rate for use in offsite dose projection calculations.

  6. Neutron activation analysis at the Livermore pool-type reactor for the environmental research program. [Identification of trace element contaminants

    Energy Technology Data Exchange (ETDEWEB)

    Ragaini, R.C.; Heft, R.E.; Garvis, D.

    1976-07-02

    Instrumental neutron activation analysis is a technique of trace analysis using measurements of radioactivity induced in the sample by exposure to a source of neutrons. The induced activity is measured by the emitted gamma radiation. Each gamma emitter can then be identified by the energy of the photopeaks produced as the nuclide decays and by the half-life of the neutron-induced activity. A complex computer program GAMANAL has been used to accomplish the major tasks of nuclide identification and quantification. The nuclide data output from GAMANAL is processed by a second computer code NADAC, which develops elemental abundance data from disintegration rates observed. The methods are those employed at the Livermore Pool-Type Reactor in support of the environmental research trace element analysis program. Among the procedures described and discussed are sample preparation, irradiation, analysis, and application of the technique.

  7. Mars Atmospheric Conversion to Methane and Water: An Engineering Model of the Sabatier Reactor with Characterization of Ru/Al2O3 for Long Duration Use on Mars

    Science.gov (United States)

    Meier, Anne J.; Shah, Malay; Petersen, Elspeth; Hintze, Paul; Muscatello, Tony

    2017-01-01

    The Atmospheric Processing Module (APM) is a Mars In-Situ Resource Utilization (ISRU) technology designed to demonstrate conversion of the Martian atmosphere into methane and water. The Martian atmosphere consists of approximately 95 carbon dioxide (CO2) and residual argon and nitrogen. APM utilizes cryocoolers for CO2 acquisition from a simulated Martian atmosphere and pressure. The captured CO2 is sublimated and pressurized as a feedstock into the Sabatier reactor, which converts CO2 and hydrogen to methane and water. The Sabatier reaction occurs over a packed bed reactor filled with Ru/Al2O3 pellets. The long duration use of the APM system and catalyst was investigated for future scaling and failure limits. Failure of the catalyst was detected by gas chromatography and temperature sensors on the system. Following this, characterization and experimentation with the catalyst was carried out with analysis including x-ray photoelectron spectroscopy and scanning electron microscopy with elemental dispersive spectroscopy. This paper will discuss results of the catalyst performance, the overall APM Sabatier approach, as well as intrinsic catalyst considerations of the Sabatier reactor performance incorporated into a chemical model.

  8. Neutron Flux and Th-U Conversion Ratio for Graphite-Molten Salt Reactor%石墨-熔盐反应堆堆芯中子通量与钍铀转换比

    Institute of Scientific and Technical Information of China (English)

    汤晓斌; 谢芹; 姚泽恩; 陈达

    2012-01-01

    作为获国际认可的第四代核电站反应堆堆型之一的熔盐堆(Molten salt reactor,MSR),具有固有安全性高、经济性好、核资源可持续发展以及易于防止核扩散等优点.针对石墨-熔盐零功率堆的几何参数,利用蒙特卡罗计算程序MCNP5建立了物理计算模型,计算临界情况下堆芯径向、轴向中子通量及增殖区厚度与Th-U转换比(Conversion ratio,CR)的关系.结果表明,(1)石墨-熔盐零功率堆堆芯中子通量密度分布较为平坦;(2)石墨-熔盐零功率堆反射层厚度和增殖区厚度在一定范围内,CR随反射层厚度或增殖区厚度的增加而增加,当超出该范围,CR不再随反射层厚度或增殖区厚度的增加而明显增加.%The molten salt reactor (MSR) is the only one liquid-fuel reactor in six candidates of Generation IV advanced nuclear reactor, which is characterized by remarkable advantages in safety, economics and sustainable development of the fissile resource and proliferation resistance of nuclear energy. A detailed computational model using the Monte Carlo code MCNP5 is set up, in order to study about radical/axis neutron flux and the influences of the reflect thickness or blanket thickness on the conversion ratio (CR) of the Th-U fuel cycle. Main results obtained in this calculation show that: (1) The neutron flux distribution of the graphite-molten zero power reactor core is relatively smooth. (2) CR will increase with the increasing of the thickness of reflector and/or the thickness of breeding region in a certain range and when it exceeds this range CR cannot get increased significantly.

  9. Hybrid reactors. [Fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.W.

    1980-09-09

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of /sup 233/U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m/sup -2/, and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid.

  10. A program-management plan with critical-path definition for Combustion Augmentation with Thermionic Energy Conversion (CATEC)

    Science.gov (United States)

    Morris, J. F.; Merrill, O. S.; Reddy, H. K.

    1981-01-01

    Thermionic energy conversion (TEC) is discussed. In recent TEC-topping analyses, overall plant efficiency (OPE) and cost of electricity (COE) improved slightly with current capabilities and substantially with fully matured technologies. Enhanced credibility derives from proven hot-corrosion protection for TEC by silicon-carbide clads in fossil fuel combustion products. Combustion augmentation with TEC (CATEC) affords minimal cost and plant perturbation, but with smaller OPE and COE improvements than more conventional topping applications. Risk minimization as well as comparative simplicity and convenience, favor CATEC for early market penetration. A program-management plan is proposed. Inputs, characteristics, outputs and capabilities are discussed.

  11. A program-management plan with critical-path definition for Combustion Augmentation with Thermionic Energy Conversion (CATEC)

    Science.gov (United States)

    Morris, J. F.; Merrill, O. S.; Reddy, H. K.

    Thermionic energy conversion (TEC) is discussed. In recent TEC-topping analyses, overall plant efficiency (OPE) and cost of electricity (COE) improved slightly with current capabilities and substantially with fully matured technologies. Enhanced credibility derives from proven hot-corrosion protection for TEC by silicon-carbide clads in fossil fuel combustion products. Combustion augmentation with TEC (CATEC) affords minimal cost and plant perturbation, but with smaller OPE and COE improvements than more conventional topping applications. Risk minimization as well as comparative simplicity and convenience, favor CATEC for early market penetration. A program-management plan is proposed. Inputs, characteristics, outputs and capabilities are discussed.

  12. Advanced gas cooled nuclear reactor materials evaluation and development program. Selection of candidate alloys. Vol. 1. Advanced gas cooled reactor systems definition

    Energy Technology Data Exchange (ETDEWEB)

    Marvin, M.D.

    1978-10-31

    Candidate alloys for a Very High Temperature Reactor (VHTR) Nuclear Process Heal (NPH) and Direct Cycle Helium Turbine (DCHT) applications in terms of the effect of the primary coolant exposure and thermal exposure were evaluated. (FS)

  13. IAEA coordinated research program on `harmonization and validation of fast reactor thermomechanical and thermohydraulic codes using experimental data`. 1. Thermohydraulic benchmark analysis on high-cycle thermal fatigue events occurred at French fast breeder reactor Phenix

    Energy Technology Data Exchange (ETDEWEB)

    Muramatsu, Toshiharu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-06-01

    A benchmark exercise on `Tee junction of Liquid Metal Fast Reactor (LMFR) secondary circuit` was proposed by France in the scope of the said Coordinated Research Program (CRP) via International Atomic Energy Agency (IAEA). The physical phenomenon chosen here deals with the mixture of two flows of different temperature. In a LMFR, several areas of the reactor are submitted to this problem. They are often difficult to design, because of the complexity of the phenomena involved. This is one of the major problems of the LMFRs. This problem has been encountered in the Phenix reactor on the secondary loop, where defects in a tee junction zone were detected during a campaign of inspections after an operation of 90,000 hours of the reactor. The present benchmark is based on an industrial problem and deal with thermal striping phenomena. Problems on pipes induced by thermal striping phenomena have been observed in some reactors and experimental facilities coolant circuits. This report presents numerical results on thermohydraulic characteristics of the benchmark problem, carried out using a direct numerical simulation code DINUS-3 and a boundary element code BEMSET. From the analysis with both the codes, it was confirmed that the hot sodium from the small pipe rise into the cold sodium of the main pipe with thermally instabilities. Furthermore, it was indicated that the coolant mixing region including the instabilities agrees approximately with the result by eye inspections. (author)

  14. Nuclear proliferation and civilian nuclear power. Report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle description

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    The Nonproliferation Alterntive Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates.

  15. City of Rio de Janeiro's Gas Distribution Network Conversion Program; Projeto de conversao do Sistema de Distribuicao de Gas Canalizado do Rio de Janeiro

    Energy Technology Data Exchange (ETDEWEB)

    Costa Filho, Manoel A.F.; Pallottino, Joao T. [Universidade do Estado do Rio de Janeiro (UERJ), RJ (Brazil); Margalef, Jose M. [Companhia Distribuidora de Gas do Rio de Janeiro (CEG), Rio de Janeiro, RJ (Brazil)

    2004-07-01

    This paper discusses the Conversion Program of Rio de Janeiro city, from Town Gas to Natural Gas, looking for evaluating its logistic and benefits, highlighting mainly those related to consumer's safety. The Conversion Program is been implemented under Companhia Distribuidora de Gas do Rio de Janeiro (CEG) supervision, with support of Rio de Janeiro State University (UERJ) in service's quality appraisal and conversion technical procedure elaboration and review. This partnership has started in 2001 and 46% of all residential and commercial customers have already been converted up to July 2004. Conversion Program is divided into the following Phases: Design and Planning, Gas Distribution Network Renovation and Adaptation, Revision, Post-Revision, Conversion and Post-Conversion. This Program may serve as model for similar programs, to be developed in other cities where Natural Gas will be distributed, displacing the Liquefied Petroleum Gas (LPG), and it is one contribution to achieve the federal governmental goal of 12%-NG-participation within Brazilian Energetic Matrix in 2010. (author)

  16. Analysis of the Microbial Community in an Acidic Hollow-Fiber Membrane Biofilm Reactor (Hf-MBfR) Used for the Biological Conversion of Carbon Dioxide to Methane

    Science.gov (United States)

    Jeon, Byoung Seung; Choi, Okkyoung; Kim, Hyun Wook; Um, Youngsoon; Lee, Dong-Hoon; Sang, Byoung-In

    2015-01-01

    Hydrogenotrophic methanogens can use gaseous substrates, such as H2 and CO2, in CH4 production. H2 gas is used to reduce CO2. We have successfully operated a hollow-fiber membrane biofilm reactor (Hf-MBfR) for stable and continuous CH4 production from CO2 and H2. CO2 and H2 were diffused into the culture medium through the membrane without bubble formation in the Hf-MBfR, which was operated at pH 4.5–5.5 over 70 days. Focusing on the presence of hydrogenotrophic methanogens, we analyzed the structure of the microbial community in the reactor. Denaturing gradient gel electrophoresis (DGGE) was conducted with bacterial and archaeal 16S rDNA primers. Real-time qPCR was used to track changes in the community composition of methanogens over the course of operation. Finally, the microbial community and its diversity at the time of maximum CH4 production were analyzed by pyrosequencing methods. Genus Methanobacterium, related to hydrogenotrophic methanogens, dominated the microbial community, but acetate consumption by bacteria, such as unclassified Clostridium sp., restricted the development of acetoclastic methanogens in the acidic CH4 production process. The results show that acidic operation of a CH4 production reactor without any pH adjustment inhibited acetogenic growth and enriched the hydrogenotrophic methanogens, decreasing the growth of acetoclastic methanogens. PMID:26694756

  17. Analysis of the Microbial Community in an Acidic Hollow-Fiber Membrane Biofilm Reactor (Hf-MBfR Used for the Biological Conversion of Carbon Dioxide to Methane.

    Directory of Open Access Journals (Sweden)

    Hyun Chul Shin

    Full Text Available Hydrogenotrophic methanogens can use gaseous substrates, such as H2 and CO2, in CH4 production. H2 gas is used to reduce CO2. We have successfully operated a hollow-fiber membrane biofilm reactor (Hf-MBfR for stable and continuous CH4 production from CO2 and H2. CO2 and H2 were diffused into the culture medium through the membrane without bubble formation in the Hf-MBfR, which was operated at pH 4.5-5.5 over 70 days. Focusing on the presence of hydrogenotrophic methanogens, we analyzed the structure of the microbial community in the reactor. Denaturing gradient gel electrophoresis (DGGE was conducted with bacterial and archaeal 16S rDNA primers. Real-time qPCR was used to track changes in the community composition of methanogens over the course of operation. Finally, the microbial community and its diversity at the time of maximum CH4 production were analyzed by pyrosequencing methods. Genus Methanobacterium, related to hydrogenotrophic methanogens, dominated the microbial community, but acetate consumption by bacteria, such as unclassified Clostridium sp., restricted the development of acetoclastic methanogens in the acidic CH4 production process. The results show that acidic operation of a CH4 production reactor without any pH adjustment inhibited acetogenic growth and enriched the hydrogenotrophic methanogens, decreasing the growth of acetoclastic methanogens.

  18. Biogenic hydrogen conversion of de-oiled jatropha waste via anaerobic sequencing batch reactor operation: process performance, microbial insights, and CO2 reduction efficiency.

    Science.gov (United States)

    Kumar, Gopalakrishnan; Lin, Chiu-Yue

    2014-01-01

    We report the semicontinuous, direct (anaerobic sequencing batch reactor operation) hydrogen fermentation of de-oiled jatropha waste (DJW). The effect of hydraulic retention time (HRT) was studied and results show that the stable and peak hydrogen production rate of 1.48 L/L ∗ d and hydrogen yield of 8.7 mL H2/g volatile solid added were attained when the reactor was operated at HRT 2 days (d) with a DJW concentration of 200 g/L, temperature 55 °C, and pH 6.5. Reduced HRT enhanced the production performance until 1.75 d. Further reduction has lowered the process efficiency in terms of biogas production and hydrogen gas content. The effluent from hydrogen fermentor was utilized for methane fermentation in batch reactors using pig slurry and cow dung as seed sources. The results revealed that pig slurry was a feasible seed source for methane generation. Peak methane production rate of 0.43 L CH4/L ∗ d and methane yield of 20.5 mL CH4/g COD were observed at substrate concentration of 10 g COD/L, temperature 30 °C, and pH 7.0. PCR-DGGE analysis revealed that combination of cellulolytic and fermentative bacteria were present in the hydrogen producing ASBR.

  19. Biogenic Hydrogen Conversion of De-Oiled Jatropha Waste via Anaerobic Sequencing Batch Reactor Operation: Process Performance, Microbial Insights, and CO2 Reduction Efficiency

    Directory of Open Access Journals (Sweden)

    Gopalakrishnan Kumar

    2014-01-01

    Full Text Available We report the semicontinuous, direct (anaerobic sequencing batch reactor operation hydrogen fermentation of de-oiled jatropha waste (DJW. The effect of hydraulic retention time (HRT was studied and results show that the stable and peak hydrogen production rate of 1.48 L/L*d and hydrogen yield of 8.7 mL H2/g volatile solid added were attained when the reactor was operated at HRT 2 days (d with a DJW concentration of 200 g/L, temperature 55°C, and pH 6.5. Reduced HRT enhanced the production performance until 1.75 d. Further reduction has lowered the process efficiency in terms of biogas production and hydrogen gas content. The effluent from hydrogen fermentor was utilized for methane fermentation in batch reactors using pig slurry and cow dung as seed sources. The results revealed that pig slurry was a feasible seed source for methane generation. Peak methane production rate of 0.43 L CH4/L*d and methane yield of 20.5 mL CH4/g COD were observed at substrate concentration of 10 g COD/L, temperature 30°C, and pH 7.0. PCR-DGGE analysis revealed that combination of celluloytic and fermentative bacteria were present in the hydrogen producing ASBR.

  20. Liquid Metal Cooled Reactor for Space Power

    Science.gov (United States)

    Weitzberg, Abraham

    2003-01-01

    The conceptual design is for a liquid metal (LM) cooled nuclear reactor that would provide heat to a closed Brayton cycle (CBC) power conversion subsystem to provide electricity for electric propulsion thrusters and spacecraft power. The baseline power level is 100 kWe to the user. For long term power generation, UN pin fuel with Nb1Zr alloy cladding was selected. As part of the SP-100 Program this fuel demonstrated lifetime with greater than six atom percent burnup, at temperatures in the range of 1400-1500 K. The CBC subsystem was selected because of the performance and lifetime database from commercial and aircraft applications and from prior NASA and DOE space programs. The high efficiency of the CBC also allows the reactor to operate at relatively low power levels over its 15-year life, minimizing the long-term power density and temperature of the fuel. The scope of this paper is limited to only the nuclear components that provide heated helium-xenon gas to the CBC subsystem. The principal challenge for the LM reactor concept was to design the reactor core, shield and primary heat transport subsystems to meet mission requirements in a low mass configuration. The LM concept design approach was to assemble components from prior programs and, with minimum change, determine if the system met the objective of the study. All of the components are based on technologies having substantial data bases. Nuclear, thermalhydraulic, stress, and shielding analyses were performed using available computer codes. Neutronics issues included maintaining adequate operating and shutdown reactivities, even under accident conditions. Thermalhydraulic and stress analyses calculated fuel and material temperatures, coolant flows and temperatures, and thermal stresses in the fuel pins, components and structures. Using conservative design assumptions and practices, consistent with the detailed design work performed during the SP-100 Program, the mass of the reactor, shield, primary heat

  1. Light Water Reactor Sustainability Program Risk-Informed Safety Margins Characterization (RISMC) PathwayTechnical Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Curtis Smith; Cristian Rabiti; Richard Martineau

    2012-11-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). As the current Light Water Reactor (LWR) NPPs age beyond 60 years, there are possibilities for increased frequency of Systems, Structures, and Components (SSCs) degradations or failures that initiate safety-significant events, reduce existing accident mitigation capabilities, or create new failure modes. Plant designers commonly “over-design” portions of NPPs and provide robustness in the form of redundant and diverse engineered safety features to ensure that, even in the case of well-beyond design basis scenarios, public health and safety will be protected with a very high degree of assurance. This form of defense-in-depth is a reasoned response to uncertainties and is often referred to generically as “safety margin.” Historically, specific safety margin provisions have been formulated, primarily based on “engineering judgment.”

  2. Deep water pipe, pump, and mooring study: Ocean Thermal Energy Conversion program. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Little, T.E.; Marks, J.D.; Wellman, K.H.

    1976-06-01

    The ocean engineering issues affecting the design, construction, deployment, and operation of Ocean Thermal Energy Conversion (OTEC) power plants are of key importance. This study addressed the problems associated with the conceptual design of the deep-water pipe, cold-water-pumping, and platform mooring arrangements. These subsystems fall into a natural grouping since the parameters affecting their design are closely related to each other and to the ocean environment. Analysis and evaluations are provided with a view toward judging the impact of the various subsystems on the overall plant concept and to provide an estimate of material and construction cost. Parametric data is provided that describes mooring line configurations, mooring line loads, cold water pipe configurations, and cold water pumping schemes. Selected parameters, issues, and evaluation criteria are used to judge the merits of candidate concepts over a range of OTEC plant size from 100 MWe to 1000 MWe net output power.

  3. Oxidative coupling of methane using inorganic membrane reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ma, Y.H.; Moser, W.R.; Dixon, A.G. [Worcester Polytechnic Institute, MA (United States)] [and others

    1995-12-31

    The goal of this research is to improve the oxidative coupling of methane in a catalytic inorganic membrane reactor. A specific target is to achieve conversion of methane to C{sub 2} hydrocarbons at very high selectivity and relatively higher yields than in fixed bed reactors by controlling the oxygen supply through the membrane. A membrane reactor has the advantage of precisely controlling the rate of delivery of oxygen to the catalyst. This facility permits balancing the rate of oxidation and reduction of the catalyst. In addition, membrane reactors minimize the concentration of gas phase oxygen thus reducing non selective gas phase reactions, which are believed to be a main route for formation of CO{sub x} products. Such gas phase reactions are a cause for decreased selectivity in oxidative coupling of methane in conventional flow reactors. Membrane reactors could also produce higher product yields by providing better distribution of the reactant gases over the catalyst than the conventional plug flow reactors. Modeling work which aimed at predicting the observed experimental trends in porous membrane reactors was also undertaken in this research program.

  4. Review of the n_TOF experimental program for Reactor Applications

    Directory of Open Access Journals (Sweden)

    Guerrero C.

    2013-03-01

    Full Text Available The n_TOF facility at CERN is devoted mainly to the measurement of neutron-induced reaction cross section of interest for Nuclear Technologies, Astrophysics and Basic Physics. In particular, the list of measurements carried out during the 2nd Phase of experiments n_TOF-Ph2 (2009-2012 includes a significant number of capture and fission experiments on actinides which are considered key for the further development of nuclear reactors. This contribution will contain a description of all these experiments, some of which will be discussed in detail. The future of the n_TOF facility will be also addressed; in particular, the new vertical neutron beam line with a flight path of only 20 m will be presented and the expected performance discussed in detail.

  5. Modular Pebble-Bed Reactor Project: Laboratory-Directed Research and Development Program FY 2002 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David Andrew; Dolan, Thomas James; Miller, Gregory Kent; Moore, Richard Leroy; Terry, William Knox; Ougouag, Abderrafi Mohammed-El-Ami; Oh, Chang H; Gougar, Hans D

    2002-11-01

    This report documents the results of our research in FY-02 on pebble-bed reactor technology under our Laboratory Directed Research and Development (LDRD) project entitled the Modular Pebble-Bed Reactor. The MPBR is an advanced reactor concept that can meet the energy and environmental needs of future generations under DOE’s Generation IV initiative. Our work is focused in three areas: neutronics, core design and fuel cycle; reactor safety and thermal hydraulics; and fuel performance.

  6. Modular Pebble-Bed Reactor Project: Laboratory-Directed Research and Development Program FY 2002 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David Andrew; Dolan, Thomas James; Miller, Gregory Kent; Moore, Richard Leroy; Terry, William Knox; Ougouag, Abderrafi Mohammed-El-Ami; Oh, Chang H; Gougar, Hans D

    2002-11-01

    This report documents the results of our research in FY-02 on pebble-bed reactor technology under our Laboratory Directed Research and Development (LDRD) project entitled the Modular Pebble-Bed Reactor. The MPBR is an advanced reactor concept that can meet the energy and environmental needs of future generations under DOE’s Generation IV initiative. Our work is focused in three areas: neutronics, core design and fuel cycle; reactor safety and thermal hydraulics; and fuel performance.

  7. Defining the "proven technology" technical criterion in the reactor technology assessment for Malaysia's nuclear power program

    Science.gov (United States)

    Anuar, Nuraslinda; Kahar, Wan Shakirah Wan Abdul; Manan, Jamal Abdul Nasir Abd

    2015-04-01

    Developing countries that are considering the deployment of nuclear power plants (NPPs) in the near future need to perform reactor technology assessment (RTA) in order to select the most suitable reactor design. The International Atomic Energy Agency (IAEA) reported in the Common User Considerations (CUC) document that "proven technology" is one of the most important technical criteria for newcomer countries in performing the RTA. The qualitative description of five desired features for "proven technology" is relatively broad and only provides a general guideline to its characterization. This paper proposes a methodology to define the "proven technology" term according to a specific country's requirements using a three-stage evaluation process. The first evaluation stage screens the available technologies in the market against a predefined minimum Technology Readiness Level (TRL) derived as a condition based on national needs and policy objectives. The result is a list of technology options, which are then assessed in the second evaluation stage against quantitative definitions of CUC desired features for proven technology. The potential technology candidates produced from this evaluation is further narrowed down to obtain a list of proven technology candidates by assessing them against selected risk criteria and the established maximum allowable total score using a scoring matrix. The outcome of this methodology is the proven technology candidates selected using an accurate definition of "proven technology" that fulfills the policy objectives, national needs and risk, and country-specific CUC desired features of the country that performs this assessment. A simplified assessment for Malaysia is carried out to demonstrate and suggest the use of the proposed methodology. In this exercise, ABWR, AP1000, APR1400 and EPR designs assumed the top-ranks of proven technology candidates according to Malaysia's definition of "proven technology".

  8. Moon base reactor system

    Science.gov (United States)

    Chavez, H.; Flores, J.; Nguyen, M.; Carsen, K.

    1989-01-01

    The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.

  9. Case studies of fifth-grade student modeling in science through programming: Comparison of modeling practices and conversations

    Science.gov (United States)

    Louca, Loucas

    This is a descriptive case study investigating the use of two computer-based programming environments (CPEs), MicroWorlds(TM) (MW) and Stagecast Creator(TM) (SC), as modeling tools for collaborative fifth grade science learning. In this study I investigated how CPEs might support fifth grade student work and inquiry in science. There is a longstanding awareness of the need to help students learn about models and modeling in science, and CPEs are promising tools for this. A computer program can be a model of a physical system, and modeling through programming may make the process more tangible: Programming involves making decisions and assumptions; the code is used to express ideas; running the program shows the implications of those ideas. In this study I have analyzed and compared students' activities and conversations in two after-school clubs, one working with MW and the other with SC. The findings confirm the promise of CPEs as tools for teaching practices of modeling and science, and they suggest advantages and disadvantages to that purpose of particular aspects of CPE designs. MW is an open-ended, textual CPE that uses procedural programming. MW students focused on breaking down phenomena into small programmable pieces, which is useful for scientific modeling. Developing their programs, the students focused on writing, testing and debugging code, which are also useful for scientific modeling. SC is a non-linear, object-oriented CPE that uses visual program language. SC students saw their work as creating games. They were focused on the overall story which they then translated it into SC rules, which was in conflict with SC's object-oriented interface. However, telling the story of individual causal agents was useful for scientific modeling. Programming in SC was easier, whereas reading code in MW was more tangible. The latter helped MW students to use the code as the representation of the phenomenon rather than merely as a tool for creating a simulation. The

  10. Toward a Conversation between ESL Teachers and Intensive English Program Administrators

    Science.gov (United States)

    Moussu, Lucie

    2010-01-01

    Several scholars have investigated the strengths and weaknesses of native-English-speaking (NES) and nonnative-English-speaking (NNES) teachers of English as a second language (ESL), but few researchers have explored intensive English program (IEP) administrators' and ESL teachers' perspectives on teacher training and the strengths and weaknesses…

  11. Highlights of the solar thermal conversion program. Semi-annual review, 30 April-2 May 1975

    Energy Technology Data Exchange (ETDEWEB)

    Finlayson, F.C.

    1975-07-01

    Overviews of the ERDA and Electric Power Research Institute program plans and the progress in the mission analysis study by the Aerospace Corporation are given. The total energy, distributed collector, surface technology, and central receiver studies are summarized briefly. Informal oral comments, recommendations, and conclusions of the Ad Hoc Review Committee relative to the information presented at the review are compiled. (LEW)

  12. Hydrogen Production from Water by Photosynthesis System I for Use as Fuel in Energy Conversion Devices (a.k.a. Understanding Photosystem I as a Biomolecular Reactor for Energy Conversion)

    Science.gov (United States)

    2014-04-01

    Hydrogen Production from Water by Photosynthesis System I for Use as Fuel in Energy Conversion Devices (a.k.a. Understanding Photosystem I as...Laboratory Adelphi, MD 20783-1197 ARL-TR-6904 April 2014 Hydrogen Production from Water by Photosynthesis System I for Use as Fuel in Energy...Final 3. DATES COVERED (From - To) 10/1/2010–10/1/2013 4. TITLE AND SUBTITLE Hydrogen Production from Water by Photosynthesis System I for Use as Fuel

  13. First results of the irradiation program of inert matrices, targets and fuels for minor actinides transmutation in fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bonnerot, Jean-Marc; Ferroud-Plattet, Marie-Pierre; Lamontagne, Jerome [CEA Cadarache, Nuclear Energy Direction, Saint-Paul les Durance Cedex, 13108 (France); Warin, Dominique [CEA Valrho, Nuclear Energy Direction, DRCP, Bagnols-sur-Ceze Cedex, 30207 (France); Gosmain, Lionel [CEA Saclay, Nuclear Energy Direction, DMN, Gif sur Yvette, 91190 (France)

    2008-07-01

    A comprehensive irradiation program was started in France in 1992 to demonstrate the technical feasibility of the transmutation of minor actinides in current and future nuclear reactors, by means of inert support targets or dedicated fuels. The first step of the program (MATINA program) consisted in the irradiation of various inert materials intended as support matrix for transmutation targets, in the fast reactor Phenix, to select the best candidates. These inert materials included as well oxide and nitride ceramics - MgO, MgAl{sub 2}O{sub 4}, Al{sub 2}O{sub 3}, Y{sub 3}Al{sub 5}O{sub 12} and TiN - as refractory metals - W, Nb, Cr and V- and were irradiated under fast neutron flux at temperatures ranged between 650 and 1040 deg. C. The results show that in comparison to MgO, MgAl{sub 2}O{sub 4} and Al{sub 2}O{sub 3} inert matrices irradiated alone, the composite pellets containing UO{sub 2} particles, showed very different behaviors under irradiation. The swelling of MgO pellets is enhanced in the presence of fissile material whereas it is lowered for the Al{sub 2}O{sub 3}-UO{sub 2} pellets. MgAl{sub 2}O{sub 4}-UO{sub 2} pellets remained stable. The second step of the program aimed at testing the behavior of inert support targets containing americium. A new experiment ECRIX H involving composite pellets with an MgO matrix and AmO{sub 2-x} particles was performed in Phenix and completed in 2006. A rather low elongation of the pellet stack was observed and no significant diameter deformation of cladding was detected after irradiation. The analysis of the filling gas of the pin after puncturing, revealed that respectively 28% and 5% of the He and Xe+Kr created under irradiation were released in the expanding volume of the pin. ECRIX H, which is the first experiment on Am base target in Phenix, will undoubtedly represent a very important step in the general design approach about inert matrix support targets once the complete results should be available by the end of

  14. Seawater pump study: Ocean Thermal Energy Conversion Program. Final report. [For ocean thermal power plants

    Energy Technology Data Exchange (ETDEWEB)

    Little, T.E.

    1978-01-01

    The pumping power required to move cold seawater and warm seawater through an Ocean Thermal Energy Conversion (OTEC) power plant is a significant portion of the plant power output; therefore, seawater pump performance, sizing, and cost information are very influential inputs into any power plant system design optimizations. The analysis and evaluation of large seawater pumping systems selected specifically for the OTEC application are provided with a view toward judging the impact of pump selection on overall OTEC power plant performance. A self-contained bulb, direct drive, axial flow pump was found to have a distinct advantage in performance and arrangement flexibility. A design of a pump operating at a net total head rise of 3.5 meters and a flow capacity of 100 m/sup 3//s is presented including pump blade geometry (profiles), pump diffuser geometry, and pump/diffuser configuration and performance. Results are presented in terms of the geometric and power requirements of several related pump designs over a range of seawater capacity from 25 m/sup 3//s to 100 m/sup 3//s. Summary analysis and evaluations include pump design weights and cost estimates.

  15. Medicare program; physician fee schedule conversion factor for calendar year 1998 and sustainable growth rate for fiscal year 1998--HCFA. Final notice.

    Science.gov (United States)

    1997-10-31

    This final notice announces the calendar year 1998 Medicare physician fee schedule conversion factor and the fiscal year 1998 sustainable growth rate for expenditures for physicians' services under the Medicare Supplementary Medical Insurance (Part B) program as required by sections 1846(d) and (f), respectively, of the Social Security Act. The 1998 Medicare physician fee schedule conversion factor is $36,6873. The sustainable growth rate for fiscal year 1998 is 1.5 percent.

  16. Ocean Thermal Energy Conversion (OTEC) test facilities study program. Final report. Volume II. Part A

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-17

    Results are presented of an 8-month study to develop alternative non-site-specific OTEC facilities/platform requirements for an integrated OTEC Test Program which may include land and floating test facilities. The document, Volume II - Appendixes is bound in three parts (A, B, and C) which together comprise a compendium of the most significant detailed data developed during the study. Part A contains definitions, baseline revisions, test plans, and energy utilization sections.

  17. Prototype Tests for the Recovery and Conversion of UF6 Chemisorbed in NaF Traps for the Molten Salt Reactor Remediation Project

    Energy Technology Data Exchange (ETDEWEB)

    Del Cul, G.D.; Icenhour, A.S.; Simmons, D.W.

    2000-04-01

    The remediation of the Molten Salt Reactor Experiment (MSRE) site includes the removal of about 37 kg of uranium. Of that inventory, about 23 kg have already been removed from the piping system and chemisorbed in 25 NaF traps. This material is being stored in Building 3019. The planned recovery of -11 kg of uranium from the fuel salt will generate another 15 to 19 NaF traps. The remaining 2 to 3 kg of uranium are present in activated charcoal beds, which are also scheduled to be removed from the reactor site. Since all of these materials (NaF traps and the uranium-laden charcoal) are not suitable for long-term storage, they will be converted to a chemical form [uranium oxide], which is suitable for long-term storage. This document describes the process that will be used to recover and convert the uranium in the NaF traps into a stable oxide for long-term storage. Included are a description of the process, equipment, test results, and lessons learned. The process was developed for remote operation in a hot cell. Lessons learned from the prototype testing were incorporated into the process design.

  18. Prototype Tests for the Recovery and Conversion of UF6Chemisorbed in NaF Traps for the Molten Salt Reactor Remediation Project

    Energy Technology Data Exchange (ETDEWEB)

    Del Cul, G.D.

    2000-06-07

    The remediation of the Molten Salt Reactor Experiment (MSRE) site includes the removal of about 37 kg of uranium. Of that inventory, about 23 kg have already been removed from the piping system and chemisorbed in 25 NaF traps. This material is being stored in Building 3019. The planned recovery of {approx}11 kg of uranium from the fuel salt will generate another 15 to 19 NaF traps. The remaining 2 to 3 kg of uranium are present in activated charcoal beds, which are also scheduled to be removed from the reactor site. Since all of these materials (NaF traps and the uranium-laden charcoal) are not suitable for long-term storage, they will be converted to a chemical form [uranium oxide (U{sub 3}O{sub 8})], which is suitable for long-term storage. This document describes the process that will be used to recover and convert the uranium in the NaF traps into a stable oxide for long-term storage. Included are a description of the process, equipment, test results, and lessons learned. The process was developed for remote operation in a hot cell. Lessons learned from the prototype testing were incorporated into the process design.

  19. Application programming interface document for the modernized Transient Reactor Analysis Code (TRAC-M)

    Energy Technology Data Exchange (ETDEWEB)

    Mahaffy, J. [Pennsylvania State Univ., University Park, PA (United States); Boyack, B.E.; Steinke, R.G. [Los Alamos National Lab., NM (United States)

    1998-05-01

    The objective of this document is to ease the task of adding new system components to the Transient Reactor Analysis Code (TRAC) or altering old ones. Sufficient information is provided to permit replacement or modification of physical models and correlations. Within TRAC, information is passed at two levels. At the upper level, information is passed by system-wide and component-specific data modules at and above the level of component subroutines. At the lower level, information is passed through a combination of module-based data structures and argument lists. This document describes the basic mechanics involved in the flow of information within the code. The discussion of interfaces in the body of this document has been kept to a general level to highlight key considerations. The appendices cover instructions for obtaining a detailed list of variables used to communicate in each subprogram, definitions and locations of key variables, and proposed improvements to intercomponent interfaces that are not available in the first level of code modernization.

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

    Energy Technology Data Exchange (ETDEWEB)

    1992-08-01

    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.

  1. Implications of Results from the Advanced Gas Reactor Fuel Development and Qualification Program on Licensing of Modular HTGRs

    Energy Technology Data Exchange (ETDEWEB)

    David Petti

    2001-10-01

    The high level of safety of modular high temperature gas-cooled reactor (HTGR) designs is achieved by passively maintaining core temperatures below fission product release thresholds under all envisioned accident scenarios. This level of fuel performance and fission product retention reduces the radioactive source term by many orders of magnitude relative to other reactor types but is predicated on exceptionally high coated-particle fuel fabrication quality and excellent fuel performance under normal operation and accident conditions. The Advanced Gas Reactor Fuel Development and Qualification (AGR) Program decided to qualify for uranium oxide/uranium carbide (UCO) TRISO coated-particle fuel in an operating envelope that would bound both pebble bed and prismatic modular HTGR options. By using a mixture of uranium oxide and uranium carbide, the kernel composition is engineered to minimize CO formation and fuel kernel migration, which is key to maintain to fuel integrity at the higher burnups, temperatures, and temperature gradients anticipated in prismatic HTGRs. Fuel fabrication conducted at both laboratory and engineering scale has demonstrated the ability to fabricate high quality UCO TRISO fuel with very low defects. The first irradiation (AGR 1) exposed about 300,000 TRISO fuel particles to a peak burnup of 19.6% FIMA, a peak fast-neutron fluence of about 4.3 × 1025 n/m2, and a maximum time-averaged fuel temperature of about 1,200°C without a single particle failure. The very low release of key metallic fission products (except silver) measured in post-irradiation examination (PIE) confirms the excellent performance measured under irradiation. Very low releases have been measured in accident simulation heatup testing (''safety testing'') after hundreds of hours at 1600 and 1700°C and no particle failures (no noble gas release measured) have been observed. Even after hundreds of hours at 1800°C, the releases are

  2. Uranium Conversion & Enrichment

    Energy Technology Data Exchange (ETDEWEB)

    Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-06

    The isotopes of uranium that are found in nature, and hence in ‘fresh’ Yellowcake’, are not in relative proportions that are suitable for power or weapons applications. The goal of conversion then is to transform the U3O8 yellowcake into UF6. Conversion and enrichment of uranium is usually required to obtain material with enough 235U to be usable as fuel in a reactor or weapon. The cost, size, and complexity of practical conversion and enrichment facilities aid in nonproliferation by design.

  3. Ocean Thermal Energy Conversion (OTEC) test facilities study program. Final report. Volume I

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-01-17

    A comprehensive test program has been envisioned by ERDA to accomplish the OTEC program objectives of developing an industrial and technological base that will lead to the commercial capability to successfully construct and economically operate OTEC plants. This study was performed to develop alternative non-site specific OTEC test facilities/platform requirements for an integrated OTEC test program including both land and floating test facilities. A progression of tests was established in which OTEC power cycle component designs proceed through advanced research and technology, component, and systems test phases. This progression leads to the first OTEC pilot plant and provides support for following developments which potentially reduce the cost of OTEC energy. It also includes provisions for feedback of results from all test phases to enhance modifications to existing designs or development of new concepts. The tests described should be considered as representative of generic types since specifics can be expected to change as the OTEC plant design evolves. Emphasis is placed on defining the test facility which is capable of supporting the spectrum of tests envisioned. All test support facilities and equipment have been identified and included in terms of space, utilities, cost, schedule, and constraints or risks. A highly integrated data acquisition and control system has been included to improve test operations and facility effectiveness through a centralized computer system capable of automatic test control, real-time data analysis, engineering analyses, and selected facility control including safety alarms. Electrical power, hydrogen, and ammonia are shown to be technically feasible as means for transmitting OTEC power to a land-based distribution point. (WHK)

  4. Where program and fantasy meet: Female fans conversing with character bots in Japan [symposium

    Directory of Open Access Journals (Sweden)

    Keiko Nishimura

    2013-03-01

    Full Text Available Character bots are automated programs that post—that is, tweet—characters' lines from popular manga, anime, games, and so on. They post regularly, and in the past few years they have become difficult to ignore, especially in fan communities. Many fans take great pleasure in interacting with favorite characters as bots; rhey also enjoy the communities that spring up around favorite series, characters, and, yes, even bots. Here I adopt an ethnographic approach to analyze the human dimensions of the phenomenon of character bots, based on participant observation among female fans in Japan.

  5. Light Water Reactor Sustainability Program: Computer-Based Procedures for Field Activities: Results from Three Evaluations at Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Oxstrand, Johanna [Idaho National Lab. (INL), Idaho Falls, ID (United States); Le Blanc, Katya [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bly, Aaron [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    The Computer-Based Procedure (CBP) research effort is a part of the Light-Water Reactor Sustainability (LWRS) Program, which is a research and development (R&D) program sponsored by Department of Energy (DOE) and performed in close collaboration with industry R&D programs that provides the technical foundations for licensing and managing the long-term, safe, and economical operation of current nuclear power plants. One of the primary missions of the LWRS program is to help the U.S. nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. One area that could yield tremendous savings in increased efficiency and safety is in improving procedure use. Nearly all activities in the nuclear power industry are guided by procedures, which today are printed and executed on paper. This paper-based procedure process has proven to ensure safety; however, there are improvements to be gained. Due to its inherent dynamic nature, a CBP provides the opportunity to incorporate context driven job aids, such as drawings, photos, and just-in-time training. Compared to the static state of paper-based procedures (PBPs), the presentation of information in CBPs can be much more flexible and tailored to the task, actual plant condition, and operation mode. The dynamic presentation of the procedure will guide the user down the path of relevant steps, thus minimizing time spent by the field worker to evaluate plant conditions and decisions related to the applicability of each step. This dynamic presentation of the procedure also minimizes the risk of conducting steps out of order and/or incorrectly assessed applicability of steps.

  6. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, April 1, 1982-June 30, 1982

    Energy Technology Data Exchange (ETDEWEB)

    1982-10-01

    Work covered in this report includes the activities associated with the status of the simulated reactor helium supply systems and testing equipment. The progress in the screening test program is described; this includes screening creep results and metallographic analyses for materials thermally exposed or tested at 750/sup 0/, 850/sup 0/, 950/sup 0/ and 1050/sup 0/C (1382/sup 0/, 1562/sup 0/, 1742/sup 0/ and 1922/sup 0/F) in controlled-purity helium. The status of creep-rupture in controlled-purity helium and air and fatigue testing in controlled-purity helium and air in the intensive screening test program is discussed. The results of metallographic studies of corrosion pins exposed at 750/sup 0/C for 6000 hours in controlled-purity helium (solid solution strengthened alloys, centrifugally cast alloys and an iron-base oxide dispersion strengthened alloy) are presented and discussed. The results of metallographic studies on the same materials after 10,000 hour exposure in controlled-purity helium at 850/sup 0/C are also presented.

  7. Studies on a membrane reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mohan, K.; Govind, R.

    1988-10-01

    Simulation is used to evaluate the performance of a catalytic reactor with permeable wall (membrane reactor) in shifting the equilibrium of three reversible reactions (cyclohexane dehydrogenation, hydrogen iodide decomposition, and propylene disproportionation). It is found that the preferred choice of cocurrernt or countercurrent operation is dependent on the physical properties and operating conditions. Methods of enhancing conversion are suggested and temperature effects are discussed.

  8. Light Water Reactor Sustainability Program: Evaluation of Localized Cable Test Methods for Nuclear Power Plant Cable Aging Management Programs

    Energy Technology Data Exchange (ETDEWEB)

    Glass, Samuel W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fifield, Leonard S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hartman, Trenton S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-05-30

    This Pacific Northwest National Laboratory (PNNL) milestone report describes progress to date on the investigation of nondestructive test (NDE) methods focusing particularly on local measurements that provide key indicators of cable aging and damage. The work includes a review of relevant literature as well as hands-on experimental verification of inspection capabilities. As NPPs consider applying for second, or subsequent, license renewal (SLR) to extend their operating period from 60 years to 80 years, it important to understand how the materials installed in plant systems and components will age during that time and develop aging management programs (AMPs) to assure continued safe operation under normal and design basis events (DBE). Normal component and system tests typically confirm the cables can perform their normal operational function. The focus of the cable test program is directed toward the more demanding challenge of assuring the cable function under accident or DBE. Most utilities already have a program associated with their first life extension from 40 to 60 years. Regrettably, there is neither a clear guideline nor a single NDE that can assure cable function and integrity for all cables. Thankfully, however, practical implementation of a broad range of tests allows utilities to develop a practical program that assures cable function to a high degree. The industry has adopted 50% elongation at break (EAB) relative to the un-aged cable condition as the acceptability standard. All tests are benchmarked against the cable EAB test. EAB is a destructive test so the test programs must apply an array of other NDE tests to assure or infer the overall set of cable’s system integrity. These cable NDE programs vary in rigor and methodology. As the industry gains experience with the efficacy of these programs, it is expected that implementation practice will converge to a more common approach. This report addresses the range of local NDE cable tests that are

  9. The Institutional Challenges of Payment for Ecosystem Service Program in China: A Review of the Effectiveness and Implementation of Sloping Land Conversion Program

    Directory of Open Access Journals (Sweden)

    Cheng Chen

    2015-05-01

    Full Text Available This study is an overview of the effectiveness and institutional challenges of China’s Sloping Land Conversion Program (SLCP. The SLCP is the Chinese government’s largest Payment for Ecosystem Services (PES program and one of the largest PES programs in the world. From an institutional perspective, the SLCP is particularly interesting because it represents a hybrid governance type that includes both voluntary and hierarchical (top-down elements rather than traditional command-and-control approach. Our analysis is based on a literature review that encompasses 164 international scientific articles. To identify institutional challenges, we linked the results regarding the effectiveness of the program to its institutional aspects. Our SLCP case study highlights the dependence of the effectiveness of a governmental PES program on the specific regulatory institutional setting and the particular actors involved. Our results show that some institutional challenges undermine the anticipated advantages of PES (local participation and eventually reshape the program outcomes through implementation process, particularly in cases of hybrid governance structures in which institutional requirements are as important as the design of the specific financial incentives. The collaboration between relevant government actors at different hierarchical levels, and specifically the motivations and interests of the government actors responsible for the implementation on the ground, play crucial roles. The SLCP can be an important milestone in environmental policy in China and the world, if more innovative elements of a theoretically ideal PES—such as local flexibility and self-interest (or at least the acceptance of the service providers supplying the relevant ecosystem services can be strengthened. The environmental goals can be achieved in combination with greater self-interest of the applicable government actors on all hierarchical levels.

  10. Conversion of Molybdenum-99 production process to low enriched uranium: Neutronic and thermal hydraulic analyses of HEU and LEU target plates for irradiation in Pakistan Research Reactor-1

    Science.gov (United States)

    Mushtaq, Ahmad; Iqbal, Masood; Bokhari, Ishtiaq Hussain; Mahmood, Tayyab; Muhammad, Atta

    2012-09-01

    Technetium-99m, the daughter product of Molybdenum-99 is the most widely needed radionuclide for diagnostic studies in Pakistan. Molybdenum-99 Production Facility has been established at PINSTECH. Highly enriched uranium (93% 235U) U/Al alloy targets have been irradiated in Pakistan Research Reactor-1 (PARR-1) for the generation of fission Mo-99, while basic dissolution technique is used for separation of Mo-99 from target matrix activity. In line with the international objective of minimizing and eventually eliminating the use of HEU in civil commerce, national and international efforts have been underway to shift the production of medical isotopes from HEU to LEU (LEU; uranium is needed. LEU aluminum uranium dispersion target has been developed, which may replace existing HEU aluminum/uranium alloy targets for production of 99Mo using basic dissolution technique. Neutronic and thermal hydraulic calculations were performed for safe irradiation of targets in the core of PARR-1.

  11. NOVEL REACTOR FOR THE PRODUCTION OF SYNTHESIS GAS

    Energy Technology Data Exchange (ETDEWEB)

    Vasilis Papavassiliou; Leo Bonnell; Dion Vlachos

    2004-12-01

    Praxair investigated an advanced technology for producing synthesis gas from natural gas and oxygen This production process combined the use of a short-reaction time catalyst with Praxair's gas mixing technology to provide a novel reactor system. The program achieved all of the milestones contained in the development plan for Phase I. We were able to develop a reactor configuration that was able to operate at high pressures (up to 19atm). This new reactor technology was used as the basis for a new process for the conversion of natural gas to liquid products (Gas to Liquids or GTL). Economic analysis indicated that the new process could provide a 8-10% cost advantage over conventional technology. The economic prediction although favorable was not encouraging enough for a high risk program like this. Praxair decided to terminate development.

  12. The Cambia Sojourns Scholars Leadership Program: Conversations with Emerging Leaders in Palliative Care.

    Science.gov (United States)

    Cruz-Oliver, Dulce M; Bernacki, Rachelle; Cooper, Zara; Grudzen, Corita; Izumi, Seiko; Lafond, Deborah; Lam, Daniel; LeBlanc, Thomas W; Tjia, Jennifer; Walter, Jennifer

    2017-08-01

    There is a pressing workforce shortage and leadership scarcity in palliative care to adequately meet the demands of individuals with serious illness and their families. To address this gap, the Cambia Health Foundation launched its Sojourns Scholars Leadership Program in 2014, an initiative designed to identify, cultivate, and advance the next generation of palliative care leaders. This report intends to summarize the second cohort of Sojourns Scholars' projects and their reflection on their leadership needs. This report summarizes the second cohort of sojourns scholars' project and their reflection on leadership needs. After providing a written reflection on their own projects, the second cohort participated in a group interview (fireside chat) to elicit their perspectives on barriers and facilitators in providing palliative care, issues facing leadership in palliative care in the United States, and lessons from personal and professional growth as leaders in palliative care. They analyzed the transcript of the group interview using qualitative content analysis methodology. Three themes emerged from descriptions of the scholars' project experience: challenges in palliative care practice, leadership strategies in palliative care, and three lessons learned to be a leader were identified. Challenges included perceptions of palliative care, payment and policy, and workforce development. Educating and collaborating with other clinicians and influencing policy change are important strategies used to advance palliative care. Time management, leading team effort, and inspiring others are important skills that promote effectiveness as a leader. Emerging leaders have a unique view of conceptualizing contemporary palliative care and shaping the future. Providing comprehensive, coordinated care that is high quality, patient and family centered, and readily available depends on strong leadership in palliative care. The Cambia Scholars Program represents a unique opportunity.

  13. H Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

  14. U.S. Department of Energy University Reactor Sharing Program at the University of Florida. Final report for period August 15, 2000 - May 31, 2001

    Energy Technology Data Exchange (ETDEWEB)

    Vernetson, William G.

    2002-01-01

    Department of Energy Grant Number DE-FG02-96NE38152 was supplied to the University of Florida Training Reactor (UFTR) facility through the U.S. Department of Energy's University Reactor Sharing Program. The renewal proposal submitted in January 2000 originally requested over $73,000 to support various external educational institutions using the UFTR facilities in academic year 2000-01. The actual Reactor Sharing Grant was only in the amount of $40,000, all of which has been well used by the University of Florida as host institution to support various educational institutions in the use of our reactor and associated facilities as indicated in the proposal. These various educational institutions are located primarily within the State of Florida. However, when the 600-mile distance from Pensacola to Miami is considered, it is obvious that this Grant provides access to reactor utilization for a broad geographical region and a diverse set of user institutions serving over fourteen million inhabitants throughout the State of Florida and still others throughout the Southeast.

  15. Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

    2016-01-01

    Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.

  16. Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

    2016-01-01

    Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration (Figure 1). The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration (Figure 2), based on current reactor use, has been defined for the fuel conversion analyses [1]. The code RELAP5/Mod 3.3 [2] was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.

  17. Army Gas-Cooled Reactor Systems program: alternator final design report

    Energy Technology Data Exchange (ETDEWEB)

    1964-06-01

    The development and testing of a demonstration brushless alternator for the ML-1 mobile nuclear power plant is described. The brushless concept was selected after it became apparent that a conventional power generator could not satisfy the ML-1 weight and size requirements. The demonstration alternator fabricated and tested under this program did not meet all performance specifications; the efficiency was low and the unit could not be operated for significant periods of time without overheating. However, a large body of useful data was accumulated during the extensive development program. Of special interest are data on the rotor and stator design, the cooling requirements and on the distribution of eddy current losses. Analysis of the data indicates that a brushless alternator, only slightly larger and heavier than was specified for the ML-1, could be developed with a modest additional effort.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-01-01

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

  19. Enhancement of the performance of an anaerobic sequencing batch reactor treating low-strength wastewater through implementation of a variable stirring rate program

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, J.A.D.; Pinto, A.G.; Ratusznei, S.M.; Gedraite, R. [Instituto Maua de Tecnologia (IMT), Sao Caetano do Sul, SP (Brazil). Escola de Engenharia. Dept. de Engenharia Quimica e de Alimentos]. E-mail: rodrigues@maua.br; Zaiat, M. [Sao Paulo Univ., Sao Carlos, SP (Brazil). Escola de Engenharia. Dept. de Hidraulica e Saneamento

    2004-09-01

    This work focuses on enhancement of the performance of an anaerobic sequencing batch reactor with a six-vertical-blade-disk-turbine impeller, containing granulated biomass treating low-strength synthetic wastewater, through a study of the feasibility of implementing a variable stirring rate program. The reactor was operated at 30 deg C and a six-hour cycle was used to treat approximately 2.0 L of the synthetic substrate with a chemical oxygen demand (COD) of nearly 500 mg/L. Two different stirring rate program were implemented: a constant rate of 50 rpm and a variable rate consisting of 75 rpm for one hour, 50 rpm for four hours and 25 rpm for 0.5 hour. The last 0.5 hour of the cycle was used for the settling step. In both cases, a very short start-up period and unfiltered and filtered substrate removal efficiencies of 81% and 88%, respectively, were attained. However, use of the variable stirring rate enhanced efficiency of the reactor dynamics without impairing biomass morphology, thus resulting in a reduction in the total cycle time and a possible decrease in energy consumption. Additionally, a simplified model of the anaerobic metabolic activity, using apparent kinetic parameters, was proposed as a consecutive first-order kinetic model with substrate and total volatile acid residual concentrations in order to analyze how the variable stirring rate affects reactor performance. (author)

  20. Draft program plan for TNS: The Next Step after the Tokamak Fusion Test Reactor. Part IV. Program planning

    Energy Technology Data Exchange (ETDEWEB)

    Wood, W. B.

    1977-02-01

    In this fourth part of the four-part TNS Draft Program Plan, project engineering concerns are considered. The TNS Project is first broken down into the major time and functional periods of feasibility study, preconceptual design, conceptual design, and line item construction, while the elements of the project are organized into an administrative work breakdown structure. With the aid of these two classifying schemes, the project tasks are described in terms of schedule, estimated cost, type of funding, and proposed type of participant. The initial constraints of completion data, anticipated scientific inputs, and budget procedures are used to develop a two-phase project in which the facilities are authorized first and the device 2 years later. This specific mechanism is fundamental to the construction of the schedule and should be reconsidered when the completion and initiation dates are reformulated.

  1. Department of Reactor Technology

    DEFF Research Database (Denmark)

    Risø National Laboratory, Roskilde

    The general development of the Department of Reactor Technology at Risø during 1981 is presented, and the activities within the major subject fields are described in some detail. Lists of staff, publications, and computer programs are included.......The general development of the Department of Reactor Technology at Risø during 1981 is presented, and the activities within the major subject fields are described in some detail. Lists of staff, publications, and computer programs are included....

  2. Preliminary nuclear safety assessment of the NEPST (Topaz 2) space reactor program

    Science.gov (United States)

    Marshall, A. C.

    The United States (US) Strategic Defense Initiative Organization (SDIO) decided to investigate the possibility of launching a Russian Topaz 2 space nuclear power system. A preliminary nuclear safety assessment was conducted to determine whether or not a space mission could be conducted safely and within budget constraints. As part of this assessment, a safety policy and safety functional requirements were developed to guide both the safety assessment and future Topaz 2 activities. A review of the Russian flight safety program was conducted and documented. Our preliminary nuclear safety assessment included a number of deterministic analyses, such as the following: neutronic analysis of normal and accident configurations, an evaluation of temperature coefficients of reactivity, a reentry and disposal analysis, an analysis of postulated launch abort impact accidents, and an analysis of postulated propellant fire and explosion accidents. Based on the assessment to date, it appears that it will be possible to safely launch the Topaz 2 system in the US with a modification to preclude water flooded criticality. A full scale safety program is now underway.

  3. Individual plant examination program: Perspectives on reactor safety and plant performance. Part 6, appendices A, B, and C

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    This report provides perspectives gained by reviewing 75 Individual Plant Examination (IPE) submittals pertaining to 108 nuclear power plant units. IPEs are probabilistic analyses that estimate the core damage frequency (CDF) and containment performance for accidents initiated by internal events (including internal flooding, but excluding internal fire). The U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research, reviewed the WE submittals with the objective of gaining perspectives in three major areas: (1) improvements made to individual plants as a result of their IPEs and the collective results of the IPE program, (2) plant-specific design and operational features and modeling assumptions that significantly affect the estimates of CDF and containment performance, and (3) strengths and weaknesses of the models and methods used in the IPEs. These perspectives are gained by assessing the core damage and containment performance results, including overall CDF, accident sequences, dominant contributions to component failure and human error, and containment failure modes. In particular, these results are assessed in relation to the design and operational characteristics of the various reactor and containment types, and by comparing the IPEs to probabilistic risk assessment characteristics. Methods, data, boundary conditions, and assumptions used in the IPEs are considered in understanding the differences and similarities observed among the various types of plants.

  4. Simulation in 3 dimensions of a cycle 18 months for an BWR type reactor using the Nod3D program; Simulacion en 3 dimensiones de un ciclo de 18 meses para un reactor BWR usando el programa Nod3D

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez, N.; Alonso, G. [ININ, A.P. 18-1027, 11801 Mexico D.F. (Mexico)]. E-mail: nhm@nuclear.inin.mx; Valle, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)

    2004-07-01

    The development of own codes that you/they allow the simulation in 3 dimensions of the nucleus of a reactor and be of easy maintenance, without the consequent payment of expensive use licenses, it can be a factor that propitiates the technological independence. In the Department of Nuclear Engineering (DIN) of the Superior School of Physics and Mathematics (ESFM) of the National Polytechnic Institute (IPN) a denominated program Nod3D has been developed with the one that one can simulate the operation of a reactor BWR in 3 dimensions calculating the effective multiplication factor (kJJ3, as well as the distribution of the flow neutronic and of the axial and radial profiles of the power, inside a means of well-known characteristics solving the equations of diffusion of neutrons numerically in stationary state and geometry XYZ using the mathematical nodal method RTN0 (Raviart-Thomas-Nedelec of index zero). One of the limitations of the program Nod3D is that it doesn't allow to consider the burnt of the fuel in an independent way considering feedback, this makes it in an implicit way considering the effective sections in each step of burnt and these sections are obtained of the code Core Master LEND. However even given this limitation, the results obtained in the simulation of a cycle of typical operation of a reactor of the type BWR are similar to those reported by the code Core Master LENDS. The results of the keJ - that were obtained with the program Nod3D they were compared with the results of the code Core Master LEND, presenting a difference smaller than 0.2% (200 pcm), and in the case of the axial profile of power, the maxim differs it was of 2.5%. (Author)

  5. Reactor Neutrino Spectra

    CERN Document Server

    Hayes, A C

    2016-01-01

    We present a review of the antineutrino spectra emitted from reactors. Knowledge of these and their associated uncertainties are crucial for neutrino oscillation studies. The spectra used to-date have been determined by either conversion of measured electron spectra to antineutrino spectra or by summing over all of the thousands of transitions that makeup the spectra using modern databases as input. The uncertainties in the subdominant corrections to beta-decay plague both methods, and we provide estimates of these uncertainties. Improving on current knowledge of the antineutrino spectra from reactors will require new experiments. Such experiments would also address the so-called reactor neutrino anomaly and the possible origin of the shoulder observed in the antineutrino spectra measured in recent high-statistics reactor neutrino experiments.

  6. Individual plant examination program: Perspectives on reactor safety and plant performance. Parts 2--5: Final report; Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    This report provides perspectives gained by reviewing 75 Individual Plant Examination (IPE) submittals pertaining to 108 nuclear power plant units. IPEs are probabilistic analyses that estimate the core damage frequency (CDF) and containment performance for accidents initiated by internal events. The US Nuclear Regulatory Commission (NRC) reviewed the IPE submittals with the objective of gaining perspectives in three major areas: (1) improvements made to individual plants as a result of their IPEs and the collective results of the IPE program, (2) plant-specific design and operational features and modeling assumptions that significantly affect the estimates of CDF and containment performance, and (3) strengths and weaknesses of the models and methods used in the IPEs. These perspectives are gained by assessing the core damage and containment performance results, including overall CDF, accident sequences, dominant contributions to component failure and human error, and containment failure modes. Methods, data, boundary conditions, and assumptions used in the IPEs are considered in understanding the differences and similarities observed among the various types of plants. This report is divided into three volumes containing six parts. Part 1 is a summary report of the key perspectives gained in each of the areas identified above, with a discussion of the NRC`s overall conclusions and observations. Part 2 discusses key perspectives regarding the impact of the IPE Program on reactor safety. Part 3 discusses perspectives gained from the IPE results regarding CDF, containment performance, and human actions. Part 4 discusses perspectives regarding the IPE models and methods. Part 5 discusses additional IPE perspectives. Part 6 contains Appendices A, B and C which provide the references of the information from the IPEs, updated PRA results, and public comments on draft NUREG-1560 respectively.

  7. Measurement and modeling of advanced coal conversion processes

    Energy Technology Data Exchange (ETDEWEB)

    Solomon, P.R.; Serio, M.A.; Hamblen, D.G. (Advanced Fuel Research, Inc., East Hartford, CT (USA)); Smoot, L.D.; Brewster, B.S. (Brigham Young Univ., Provo, UT (USA))

    1990-01-01

    The overall objective of this program is the development of predictive capability for the design, scale up, simulation, control and feedstock evaluation in advanced coal conversion devices. This technology is important to reduce the technical and economic risks inherent in utilizing coal, a feedstock whose variable and often unexpected behavior presents a significant challenge. This program will merge significant advances made at Advanced Fuel Research, Inc. (AFR) in measuring and quantitatively describing the mechanisms in coal conversion behavior, with technology being developed at Brigham Young University (BYU) in comprehensive computer codes for mechanistic modeling of entrained-bed gasification. Additional capabilities in predicting pollutant formation will be implemented and the technology will be expanded to fixed-bed reactors. The foundation to describe coal-specific conversion behavior is AFR's Functional Group (FG) and Devolatilization, Vaporization, and Crosslinking (DVC) models, developed under previous and on-going METC sponsored programs. These models have demonstrated the capability to describe the time dependent evolution of individual gas species, and the amount and characteristics of tar and char. The combined FG-DVC model will be integrated with BYU's comprehensive two-dimensional reactor model, PCGC-2, which is currently the most widely used reactor simulation for combustion or gasification. Success in this program will be a major step in improving in predictive capabilities for coal conversion processes including: demonstrated accuracy and reliability and a generalized first principles'' treatment of coals based on readily obtained composition data. The progress during the fifteenth quarterly of the program is presented. 56 refs., 41 figs., 5 tabs.

  8. Exploring new coolants for nuclear breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lafuente, A., E-mail: anlafuente@etsii.upm.e [ETSII-UPM, c/Jose Gutierrez Abascal, 2, 28006 Madrid (Spain); Piera, M. [ETSII:UNED, c/Juan del Rosal, 12, 28040 Madrid (Spain)

    2010-06-15

    Breeder reactors are considered a unique tool for fully exploiting natural nuclear resources. In current Light Water Reactors (LWR), only 0.5% of the primary energy contained in the nuclei removed from a mine is converted into useful heat. The rest remains in the depleted uranium or spent fuel. The need to improve resource-efficiency has stimulated interest in Fast-Reactor-based fuel cycles, which can exploit a much higher fraction of the energy content of mined uranium by burning U-238, mainly after conversion into Pu-239. Thorium fuel cycles also offer several potential advantages over a uranium fuel cycle. The coolant initially selected for most of the FBR programs launched in the 1960s was sodium, which is still considered the best candidate for these reactors. However, Na-cooled FBRs have a positive void reactivity coefficient. Among other factors, this fundamental drawback has resulted in the canceled deployment of these reactors. Therefore, it seems reasonable to explore new options for breeder coolants. In this paper, a proposal is presented for a new molten salt (F{sub 2}Be) coolant that could overcome the safety issues related to the positive void reactivity coefficient of molten metal coolants. Although it is a very innovative proposal that would require an extensive R and D program, this paper presents the very appealing properties of this salt when using a specific type of fuel that is similar to that of pebble bed reactors. The F{sub 2}Be concept was studied over a typical MOX composition and extended to a thorium-based cycle. The general analysis took into account the requirements for criticality (opening the option of hybrid subcritical systems); the requirements for breeding; and the safety requirement of having a negative coolant void reactivity coefficient. A design window was found in the definition of a F{sub 2}Be cooled reactor where the safety requirement was met, unlike for molten metal-cooled reactors, which always have positive void

  9. Fuel and target programs for the transmutation at Phenix and other reactors; Programmes combustibles et cibles pour la transmutation dans Phenix et autres reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Gaillard-Groleas, G

    2002-07-01

    The fuels and targets program for transmutation, performed in the framework of the axis 1 of the December 1991 law about the researches on the management of long-lived radioactive wastes, is in perfect consistency with the transmutation scenario studies carried out in the same framework. These studies put forward the advantage of fast breeder reactors (FBR) in the incineration of minor actinides and long-lived fission products. The program includes exploratory and technological demonstration studies covering the different design options. It aims at enhancing our knowledge of the behaviour of materials under irradiation and at ensuring the mastery of processes. The goals of the different experiments foreseen at Phenix reactor are presented. The main goal is to supply a set of results allowing to precise the conditions of the technical feasibility of minor actinides and long-lived fission products incineration in FBRs. (J.S.)

  10. Light Water Reactor Sustainability Program: Computer-based procedure for field activities: results from three evaluations at nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Oxstrand, Johanna [Idaho National Laboratory; Bly, Aaron [Idaho National Laboratory; LeBlanc, Katya [Idaho National Laboratory

    2014-09-01

    Nearly all activities that involve human interaction with the systems of a nuclear power plant are guided by procedures. The paper-based procedures (PBPs) currently used by industry have a demonstrated history of ensuring safety; however, improving procedure use could yield tremendous savings in increased efficiency and safety. One potential way to improve procedure-based activities is through the use of computer-based procedures (CBPs). Computer-based procedures provide the opportunity to incorporate context driven job aids, such as drawings, photos, just-in-time training, etc into CBP system. One obvious advantage of this capability is reducing the time spent tracking down the applicable documentation. Additionally, human performance tools can be integrated in the CBP system in such way that helps the worker focus on the task rather than the tools. Some tools can be completely incorporated into the CBP system, such as pre-job briefs, placekeeping, correct component verification, and peer checks. Other tools can be partly integrated in a fashion that reduces the time and labor required, such as concurrent and independent verification. Another benefit of CBPs compared to PBPs is dynamic procedure presentation. PBPs are static documents which limits the degree to which the information presented can be tailored to the task and conditions when the procedure is executed. The CBP system could be configured to display only the relevant steps based on operating mode, plant status, and the task at hand. A dynamic presentation of the procedure (also known as context-sensitive procedures) will guide the user down the path of relevant steps based on the current conditions. This feature will reduce the user’s workload and inherently reduce the risk of incorrectly marking a step as not applicable and the risk of incorrectly performing a step that should be marked as not applicable. As part of the Department of Energy’s (DOE) Light Water Reactors Sustainability Program

  11. Revamping Program of Urea Reactor and its Mechanism Analysis%尿素合成塔的改造方案及机理分析

    Institute of Scientific and Technical Information of China (English)

    付海华; 廖容波

    2014-01-01

    According to the urea reactor structural characteristics in the CO2 stripping process , analyze and discuss urea synthesis reaction mechanism , propose the revamping program and its technical and economic effects .%结合CO2汽提工艺尿素合成塔的结构特点,分析与探讨尿素合成反应机理,并提出改造方案,以及其技术经济效果。

  12. Photovoltaic energy conversion in vocational training. Training program for electrical engineering; Photovoltaik in der beruflichen Ausbildung. Lernprogramm. Berufsfeld Elektrotechnik

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-07-01

    ISET is an application-orientated research institute for designing electric systems for renewable energy use, rational energy use, andwater treatment. ISET focuses on environmentally acceptable an sustainable use of energy and water and the saving of natural resources. This training program was developed by ISET in cooperation with the department of vocational training of Gesamthochschule Kassel University and is based on two diploma dissertations. It is available on CD-ROM and presents the fundamentals of photovoltaic energy conversion for vocational training in the field of electrical engineering. [German] Das ISET ist ein anwendungsorientiertes Forschungsinstitut zur Gestaltung der elektrischen Systemtechnik fuer die Nutzung erneuerbarer Energien, der rationellen Energieverwendung und der Wasserbehandlung. Wir setzen uns fuer einen umweltvertraeglichen und nachhaltigen Umgang mit Energie und Wasser zur Erhaltung der natuerlichen Lebensgrundlagen ein. Bei der Entwicklung des Lernprogrammes hat das ISET die technisch-wissenschaftliche Betreuung uebernommen. Das Lernprogramm wurde im Rahmen von zwei Diplomarbeiten an der Universitaet Gesamthochschule Kassel, Fachbereich Berufs- und Fachdidaktik, entwickelt und wird in Form einer CD-ROM angeboten. Die curriculare und didaktische Entwicklung eines Lerngebietes zur Photovoltaik fuer die Berufsausbildung im Berufsfeld Elektrotechnik waren Schwerpunkt der Arbeiten. (orig./AKF)

  13. Impacts of Sloping Land Conversion Program on the vegetation in loess hilly and gully area of northern Shaanxi

    Institute of Scientific and Technical Information of China (English)

    Ma Junfei; Lu Changhe; Yu Bohua

    2009-01-01

    Aiming at alleviating the serious soil erosion,the Chinese government initiated the Sloping Land Con.version Program (SLCP) in 1999.Now,after 8 years of project implementation,the ecological recovery effects of the SLCP have become the hot issue of academic circle.This paper,taking the loess hdl and gully area of northern Shaanxi as an example,presents a methodology for assessing the vegetation restoration effect of SLCP with normalized difference vegetation index (NDVI).The key components include calculation of the Growing Season NDVI (GSNDVI),and estimation of the NVDI change induced by climate and SLCP,re,wectively.Based on the method,the NDVI change between 2000 and 2006 was obtained using the GSNDVI that excluded the noise from snow and ice.Afier the part of total NDVI change caused by climate variation was estimated using empiric formulae,we obtained the part induced by human factors,i.e.the SLCP.The human induced part of NDVI change was considered as an approximation indicating the effect of the SLCP on the vegetation.Finally,we analyzed the NDVI change characters of the whole study area,different slope lands and different land use types by spatial statistics method Results show that the vegetation condition is significantly improved by the SLCP,particularly land types that directly involved in the SLCP,such as steeply slope farmlands,degraded grasslands,etc.

  14. Automated Work Packages Prototype: Initial Design, Development, and Evaluation. Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Oxstrand, Johanna Helene [Idaho National Lab. (INL), Idaho Falls, ID (United States); Al Rashdan, Ahmad [Idaho National Lab. (INL), Idaho Falls, ID (United States); Le Blanc, Katya Lee [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bly, Aaron Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Agarwal, Vivek [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-07-01

    The goal of the Automated Work Packages (AWP) project is to demonstrate how to enhance work quality, cost management, and nuclear safety through the use of advanced technology. The work described in this report is part of the digital architecture for a highly automated plant project of the technical program plan for advanced instrumentation, information, and control (II&C) systems technologies. This report addresses the DOE Milestone M2LW-15IN0603112: Describe the outcomes of field evaluations/demonstrations of the AWP prototype system and plant surveillance and communication framework requirements at host utilities. A brief background to the need for AWP research is provided, then two human factors field evaluation studies are described. These studies focus on the user experience of conducting a task (in this case a preventive maintenance and a surveillance test) while using an AWP system. The remaining part of the report describes an II&C effort to provide real time status updates to the technician by wireless transfer of equipment indications and a dynamic user interface.

  15. Preliminary Multirod Burst Test Program results and implications of interest to reactor safety evaluations

    Energy Technology Data Exchange (ETDEWEB)

    Chapman, R.H. (comp.)

    1978-01-01

    The Multirod Burst Test (MRBT) Program, in progress at Oak Ridge National Laboratory, is investigating LWR cladding deformation in single- and multirod test arrays under conditions representative of reflood and refill phases of a LOCA. In these tests internally pressurized, unirradiated Zircaloy-4 tubes containing electrically heated fuel simulators are tested to failure in a low-pressure, superheated-steam environment. The tubes are ''uniformly'' heated over a 915-mm length; the simulator pressure, due to the small enclosed gas volume, also varies with temperature (and deformation) during the test. Two 4 x 4 multirod tests (B-1 and B-2), one with and one without the shroud being heated, have been conducted with a bundle heating rate of approx. 29/sup 0/C/sec; initial pressure conditions for these tests were selected to cause failure at about 860/sup 0/C. An additional 4 x 4 array (B-3) was tested using a bundle heating rate of approx. 10/sup 0/C/sec; the shroud was also heated in this test. Initial conditions were adjusted to cause failure at approx. 760/sup 0/C. Posttest examination (including flow tests) of the B-1 and B-2 test arrays, is essentially complete, and pertinent data are included in this summary.

  16. Waste conversion into n-caprylate and n-caproate: resource recovery from wine lees using anaerobic reactor microbiomes and in-line extraction

    Directory of Open Access Journals (Sweden)

    Leo A. Kucek

    2016-11-01

    Full Text Available To convert wastes into sustainable liquid fuels and chemicals, new resource recovery technologies are required. Chain elongation is a carboxylate-platform bioprocess that converts short-chain carboxylates (SCCs (e.g., acetate C2 and n-butyrate C4 into medium-chain carboxylates (MCCs (e.g., n-caprylate C8 and n-caproate C6 with hydrogen gas as a side product. Ethanol or another electron donor (e.g., lactate, carbohydrate is required. Competitive MCC productivities, yields (product vs. substrate fed, and specificities (product vs. all products were only achieved previously from an organic waste material when exogenous ethanol had been added. Here, we converted a real organic waste, which inherently comprised of ethanol, into MCCs with n-caprylate as the target product. We used wine lees, which consisted primarily of settled yeast cells and ethanol from wine fermentation, and produced MCCs with a reactor microbiome. We operated the bioreactor at a pH of 5.2 and with continuous in-line extraction and achieved a MCC productivity of 3.9 g COD/L-d at an organic loading rate of 5.8 g COD/L-d, resulting in a promising MCC yield of 67% and specificities of 36% for each n-caprylate and n-caproate (72% for both. Compared to all other studies that used complex organic substrates, we achieved the highest n-caprylate-to-n-caproate product ratio of 1.0 (COD basis, because we used increased broth-recycle rates through the forward membrane contactor, which improved in-line extraction rates. Increased recycle rates also allowed us to achieve the highest reported MCC production flux per membrane surface area thus far (20.1 g COD/m2-d. Through microbial community analyses, we determined that an operational taxonomic unit (OTU for Bacteroides spp. was dominant and was positively correlated with increased MCC productivities. Our data also suggested that the microbiome may have been shaped for improved MCC production by the high broth-recycle rates. Comparable abiotic

  17. Waste Conversion into n-Caprylate and n-Caproate: Resource Recovery from Wine Lees Using Anaerobic Reactor Microbiomes and In-line Extraction

    Science.gov (United States)

    Kucek, Leo A.; Xu, Jiajie; Nguyen, Mytien; Angenent, Largus T.

    2016-01-01

    To convert wastes into sustainable liquid fuels and chemicals, new resource recovery technologies are required. Chain elongation is a carboxylate-platform bioprocess that converts short-chain carboxylates (SCCs) (e.g., acetate [C2] and n-butyrate [C4]) into medium-chain carboxylates (MCCs) (e.g., n-caprylate [C8] and n-caproate [C6]) with hydrogen gas as a side product. Ethanol or another electron donor (e.g., lactate, carbohydrate) is required. Competitive MCC productivities, yields (product vs. substrate fed), and specificities (product vs. all products) were only achieved previously from an organic waste material when exogenous ethanol had been added. Here, we converted a real organic waste, which inherently contains ethanol, into MCCs with n-caprylate as the target product. We used wine lees, which consisted primarily of settled yeast cells and ethanol from wine fermentation, and produced MCCs with a reactor microbiome. We operated the bioreactor at a pH of 5.2 and with continuous in-line extraction and achieved a MCC productivity of 3.9 g COD/L-d at an organic loading rate of 5.8 g COD/L-d, resulting in a promising MCC yield of 67% and specificities of 36% for each n-caprylate and n-caproate (72% for both). Compared to all other studies that used complex organic substrates, we achieved the highest n-caprylate-to-ncaproate product ratio of 1.0 (COD basis), because we used increased broth-recycle rates through the forward membrane contactor, which improved in-line extraction rates. Increased recycle rates also allowed us to achieve the highest reported MCC production flux per membrane surface area thus far (20.1 g COD/m2-d). Through microbial community analyses, we determined that an operational taxonomic unit (OTU) for Bacteroides spp. was dominant and was positively correlated with increased MCC productivities. Our data also suggested that the microbiome may have been shaped for improved MCC production by the high broth-recycle rates. Comparable abiotic

  18. Electro-catalytic membrane reactors and the development of bipolar membrane technology

    NARCIS (Netherlands)

    Balster, J.; Stamatialis, D.F.; Wessling, M.

    2004-01-01

    Membrane reactors are currently under extensive research and development. Hardly any concept, however, is realized yet in practice. Frequently, forgotten as membrane reactors are electro-catalytic membrane reactors where electrodes perform chemical conversations and membranes separate the locations

  19. Light Water Reactor Sustainability Program Support and Modeling for the Boiling Water Reactor Station Black Out Case Study Using RELAP and RAVEN

    Energy Technology Data Exchange (ETDEWEB)

    Diego Mandelli; Curtis Smith; Thomas Riley; John Schroeder; Cristian Rabiti; Aldrea Alfonsi; Joe Nielsen; Dan Maljovec; Bie Wang; Valerio Pascucci

    2013-09-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated. In order to evaluate the impact of these two factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the impact of power uprate on the safety of a boiled water reactor system. The case study considered is a loss of off-site power followed by the loss of diesel generators, i.e., a station black out (SBO) event. Analysis is performed by using a thermo-hydraulic code, i.e. RELAP-5, and a stochastic analysis tool currently under development at INL, i.e. RAVEN. Starting from the event tree models contained in SAPHIRE, we built the input file for RELAP-5 that models in great detail system dynamics under SBO conditions. We also interfaced RAVEN with RELAP-5 so that it would be possible to run multiple RELAP-5 simulation runs by changing specific keywords of the input file. We both employed classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. We also employed advanced data analysis and visualization tools that helped us to correlate simulation outcome such as maximum core temperature with a set of input uncertain parameters. Results obtained gave a detailed overview of the issues associated to power uprate for a SBO accident scenario. We were able to quantify how timing of safety related events were impacted by a higher reactor core power. Such insights can provide useful material to the decision makers to perform risk-infomed safety margins management.

  20. Studies on advanced water-cooled reactors beyond generation Ⅲ for power generation

    Institute of Scientific and Technical Information of China (English)

    CHENG Xu

    2007-01-01

    China's ambitious nuclear power program motivates the country's nuclear community to develop advanced reactor concepts beyond generation Ⅲ to ensure a long-term, stable, and sustainable development of nuclear power. The paper discusses some main criteria for the selection of future water-cooled reactors by considering the specific Chinese situation. Based on the suggested selection criteria, two new types of water-cooled reactors are recommended for future Chinese nuclear power generation. The high conversion pressurized water reactor utilizes the present PWR technology to a large extent. With a conversion ratio of about 0.95, the fuel utilization is increased about 5 times. This significantly improves the sustainability of fuel resources. The supercritical water-cooled reactor has favorable features in economics,sustainability and technology availability. It is a logical extension of the generation Ⅲ PWR technology in China.The status of international R&D work is reviewed. A new supercritieal water-cooled reactor (SCWR) core structure (the mixed reactor core) and a new fuel assembly design (two-rows FA) are proposed. The preliminary analysis using a coupled neutron-physics/thermal-hydranlics method is carded out. It shows good feasibility for the new design proposal.

  1. User's instructions for ORCENT II: a digital computer program for the analysis of steam turbine cycles supplied by light-water-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Fuller, L.C.

    1979-02-01

    The ORCENT-II digital computer program will perform calculations at valves-wide-open design conditions, maximum guaranteed rating conditions, and an approximation of part-load conditions for steam turbine cycles supplied with throttle steam characteristic of contemporary light-water reactors. Turbine performance calculations are based on a method published by the General Electric Company. Output includes all information normally shown on a turbine-cycle heat balance diagram. The program is written in FORTRAN IV for the IBM System 360 digital computers at the Oak Ridge National Laboratory.

  2. ECUT: Energy Conversion and Utilization Technologies program. Industry, university and research interest in the US Department of Energy ECUT biocatalysis research activity

    Science.gov (United States)

    Wilcox, R. E.

    1983-01-01

    The results of a Research Opportunity Notice (RON) disseminated by the Jet Propulsion Laboratory for the U.S. Department of Energy Conversion and Utilization Technologies (ECUT) Program's Biocatalysis Research Activity are presented. The RON was issued in late April of 1983 and solicited expressions of interest from petrochemical and chemical companies, bioengineering firms, biochemical engineering consultants, private research laboratories, and universities for participating in a federal research program to investigate potential applications of biotechnology in producing chemicals. The RON results indicate that broad interest exists within the nation's industry, universities, and research institutes for the Activity and its planned research and development program.

  3. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1981-08-01

    Research activities are described concerning HTGR chemistry; fueled graphite development; prestressed concrete pressure vessel development; structural materials; HTGR graphite studies; HTR core evaluation; reactor physics; shielding; application and project assessments; and HTR Core Flow Test Loop studies.

  4. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

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

  5. Research and Development Program in Reactor Diagnostics and Monitoring with Neutron Noise Methods. Stage 11 and 12. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Sunde, C.; Pazsit, I.; Demaziere, C.; Dahl, O.; Mileshina, L. [Chalmers Univ. of Technology, Goeteborg (Sweden). Dept. of Nuclear Engineering

    2006-06-15

    This report describes the results obtained during Stage 11 and 12 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. The long-term goals are elaborated in more detail in e.g. the Final Reports of stage 1 and 2 (SKI Report 95:14 and 96:50). Higher order eigenfunctions and eigenvalues of the diffusion equation, describing a static core, have lost their significance when doing calculations in realistic systems, since there are usually made by nodal methods or other direct numerical techniques. However, there are situations when knowledge of the higher order modes is still useful. Such case is the separation of the global and regional flux oscillations in the case of BWR instability. Another case is the investigation of the stability properties of large inhomogeneous cores, which is usually quantified with the so-called eigenvalue separation, ES=1/k{sub 1} - 1/k{sub 0} where k{sub 0} and k{sub 1}, are the fundamental and first higher order eigenvalues, respectively. Numerical codes used for ICMF calculations usually do not have the option of calculating the higher order eigenvalues and eigenfunctions. In idealised systems, these can be calculated with semi-analytical methods. In order to get insight into the characteristics of the higher order eigenmodes of the neutron flux and the adjoint, these were calculated in a reflected homogeneous system with two-group theory. The adjoint eigenmodes are necessary if an orthogonality property needs to be used, such as the separation of the modes from a flux shape which is a sum of several eigenmodes. At the same time, the so-called numerical noise simulator, developed at the Dept., was extended such that it became suitable for the calculation of the higher order eigenmodes, both direct and adjoint ones. This simulator can treat real inhomogeneous cores, with an input deck compatible with that of SIMULATE. The simulator was used to calculate the

  6. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

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

  7. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  8. A LINEAR PROGRAMMING MODEL OF THE GASEOUSDIFFUSION ISOTOPE-SEPARATION PROCESS,

    Science.gov (United States)

    ISOTOPE SEPARATION, LINEAR PROGRAMMING ), (*GASEOUS DIFFUSION SEPARATION, LINEAR PROGRAMMING ), (* LINEAR PROGRAMMING , GASEOUS DIFFUSION SEPARATION), NUCLEAR REACTORS, REACTOR FUELS, URANIUM, PURIFICATION

  9. Natural gas conversion program of Rio de Janeiro-Brazil; Projeto de conversao do sistema de distribuicao de gas canalizado na cidade do Rio de Janeiro

    Energy Technology Data Exchange (ETDEWEB)

    Pallottino, J.T.; Costa Filho, M.A.F.; Guarana, L.F.M.L. [Universidade do Estado do Rio de Janeiro (UERJ), RJ (Brazil); Margalef, J.M. [Companhia Estadual de Gas (CEG), Rio de Janeiro, RJ (Brazil)

    2008-07-01

    This paper presents the Conversion Project of Rio de Janeiro, Brazil, that converted all customers and the distribution network from Manufactured Gas (MG) to Natural Gas (NG), evaluating its logistic and benefits, highlighting those related to consumer safety. The Conversion Project was executed under the responsibility of CEG with the support of UERJ, specially on the definition and revision of the technical procedures as well as the quality assurance of the project execution. As main advantages followed by MG to NG network conversion, we can note: increasing of gas distribution network capacity and gas network operational condition improvements; elimination of production plant; revision and renovation of gas network; revision and repairing of building gas facilities; augmenting of NG usage; elimination of carbon monoxide as a supplied gas constituent. Conservation condition of most consumer internal gas facilities and ambient ventilation were so critical that conversion program resulted in no measurable accident prevention. The achieved safety level should be maintained through a periodically re-evaluation program of customer's gas facilities and appliances, with special focus on permanent ventilation required for NG safety usage. (author)

  10. Conversational Dominance.

    Science.gov (United States)

    Esau, Helmut; Poth, Annette

    Details of conversational behavior can often not be interpreted until the social interaction, including the rights and obligations of the participants, their intent, the topic, etc., has been defined. This paper presents a model of conversation in which the conversational image a person presents in a given conversational situation is a function of…

  11. Supplemental Thermal-Hydraulic Transient Analyses of BR2 in Support of Conversion to LEU Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

    2016-01-01

    Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The RELAP5/Mod 3.3 code has been used to perform transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. A RELAP5 model of BR2 has been validated against select transient BR2 reactor experiments performed in 1963 by showing agreement with measured cladding temperatures. Following the validation, the RELAP5 model was then updated to represent the current use of the reactor; taking into account core configuration, neutronic parameters, trip settings, component changes, etc. Simulations of the 1963 experiments were repeated with this updated model to re-evaluate the boiling risks associated with the currently allowed maximum heat flux limit of 470 W/cm2 and temporary heat flux limit of 600 W/cm2. This document provides analysis of additional transient simulations that are required as part of a modern BR2 safety analysis report (SAR). The additional simulations included in this report are effect of pool temperature, reduced steady-state flow rate, in-pool loss of coolant accidents, and loss of external cooling. The simulations described in this document have been performed for both an HEU- and LEU-fueled core.

  12. Content for Conversation Partners.

    Science.gov (United States)

    Olson, Kathleen

    2002-01-01

    Suggests that a good strategy for helping English language learners to develop communicative competence in English is by pairing them with native English speakers. In such conversation programs, conversation partners should be provided with topics and activities that incorporate the goals, interests, and experiences of the learners. Recommends…

  13. Gas-cooled reactor programs. Fuel-management positioning and accounting module: FUELMANG Version V1. 11, September 1981

    Energy Technology Data Exchange (ETDEWEB)

    Medlin, T.W.; Hill, K.L.; Johnson, G.L.; Jones, J.E.; Vondy, D.R.

    1982-01-01

    This report documents the code module FUELMANG for fuel management of a reactor. This code may be used to position fuel during the calculation of a reactor history, maintain a mass balance history of the fuel movement, and calculate the unit fuel cycle component of the electrical generation cost. In addition to handling fixed feed fuel without recycle, provision has been made for fuel recycle with various options applied to the recycled fuel. A continuous fueling option is also available with the code. A major edit produced by the code is a detailed summary of the mass balance history of the reactor and a fuel cost analysis of that mass balance history. This code is incorporated in the system containing the VENTURE diffusion theory neutronics code for routine use. Fuel movement according to prescribed instructions is performed without the access of additional user input data during the calculation of a reactor operating history. Local application has been primarily for analysis of the performance of gas-cooled thermal reactor core concepts.

  14. Reactor Neutrinos

    OpenAIRE

    Soo-Bong Kim; Thierry Lasserre; Yifang Wang

    2013-01-01

    We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very ...

  15. BOILING REACTORS

    Science.gov (United States)

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  16. Status of core conversion with LEU silicide fuel in JRR-4

    Energy Technology Data Exchange (ETDEWEB)

    Nakajima, Teruo; Ohnishi, Nobuaki; Shirai, Eiji [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)

    1997-08-01

    Japan Research Reactor No.4 (JRR-4) is a light water moderated and cooled, 93% enriched uranium ETR-type fuel used and swimming pool type reactor with thermal output of 3.5MW. Since the first criticality was achieved on January 28, 1965, JRR-4 has been used for shielding experiments, radioisotope production, neutron activation analyses, training for reactor engineers and so on for about 30 years. Within the framework of the RERTR Program, the works for conversion to LEU fuel are now under way, and neutronic and thermal-hydraulic calculations emphasizing on safety and performance aspects are being carried out. The design and evaluation for the core conversion are based on the Guides for Safety Design and Evaluation of research and testing reactor facilities in Japan. These results show that the JRR-4 will be able to convert to use LEU fuel without any major design change of core and size of fuel element. LEU silicide fuel (19.75%) will be used and maximum neutron flux in irradiation hole would be slightly decreased from present neutron flux value of 7x10{sup 13}(n/cm{sup 2}/s). The conversion works are scheduled to complete in 1998, including with upgrade of the reactor building and utilization facilities.

  17. Neutronic Analyses for HEU to LEU fuel conversion of the Massachusetts Institute of Technology.

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, E. H.; Newton, T. H.; Bergeron, A.; Horelik, N.; Stevens, J. G (Nuclear Engineering Division); ( NS)

    2011-03-02

    The Massachusetts Institute of Technology (MIT) reactor (MITR-II), based in Cambridge, Massachusetts, is a research reactor designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on a mixture of uranium and molybdenum (UMo) is expected to allow the conversion of compact high performance reactors like the MITR-II. This report presents the results of steady state neutronic safety analyses for conversion of MITR-II from the use of HEU fuel to the use of U-Mo LEU fuel. The objective of this work was to demonstrate that the safety analyses meet current requirements for an LEU core replacement of MITR-II.

  18. Super safe small reactor RAPID-L conceptual design and R and D. JAERI's nuclear research promotion program, H11-002 (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Kambe, Mitsuru [Central Research Inst. of Electric Power Industry, Tokyo (Japan); Tsunoda, Hirokazu [Mitsubishi Research Inst., Inc., Tokyo (Japan); Mishima, Kaichiro [Kyoto Univ., Kyoto (Japan); Kawasaki, Akira [Tohoku Univ., School of Engineering, Sendai, Miyagi (Japan); Iwamura, Takamichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    The 200 kWe uranium nitride fueled lithium cooled fast reactor 'RAPID-L' combined with thermoelectric power conversion system that can be operated unmanned without refueling for up to ten years has been demonstrated. The RAPID refueling concept enables quick and simplified refueling, and achieves plant design lifetime over 20 years. A significant advantage of the RAPID-L design, which does not require the use of control rods - is the introduction of the innovative reactivity control systems: lithium expansion module (LEM), lithium injection module (LIM) and lithium release module (LRM). LEM is the most promising candidate for improving inherent reactivity feedback. LEMs could realize burnup compensation. LIMs assure sufficient negative reactivity feedback in unprotected transients. LRMs enable an automated reactor startup by detecting the hot standby temperature of the primary coolant. All these systems use {sup 6}Li as liquid poison and are actuated by highly reliable physical properties (volume expansion of {sup 6}Li for LEM, and freeze seal melting for LIM and LRM). (author)

  19. Membrane reactor. Membrane reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shindo, Y.; Wakabayashi, K. (National Chemical Laboratory for Industry, Tsukuba (Japan))

    1990-08-05

    Many reaction examples were introduced of membrane reactor, to be on the point of forming a new region in the field of chemical technology. It is a reactor to exhibit excellent function, by its being installed with membrane therein, and is generally classified into catalyst function type and reaction promotion type. What firstly belongs to the former is stabilized zirconia, where oxygen, supplied to the cathodic side of membrane with voltage, impressed thereon, becomes O {sup 2 {minus}} to be diffused through the membrane and supplied, as variously activated oxygenous species, on the anodic side. Examples with many advantages can be given such as methane coupling, propylene oxidation, methanating reaction of carbon dioxide, etc. Apart, palladium film and naphion film also belong to the former. While examples of the latter comprise, among others, decomposition of hydrogen sulfide by porous glass film and dehydrogenation of cyclohexane or palladium alloy film, which are expected to be developed and materialized in the industry. 33 refs., 8 figs.

  20. LOCA Simulation in the National Research Universal Reactor Program Postirradiation Examination Results for the Third Materials Experiment (MT-3) - Second Campaign

    Energy Technology Data Exchange (ETDEWEB)

    Haberman, J. H.

    1985-06-01

    A series of in-reactor experiments were conducted using full-length 32-rod pressurized water reactor (PWR} fuel bundles as part of the Loss-of-Coolant Accident (LOCA} Simulation Program by Pacific Northwest Laboratory (PNL). The third materials test (MT-3} was the sixth experiment in a series of thermalhydraulic and materials deformation/rupture experiments conducted in the National Research Universal (NRU) Reactor, Chalk River, Ontario, Canada. The MT-3 experiment was jointly funded by the U.S. Nuclear Regulatory Commission (NRC) and the United Kingdom Atomic Energy Authority (UKAEA) with the main objective of evaluating ballooning and rupture during active two-phase cooling at elevated temperatures. All 12 test rods in the center of the 32-rod bundle failed with an average peak strain of 55.4%. At the request of the UKAEA, a destructive postirradiation examination (PIE) was performed on 7 of the 12 test rods. The results of this examination were presented in a previous report. Subsequently, and at the request of UKAEA, PIE was performed on three additional rods along with further examination of one of the previously examined rods. Information obtained from the PIE included cladding thickness measurements, cladding metallography, and particle size analysis of the fractured fuel pellets. This report describes the additional PIE work performed and presents the results of the examinations.

  1. TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Liles, D.R.; Mahaffy, J.H.

    1986-07-01

    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.

  2. Sodium Based Heat Pipe Modules for Space Reactor Concepts: Stainless Steel SAFE-100 Core

    Science.gov (United States)

    Martin, James J.; Reid, Robert S.

    2004-01-01

    A heat pipe cooled reactor is one of several candidate reactor cores being considered for advanced space power and propulsion systems to support future space exploration applications. Long life heat pipe modules, with designs verified through a combination of theoretical analysis and experimental lifetime evaluations, would be necessary to establish the viability of any of these candidates, including the heat pipe reactor option. A hardware-based program was initiated to establish the infrastructure necessary to build heat pipe modules. This effort, initiated by Los Alamos National Laboratory and referred to as the Safe Affordable Fission Engine (SAFE) project, set out to fabricate and perform non-nuclear testing on a modular heat pipe reactor prototype that can provide 100 kilowatt from the core to an energy conversion system at 700 C. Prototypic heat pipe hardware was designed, fabricated, filled, closed-out and acceptance tested.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-09-01

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

  4. Exploring new coolants for nuclear breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lafuente, A. [ETSI Industriales-Universidad Politecnica de Madrid, C/Jose Gutierrez Abascal, 2. 28006 Madrid (Spain)

    2010-07-01

    Breeder reactors are considered the unique tool for fully exploiting the natural nuclear resources. In current LWR, only a 0.5% of the primary energy contained in the nuclei removed from the mine is converted into useful heat, with the rest remaining in the depleted uranium or in the spent fuel. The objective of resource-efficiency stimulated the interest in Fast- Reactor-based fuel cycles which can exploit a much higher fraction of the energy content of the mined uranium by burning U-238, mainly after conversion into Pu-239. Thorium fuel cycles would also offers several potential advantages over a uranium fuel cycle. The coolant initially chosen for most of the FBR programs launched in the 60's was sodium, which still is considered the best candidate for these reactors. However, Na-cooled FBR have a positive void reactivity coefficient, which has been among others, a fundamental drawback that has cancelled the deployment of these reactors. Therefore, it seems reasonable to explore totally new options on coolants for breeders. In this paper, a proposal is presented on a new molten salt (F{sub 2}Be) coolant that could overcome the safety issues related to the positive void reactivity coefficient of molten metal coolants. Although it is a very innovative proposal that would need an extensive R and D programme, this paper presents the very appealing properties of this salt, in the case of using a specific type of fuel, similar to that of pebble bed reactors. The concept will be studied over a typical MOX composition and extended to a Thorium-based cycle. The general analysis takes into account requirements for criticality (opening the option of hybrid subcritical systems); requirements for breeding; and the safety requirement of having a negative coolant void reactivity coefficient. A design window is found in the definition of a F{sub 2}Be cooled reactor where the safety requirement is met, unlike for molten metal cooled reactors which always have positive void

  5. Power distributions in fresh and depleted LEU and HEU cores of the MITR reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, E.H.; Horelik, N.E.; Dunn, F.E.; Newton, T.H., Jr.; Hu, L.; Stevens, J.G. (Nuclear Engineering Division); (2MIT Nuclear Reactor Laboratory and Nuclear Science and Engineering Department)

    2012-04-04

    The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (UMo) is expected to allow the conversion of U.S. domestic high performance reactors like the MITR-II reactor. Toward this goal, core geometry and power distributions are presented. Distributions of power are calculated for LEU cores depleted with MCODE using an MCNP5 Monte Carlo model. The MCNP5 HEU and LEU MITR models were previously compared to experimental benchmark data for the MITR-II. This same model was used with a finer spatial depletion in order to generate power distributions for the LEU cores. The objective of this work is to generate and characterize a series of fresh and depleted core peak power distributions, and provide a thermal hydraulic evaluation of the geometry which should be considered for subsequent thermal hydraulic safety analyses.

  6. Conversion disorder

    Science.gov (United States)

    ... this page: //medlineplus.gov/ency/article/000954.htm Conversion disorder To use the sharing features on this page, please enable JavaScript. Conversion disorder is a mental condition in which a person ...

  7. Disk MHD Conversion System for Nerva Reactor

    Science.gov (United States)

    2007-11-02

    connection thyristor and diode operation in high voltage ( HVDC ) transmission systems and no difficulty is foreseen in applying this to the specific...conditions of the space environment. Further, experience is now being gained with the series operation of GTOs, both for HVDC and other large scale

  8. Conversion of Methanogenic Substrates in Anaerobic Reactors

    NARCIS (Netherlands)

    Gonzalez-Gil, G.

    2000-01-01

    The EGSB systems represents an attractive option to extend further the use of anaerobic technology for wastewater treatment, particularly with respect to waste streams originating from chemical industries. Frequently chemical waste streams are unbalanced with respect to nutrients and/or micronutrien

  9. Improved enzymatic conversion of cellulose by ethylene treatment. Final report of innovative research program subtask, December 1977--September 1978

    Energy Technology Data Exchange (ETDEWEB)

    Linden, J C; Murphy, V G

    1978-10-01

    The stimulating effect of ethylene upon the enzymatic hydrolysis of cellulose was investigated to determine the mechanism of the stimulatory action. The mechanism was not determined; however, it was tentatively concluded that prolonged ethylene treatment unmasks the cellulose by dissolving contaminating materials such as lignin, hemicellulose, pectin, or ash. The presence of ethylene during the enzymatic hydrolysis appears to involve enzyme-substrate interactions in many cases. The evidence supports the opinion that a 20 to 40% improvement in the rate of enzymatic hydrolysis of cellulose may be obtained by the use of ethylene in the enzyme reactor. (JSR)

  10. Metallic fuels for advanced reactors

    Science.gov (United States)

    Carmack, W. J.; Porter, D. L.; Chang, Y. I.; Hayes, S. L.; Meyer, M. K.; Burkes, D. E.; Lee, C. B.; Mizuno, T.; Delage, F.; Somers, J.

    2009-07-01

    In the framework of the Generation IV Sodium Fast Reactor Program, the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. This paper presents an evaluation of metallic alloy fuels. Early US fast reactor developers originally favored metal alloy fuel due to its high fissile density and compatibility with sodium. The goal of fast reactor fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a conventional fast spectrum nuclear fuel while destroying recycled actinides. This will provide a mechanism for closure of the nuclear fuel cycle. Metal fuels are candidates for this application, based on documented performance of metallic fast reactor fuels and the early results of tests currently being conducted in US and international transmutation fuel development programs.

  11. A summary of the results from the DOE advanced gas reactor (AGR) fuel development and qualification program

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-04-01

    Modular high temperature gas-cooled reactor (HTGR) designs were developed to provide natural safety, which prevents core damage under all licensing basis events. The principle that guides their design concepts is to passively maintain core temperatures below fission product release thresholds under all accident scenarios. The required level of fuel performance and fission product retention reduces the radioactive source term by many orders of magnitude relative to source terms for other reactor types and allows a graded approach to emergency planning and the potential elimination of the need for evacuation and sheltering beyond a small exclusion area. Achieving this level, however, is predicated on exceptionally high coated-particle fuel fabrication quality and excellent performance under normal operation and accident conditions. The design goal of modular HTGRs is to meet the Environmental Protection Agency (EPA) Protective Action Guides (PAGs) for offsite dose at the Exclusion Area Boundary (EAB). To achieve this, the reactor design concepts require a level of fuel integrity that is far better than that achieved for all prior U.S.-manufactured tristructural isotropic (TRISO) coated particle fuel.

  12. Advanced light water reactor plants System 80+{trademark} design certification program. Annual progress report, October 1, 1994--September 30, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-09-01

    The purpose of this report is to provide the status of the progress that was made towards Design Certification of System 80+{trademark} during the US government`s 1995 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.

  13. Direct conversion of fusion energy

    Energy Technology Data Exchange (ETDEWEB)

    Johansson, Markus

    2003-03-01

    Deuterium and tritium are expected to be used as fuel in the first fusion reactors. Energy is released as kinetic energy of ions and neutrons, when deuterium reacts with tritium. One way to convert the kinetic energy to electrical energy, is to let the ions and neutrons hit the reactor wall and convert the heat that is caused by the particle bombardment to electrical energy with ordinary thermal conversion. If the kinetic energy of the ions instead is converted directly to electrical energy, a higher efficiency of the energy conversion is possible. The majority of the fusion energy is released as kinetic energy of neutrons, when deuterium reacts with tritium. Fusion reactions such as the D-D reactions, the D-{sup 3}He reaction and the p-{sup 11}B reaction, where a larger part of the fusion energy becomes kinetic energy of charged particles, appears therefore more suitable for direct conversion. Since they have lower reactivity than the D-T reaction, they need a larger {beta}B{sup 2}{sub 0} to give sufficiently high fusion power density. Because of this, the fusion configurations spherical torus (ST) and field-reversed configuration (FRC), where high {beta} values are possible, appear interesting. Rosenbluth and Hinton come to the conclusion that efficient direct conversion isn't possible in closed field line systems and that open geometries, which facilitate direct conversion, provide inadequate confinement for D-{sup 3}He. It is confirmed in this study that it doesn't seem possible to achieve as high direct conversion efficiency in closed systems as in open systems. ST and FRC fusion power plants that utilize direct conversion seem however interesting. Calculations with the help of Maple indicate that the reactor parameters needed for a D-D ST and a D{sub 3} He ST hopefully are possible to achieve. The best energy conversion option for a D-D or D{sub 3} He ST appears to be direct electrodynamic conversion (DEC) together with ordinary thermal conversion

  14. 1D Colloidal Hetero-Nanomaterials with Programmed Semiconductor Morphology and Metal Location for Enhancing Solar Energy Conversion.

    Science.gov (United States)

    Zhuang, Tao-Tao; Liu, Yan; Li, Yi; Sun, Meng; Sun, Zi-Jun; Du, Ping-Wu; Jiang, Jun; Yu, Shu-Hong

    2017-01-30

    A new kind of multitetrahedron sheath ternary ZnS-(CdS/Au) hetero-nanorod is prepared, in which one 1D ultrathin ZnS nanorod is integrated with segmented tetrahedron sheaths made of CdS, and more importantly, Au nanoparticles can be decorated in a targeted manner onto the vertexes and edges of CdS tetrahedron sheaths solely, for achieving performance improvement in photoelectric and photochemical conversion applications.

  15. Development of an on-line control system for the cultivation of animal cells in a hollow-fiber reactor using flowing injection analysis and a visual programming language.

    Science.gov (United States)

    Stoll, T S; Ruffieux, P A; von Stockar, U; Marison, I W

    1996-10-18

    A system for the independent control of ammonia and glutamine in a hollow-fiber reactor has been developed. On-line analysis of the two chemical species was performed using FIA. A custom-made program has been written to fully automate the FIA, to perform measurements, calibrations and data acquisition, and to activate two peristaltic feed pumps. Diagnostic tests were also automatically performed in order to identify, and in some cases eliminate, erroneous measurements and calibrations due to instrument failures. For this purpose, a visual programming language (LabVIEW) has been chosen. Ammonia concentration was controlled by varying the medium supply rate to the hollow-fiber reactor which was shown to behave like an ideal-stirred tank reactor. The glutamine concentration was controlled through the intermittent activation of a pump delivering a concentrated solution of this species to the reactor. This control system was tested with a culture of hybridoma cells at three different ammonia and one glutamine set-points. The hollow-fiber reactor was successfully operated automatically for more than 1200 h with ammonia and glutamine concentrations accurately controlled at non-limiting levels. The visual programming environment was found to be reliable and highly suitable for the development of custom programs for bioprocess control.

  16. Research and development program in reactor diagnostics and monitoring with neutron noise methods. Stage 7. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Pazsit, I.; Demaziere, C.; Arzhanov, V. [Chalmers Univ. of Technology, Goeteborg (Sweden). Department of Reactor Physics; Garis, N.S. [Swedish Nuclear Power Inspectorate, Stockholm (Sweden)

    2001-08-01

    This report constitutes stage 7 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. A proposal for the continuation of this program in stage 8 is also given at the end of the report. In stage 6, the basic principles of a 3-D fully coupled neutronic/thermal-hydraulic simulator in the frequency domain were presented. The neutronic model relied on the two-group diffusion approximation, whereas the thermal-hydraulic algorithms relied on the so called 'lumped' model. The key element of this simulator was that only the static data were required which could be obtained from the Studsvik Scandpower CASMO-4/TABLES-3/ SIMULATE-3 code package. The simulator was developed with this underlying idea, which means that the calculation of the static fluxes and the eigenvalue were avoided. Depending on what kind of spatial discretization scheme which is used in the noise simulator to calculate the 'leakage' noise, it is not granted that the system remains critical by using the group constants supplied by SIMULATE. Nevertheless, when the system is critical, the balance equations should be fulfilled in all nodes with respect to the discretization scheme used. In concrete terms, the calculation of the static fluxes and eigenvalue can be avoided if the system is brought back to criticality by modifying the cross-sections so that the balance equations are always fulfilled with the chosen spatial discretization scheme. This approach was used in this study with the finite difference scheme. As pointed out in stage 6, the finite difference scheme is relatively inefficient compared to finite elements or nodal methods, but on the other hand it is rather easy to implement. These two more sophisticated schemes are planned to be investigated at a later stage, but for the time being the simulator relying on the finite difference scheme was improved as much as possible so that a 2-D entirely

  17. Multifunctional reactors

    NARCIS (Netherlands)

    Westerterp, K.R.

    1992-01-01

    Multifunctional reactors are single pieces of equipment in which, besides the reaction, other functions are carried out simultaneously. The other functions can be a heat, mass or momentum transfer operation and even another reaction. Multifunctional reactors are not new, but they have received much

  18. Nuclear research reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Cota, Anna Paula Leite; Mesquita, Amir Zacarias, E-mail: aplc@cdtn.b, E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

  19. Nuclear research reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Cota, Anna Paula Leite; Mesquita, Amir Zacarias, E-mail: aplc@cdtn.b, E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

  20. Chromatographic and Related Reactors.

    Science.gov (United States)

    1988-01-07

    special information about effects of surface heteroge- neity in the methanation reaction. Studies of an efficient multicolumn assembly for measuring...of organic basic catalysts such as pyridine and 4-methylpicoline. It was demonstrated that the chromatographic reactor gave special information about...Programmed Reaction to obtain special information about surface heterogeneity in the methanation reaction. Advantages of stopped flow over steady state

  1. The research reactors their contribution to the reactors physics; Les reacteurs de recherche leur apport sur la physique des reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Barral, J.C. [Electricite de France (EDF), 75 - Paris (France); Zaetta, A. [CEA/Cadarache, Direction des Reacteurs Nucleaires, DRN, 13 - Saint-Paul-lez-Durance (France); Johner, J. [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee (DRFC), 13 - Saint Paul lez Durance (France); Mathoniere, G. [CEA/Saclay, DEN, 91 - Gif sur Yvette (France)] [and others

    2000-07-01

    The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

  2. High Performance Photocatalytic Oxidation Reactor System Project

    Data.gov (United States)

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

  3. Systems Engineering Model for ART Energy Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Mendez Cruz, Carmen Margarita [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Rochau, Gary E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wilson, Mollye C. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-02-01

    The near-term objective of the EC team is to establish an operating, commercially scalable Recompression Closed Brayton Cycle (RCBC) to be constructed for the NE - STEP demonstration system (demo) with the lowest risk possible. A systems engineering approach is recommended to ensure adequate requirements gathering, documentation, and mode ling that supports technology development relevant to advanced reactors while supporting crosscut interests in potential applications. A holistic systems engineering model was designed for the ART Energy Conversion program by leveraging Concurrent Engineering, Balance Model, Simplified V Model, and Project Management principles. The resulting model supports the identification and validation of lifecycle Brayton systems requirements, and allows designers to detail system-specific components relevant to the current stage in the lifecycle, while maintaining a holistic view of all system elements.

  4. Thread-Based Automatic Parallel Conversion Technique for Java Program%基于线程的Java程序自动并行转换技术

    Institute of Scientific and Technical Information of China (English)

    刘英; 刘磊; 张乃孝

    2001-01-01

    Java程序的并行化研究是一个重要课题.提出一种Java程序的自动并行转换技术,并充分利用Java语言本身提供的多线程机制,通过操作冲突性检测等方法将串行化的Java程序自动转化成并行化程序.使得转化后的并行化程序在多处理机操作系统的支持下,能在共享内存的多处理机系统上运行,从而提高了程序的运行效率.%The study of parallelism for Java program is one of the mostimportant subjects at present. In this paper, a kind of automatic parallel conversion technique is given. The serial Java program is transformed to parallel program utilizing the multithreading machanism and testing technique of commutativity operations. The parallel program after transforming can run in the supercomputer with multi-CPU under the multi-processor operating system, which will enhance the programs' efficiency.

  5. Idaho National Laboratory Experimental Program to Measure the Flow Phenomena in a Scaled Model of a Prismatic Gas-Cooled Reactor Lower Plenum for Validation of CFD Codes

    Energy Technology Data Exchange (ETDEWEB)

    Hugh M. McIlroy Jr.; Donald M. McEligot; Robert J. Pink

    2008-09-01

    The experimental program that is being conducted at the Matched Index-of-Refraction (MIR) Flow Facility at Idaho National Laboratory (INL) to obtain benchmark data on measurements of flow phenomena in a scaled model of a prismatic gas-cooled reactor lower plenum using 3-D Particle Image Velocimetry (PIV) is presented. A description of the scaling analysis, experimental facility, 3-D PIV system, measurement uncertainties and analysis, experimental procedures and samples of the data sets that have been obtained are included. Samples of the data set that will be presented include mean-velocity-field and turbulence data in an approximately 1:7 scale model of a region of the lower plenum of a typical prismatic gas-cooled reactor (GCR) similar to a General Atomics Gas-Turbine-Modular Helium Reactor (GTMHR) design. This experiment has been selected as the first Standard Problem endorsed by the Generation IV International Forum. The flow in the lower plenum consists of multiple jets injected into a confined cross flow - with obstructions. The model consists of a row of full circular posts along its centerline with half-posts on the two parallel walls to approximate flow scaled to that expected from the staggered parallel rows of posts in the reactor design. The model is fabricated from clear, fused quartz to match the refractive-index of the mineral oil working fluid. The benefit of the MIR technique is that it permits high-quality measurements to be obtained without locating intrusive transducers that disturb the flow field and without distortion of the optical paths. An advantage of the INL MIR system is its large size which allows improved spatial and temporal resolution compared to similar facilities at smaller scales. Results concentrate on the region of the lower plenum near its far reflector wall (away from the outlet duct). Inlet jet Reynolds numbers (based on the jet diameter and the time-mean average flow rate) are approximately 4,300 and 12,400. The measurements

  6. Conversational Narcissism.

    Science.gov (United States)

    Vangelisti, Anita L.; And Others

    1990-01-01

    Examines narcissistic communication and the ways it is exhibited in everyday conversation. Identifies the following behavioral referents: boasting, refocusing the topic of conversation on the self, exaggerating hand and body movements, using a loud tone of voice, and "glazing over" when others speak. Suggests that conversational…

  7. Contentious Conversations

    Science.gov (United States)

    Zuidema, Leah A.

    2011-01-01

    The idea of joining a conversation through reading and writing is not new; in his 1941 book "The Philosophy of Literary Form: Studies in Symbolic Action," Kenneth Burke suggests that the acts of reading and writing are like entering a parlor where others are already conversing. The author explores the place of professional debate within NCTE and…

  8. Reactor vessel

    OpenAIRE

    Makkee, M.; Kapteijn, F.; Moulijn, J.A

    1999-01-01

    A reactor vessel (1) comprises a reactor body (2) through which channels (3) are provided whose surface comprises longitudinal inwardly directed parts (4) and is provided with a catalyst (6), as well as buffer bodies (8, 12) connected to the channels (3) on both sides of the reactor body (2) and comprising connections for supplying (9, 10, 11) and discharging (13, 14, 15) via the channels (3) gases and/or liquids entering into a reaction with each other and substances formed upon this reactio...

  9. Study on the Adaptability of Etheriifcation Feedstock to Reactor Type

    Institute of Scientific and Technical Information of China (English)

    Mao Junyi; Yuan Qing; Wang Lei; Huang Tao

    2016-01-01

    A reactive C5 oleifns and methanol etheriifcation kinetic model based on E-R mechanism was established and three different types of reactors including the adiabatic ifxed-bed liquid reactor, the external loop reactor and the mixed-phase reactor were constructed by Aspen Plus. The adaptability of reactive C5 oleifns to these reactors was studied and simulated using various gasoline fractions with different oleifns content. After the theoretical model was validated by the experimental data of the etheriifcation of three C5 light cut fractions from different gasoline sources in different reactors, the simulated isoamylene conversion with reactive C5 olefin contents increasing from 10% to 60% was studied in the three different types of reactors for etheriifcation with methanol, respectively. Test results show that there is an obvious adaptability of the feedstock composition to the reactor type to achieve a high conversion.

  10. A gas-cooled reactor surface power system

    Science.gov (United States)

    Lipinski, Ronald J.; Wright, Steven A.; Lenard, Roger X.; Harms, Gary A.

    1999-01-01

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  11. Estimating boiling water reactor decommissioning costs: A user`s manual for the BWR Cost Estimating Computer Program (CECP) software. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Bierschbach, M.C. [Pacific Northwest National Lab., Richland, WA (United States)

    1996-06-01

    Nuclear power plant licensees are required to submit to the US Nuclear Regulatory Commission (NRC) for review their decommissioning cost estimates. This user`s manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personal computer, provides estimates for the cost of decommissioning boiling water reactor (BWR) power stations to the point of license termination. Such cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning.

  12. Study on core design for reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    The Reduced-Moderation Water Reactor (RMWR) is a water-cooled reactor with the harder neutron spectrum comparing with the LWR, resulting from low neutron moderation due to reduced water volume fraction. Based on the difference from the spectrum from the LWR, the conversion from U-238 to Pu-239 is promoted and the new cores preferable to effective utilization of uranium resource can be possible Design study of the RMWR core started in 1997 and new four core concepts (three BWR cores and one PWR core) are recently evaluated in terms of control rod worths, plutonium multiple recycle, high burnup and void coefficient. Comparative evaluations show needed incorporation of control rod programming and simplified PUREX process as well as development of new fuel cans for high burnup of 100 GW-d/t. Final choice of design specifications will be made at the next step aiming at realization of the RMWR. (T. Tanaka)

  13. Program management plan for the Molten Salt Reactor Experiment Remediation Project at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    The primary mission of the Molten Salt Reactor Experiment (MSRE) Remediation Project is to effectively implement the risk-reduction strategies and technical plans to stabilize and prevent further migration of uranium within the MSRE facility, remove the uranium and fuel salts from the system, and dispose of the fuel and flush salts by storage in appropriate depositories to bring the facility to a surveillance and maintenance condition before decontamination and decommissioning. This Project Management Plan (PMP) for the MSRE Remediation Project details project purpose; technical objectives, milestones, and cost objectives; work plan; work breakdown structure (WBS); schedule; management organization and responsibilities; project management performance measurement planning, and control; conduct of operations; configuration management; environmental, safety, and health compliance; quality assurance; operational readiness reviews; and training.

  14. Program management plan for the Molten Salt Reactor Experiment Remediation Project at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    The primary mission of the Molten Salt Reactor Experiment (MSRE) Remediation Project is to effectively implement the risk-reduction strategies and technical plans to stabilize and prevent further migration of uranium within the MSRE facility, remove the uranium and fuel salts from the system, and dispose of the fuel and flush salts by storage in appropriate depositories to bring the facility to a surveillance and maintenance condition before decontamination and decommissioning. This Project Management Plan (PMP) for the MSRE Remediation Project details project purpose; technical objectives, milestones, and cost objectives; work plan; work breakdown structure (WBS); schedule; management organization and responsibilities; project management performance measurement planning, and control; conduct of operations; configuration management; environmental, safety, and health compliance; quality assurance; operational readiness reviews; and training.

  15. Defining the “proven technology” technical criterion in the reactor technology assessment for Malaysia’s nuclear power program

    Energy Technology Data Exchange (ETDEWEB)

    Anuar, Nuraslinda, E-mail: nuraslinda@uniten.edu.my [College of Engineering, Universiti Tenaga Nasional, Jalan IKRAM-UNITEN, 43000 Kajang, Selangor (Malaysia); Kahar, Wan Shakirah Wan Abdul, E-mail: shakirah@tnb.com.my; Manan, Jamal Abdul Nasir Abd [Nuclear Energy Department, Regulatory Economics and Planning Division, Tenaga Nasional Berhad, No. 8 Jalan Tun Sambanthan, Brickfields, 50470 Kuala Lumpur (Malaysia)

    2015-04-29

    Developing countries that are considering the deployment of nuclear power plants (NPPs) in the near future need to perform reactor technology assessment (RTA) in order to select the most suitable reactor design. The International Atomic Energy Agency (IAEA) reported in the Common User Considerations (CUC) document that “proven technology” is one of the most important technical criteria for newcomer countries in performing the RTA. The qualitative description of five desired features for “proven technology” is relatively broad and only provides a general guideline to its characterization. This paper proposes a methodology to define the “proven technology” term according to a specific country’s requirements using a three-stage evaluation process. The first evaluation stage screens the available technologies in the market against a predefined minimum Technology Readiness Level (TRL) derived as a condition based on national needs and policy objectives. The result is a list of technology options, which are then assessed in the second evaluation stage against quantitative definitions of CUC desired features for proven technology. The potential technology candidates produced from this evaluation is further narrowed down to obtain a list of proven technology candidates by assessing them against selected risk criteria and the established maximum allowable total score using a scoring matrix. The outcome of this methodology is the proven technology candidates selected using an accurate definition of “proven technology” that fulfills the policy objectives, national needs and risk, and country-specific CUC desired features of the country that performs this assessment. A simplified assessment for Malaysia is carried out to demonstrate and suggest the use of the proposed methodology. In this exercise, ABWR, AP1000, APR1400 and EPR designs assumed the top-ranks of proven technology candidates according to Malaysia’s definition of “proven technology”.

  16. Chemical Reactors.

    Science.gov (United States)

    Kenney, C. N.

    1980-01-01

    Describes a course, including content, reading list, and presentation on chemical reactors at Cambridge University, England. A brief comparison of chemical engineering education between the United States and England is also given. (JN)

  17. Reactor Neutrinos

    Directory of Open Access Journals (Sweden)

    Soo-Bong Kim

    2013-01-01

    Full Text Available We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very recently the most precise determination of the neutrino mixing angle θ13. This paper provides an overview of the upcoming experiments and of the projects under development, including the determination of the neutrino mass hierarchy and the possible use of neutrinos for society, for nonproliferation of nuclear materials, and geophysics.

  18. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  19. Reactor Engineering

    Science.gov (United States)

    Lema, Juan M.; López, Carmen; Eibes, Gemma; Taboada-Puig, Roberto; Moreira, M. Teresa; Feijoo, Gumersindo

    In this chapter, the engineering aspects of processes catalyzed by peroxidases will be presented. In particular, a discussion of the existing technologies that utilize peroxidases for different purposes, such as the removal of recalcitrant compounds or the synthesis of polymers, is analyzed. In the first section, the essential variables controlling the process will be investigated, not only those that are common in any enzymatic system but also those specific to peroxidative reactions. Next, different reactor configurations and operational modes will be proposed, emphasizing their suitability and unsuitability for different systems. Finally, two specific reactors will be described in detail: enzymatic membrane reactors and biphasic reactors. These configurations are especially valuable for the treatment of xenobiotics with high and poor water solubility, respectively.

  20. China’s Conversion of Cropland to Forest Program for Household Delivery of Ecosystem Services: How Important is a Local Implementation Regime to Survival Rate Outcomes?

    Directory of Open Access Journals (Sweden)

    Michael T. Bennett

    2014-09-01

    Full Text Available China’s Conversion of Cropland to Forests Program (CCFP is the world’s largest afforestation-based Payments for Ecosystem Services (PES program, having retired and afforested over 24 million ha involving 32 million rural households. Prior research has primarily focused on the CCFP’s rural welfare impacts, with few studies on program-induced environmental improvements, particularly at the household level. In this study, data from a 2010 survey covering 2808 rural households from across China was analyzed using an interval regression model to explain household-reported survival rates of trees planted on program-enrolled cropland. In addition to household-level factors, we explore the influence of local conditions and institutional configurations by exploiting the wide diversity of contexts covered by the data set. We find that households with more available labor and more forestry experience manage trees better, but that higher opportunity costs for both land and labor have the opposite effect. We also find that the local implementation regime- e.g., the degree of prior consultation with participants and regular monitoring - has a strong positive effect on reported survivorship. We suggest that the level of subsidy support to participating households will be key to survivorship of trees in planted CCFP forests for some time to come.