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Sample records for generation iv reactors

  1. An Economic Analysis of Generation IV Small Modular Reactors

    SciTech Connect

    Stewart, J S; Lamont, A D; Rothwell, G S; Smith, C F; Greenspan, E; Brown, N; Barak, A

    2002-03-01

    This report examines some conditions necessary for Generation IV Small Modular Reactors (SMRs) to be competitive in the world energy market. The key areas that make nuclear reactors an attractive choice for investors are reviewed, and a cost model based on the ideal conditions is developed. Recommendations are then made based on the output of the cost model and on conditions and tactics that have proven successful in other industries. The Encapsulated Nuclear Heat Source (ENHS), a specific SMR design concept, is used to develop the cost model and complete the analysis because information about the ENHS design is readily available from the University of California at Berkeley Nuclear Engineering Department. However, the cost model can be used to analyze any of the current SMR designs being considered. On the basis of our analysis, we determined that the nuclear power industry can benefit from and SMRs can become competitive in the world energy market if a combination of standardization and simplification of orders, configuration, and production are implemented. This would require wholesale changes in the way SMRs are produced, manufactured and regulated, but nothing that other industries have not implemented and proven successful.

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

    SciTech Connect

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

    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 the structural

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

    SciTech Connect

    Corwin, William R; Burchell, Timothy D; Halsey, William; Hayner, George; Katoh, Yutai; Klett, James William; McGreevy, Timothy E; Nanstad, Randy K; Ren, Weiju; Snead, Lance Lewis; Stoller, Roger E; Wilson, Dane F

    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.

  4. Generation IV reactors and the ASTRID prototype: Lessons from the Fukushima accident

    NASA Astrophysics Data System (ADS)

    Gauché, François

    2012-05-01

    In France, the ASTRID prototype is a sodium-cooled fast neutron industrial demonstrator, fulfilling the criteria for Generation IV reactors. ASTRID will meet safety requirements as stringent as for 3rd generation reactors, and take into account lessons from the Fukushima accident. The objectives are to reinforce the robustness of the safety demonstration for all safety functions. ASTRID will feature an innovative core with a negative sodium void coefficient, take advantage of the large thermal inertia of SFRs for decay heat removal, and provide for a design either eliminating the sodium-water reaction, or guaranteeing no consequences for safety in case such reaction would take place.

  5. DESIGN AND LAYOUT CONCEPTS FOR COMPACT, FACTORY-PRODUCED, TRANSPORTABLE, GENERATION IV REACTOR SYSTEMS

    SciTech Connect

    Mynatt Fred R.; Townsend, L.W.; Williamson, Martin; Williams, Wesley; Miller, Laurence W.; Khan, M. Khurram; McConn, Joe; Kadak, Andrew C.; Berte, Marc V.; Sawhney, Rapinder; Fife, Jacob; Sedler, Todd L.; Conway, Larry E.; Felde, Dave K.

    2003-11-12

    The purpose of this research project is to develop compact (100 to 400 MWe) Generation IV nuclear power plant design and layout concepts that maximize the benefits of factory-based fabrication and optimal packaging, transportation and siting. The reactor concepts selected were compact designs under development in the 2000 to 2001 period. This interdisciplinary project was comprised of three university-led nuclear engineering teams identified by reactor coolant type (water, gas, and liquid metal) and a fourth Industrial Engineering team. The reactors included a Modular Pebble Bed helium-cooled concept being developed at MIT, the IRIS water-cooled concept being developed by a team led by Westinghouse Electric Company, and a Lead-Bismuth-cooled concept developed by UT. In addition to the design and layout concepts this report includes a section on heat exchanger manufacturing simulations and a section on construction and cost impacts of proposed modular designs.

  6. Generation IV Nuclear Energy Systems Construction Cost Reductions through the Use of Virtual Environments - Task 5 Report: Generation IV Reactor Virtual Mockup Proof-of-Principle Study

    SciTech Connect

    Timothy Shaw; Anthony Baratta; Vaughn Whisker

    2005-02-28

    Task 5 report is part of a 3 year DOE NERI-sponsored effort evaluating immersive virtual reality (CAVE) technology for design review, construction planning, and maintenance planning and training for next generation nuclear power plants. Program covers development of full-scale virtual mockups generated from 3D CAD data presented in a CAVE visualization facility. Created a virtual mockup of PBMR reactor cavity and discussed applications of virtual mockup technology to improve Gen IV design review, construction planning, and maintenance planning.

  7. A Project Management and Systems Engineering Structure for a Generation IV Very High Temperature Reactor

    SciTech Connect

    Ed Gorski; Dennis Harrell; Finis Southworth

    2004-09-01

    The Very High Temperature Reactor (VHTR) will be an advanced, very high temperature (approximately 1000o C. coolant outlet temperature), gas cooled nuclear reactor and is the nearest term of six Generation IV reactor technologies for nuclear assisted hydrogen production. In 2001, the Generation IV International Forum (GIF), a ten nation international forum working together with the Department of Energy’s (DOE) Nuclear Energy Research Advisory Committee (NERAC), agreed to proceed with the development of a technology roadmap and identified the next generation of nuclear reactor systems for producing new sources of power. Since a new reactor has not been licensed in the United States since the 1970s, the risks are too large for a single utility to assume in the development of an unprecedented Generation IV reactor. The government must sponsor and invest in the research to resolve major first of a kind (FOAK) issues through a full-scale demonstration prior to industry implementation. DOE’s primary mission for the VHTR is to demonstrate nuclear reactor assisted cogeneration of electricity and hydrogen while meeting the Generation IV goals for safety, sustainability, proliferation resistance and physical security and economics. The successful deployment of the VHTR as a demonstration project will aid in restarting the now atrophied U.S. nuclear power industry infrastructure. It is envisioned that VHTR project participants will include DOE Laboratories, industry partners such as designers, constructors, manufacturers, utilities, and Generation IV international countries. To effectively mange R&D, engineering, procurement, construction, and operation for this multi-organizational and technologically complex project, systems engineering will be used extensively to ensure delivery of the final product. Although the VHTR is an unprecedented FOAK system, the R&D, when assessed using the Office of Science and Technology Gate Model, falls primarily in the 3rd - Exploratory

  8. Fundamental Understanding of Crack Growth in Structural Components of Generation IV Supercritical Light Water Reactors

    SciTech Connect

    Iouri I. Balachov; Takao Kobayashi; Francis Tanzella; Indira Jayaweera; Palitha Jayaweera; Petri Kinnunen; Martin Bojinov; Timo Saario

    2004-11-17

    This work contributes to the design of safe and economical Generation-IV Super-Critical Water Reactors (SCWRs) by providing a basis for selecting structural materials to ensure the functionality of in-vessel components during the entire service life. During the second year of the project, we completed electrochemical characterization of the oxide film properties and investigation of crack initiation and propagation for candidate structural materials steels under supercritical conditions. We ranked candidate alloys against their susceptibility to environmentally assisted degradation based on the in situ data measure with an SRI-designed controlled distance electrochemistry (CDE) arrangement. A correlation between measurable oxide film properties and susceptibility of austenitic steels to environmentally assisted degradation was observed experimentally. One of the major practical results of the present work is the experimentally proven ability of the economical CDE technique to supply in situ data for ranking candidate structural materials for Generation-IV SCRs. A potential use of the CDE arrangement developed ar SRI for building in situ sensors monitoring water chemistry in the heat transport circuit of Generation-IV SCWRs was evaluated and proved to be feasible.

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

    SciTech Connect

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

    2013-10-01

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

  10. Multidimensional Model of Fluid Flow and Heat Transfer in Generation-IV Supercritical Water Reactors

    SciTech Connect

    Gallaway, Tara; Antal, Steven P.; Podowski, Michael Z.

    2006-07-01

    This paper is concerned with the mechanistic modeling and theoretical/computational analysis of flow and heat transfer in future Generation-IV Supercritical Water Cooled Reactors (SCWR). The issues discussed in the paper include: the development of analytical models of the properties of supercritical water, and the application of full three-dimensional computational modeling framework to simulate fluid flow and heat transfer in SCWRs. Several results of calculations are shown, including the evaluation of water properties (density, specific heat, thermal conductivity, viscosity, and Prandtl number) near the pseudo-critical temperature for various supercritical pressures, and the CFD predictions using the NPHASE computer code. It is demonstrated that the proposed approach is very promising for future mechanistic analyses of SCWR thermal-hydraulics and safety. (authors)

  11. Simplified Design Criteria for Very High Temperature Applications in Generation IV Reactors

    SciTech Connect

    McGreevy, TE

    2004-12-15

    The goal of this activity is to provide simplified criteria which can be used in rapid feasibility assessments of the structural viability of very high temperature components in conceptual and early preliminary design phases for Generation IV reactors. The current criteria in ASME Code Section III, Subsection NH, hereafter referred to as NH, (and Code Case N-201 for core support structures) are difficult and require a complex deconstruction of finite element analysis results for their implementation. Further, and most important, times, temperatures and some materials of interest to the very high temperature Generation IV components are not covered by the current provisions of NH. Future revisions to NH are anticipated that will address very high temperature Generation IV components and materials requirements but, until that occurs interim guidance is required for design activities to proceed. These simplified criteria are for design guidance and are not necessarily in rigorous compliance with NH methodology. Rather, the objective is for criteria which address the early design needs of very high temperature Generation IV components and materials. The intent is to provide simplified but not overly conservative design methods. When more rigorous criteria and methods are incorporated in NH, the degree of conservatism should obviously be reduced. These criteria are based on currently available information. Although engineering judgments have been made in the formulation of these criteria they are not intended to require additional development or testing prior to implementation as a tool for use in conceptual and early preliminary design. Appendices are provided herein that contain useful information. The simplified methods were developed specifically with Alloy 617 in mind; however, they could be applied for the same intended purpose for other materials such as 9Cr-1Mo, Alloy 800H, etc. However, supporting design curves, stress allowables, and isochronous curves may

  12. Assessment of the Technical Maturity of Generation IV Concepts for Test or Demonstration Reactor Applications, Revision 2

    SciTech Connect

    Gougar, Hans David

    2015-10-01

    The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each of the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.

  13. Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors: Final Scientific/Technical Report

    SciTech Connect

    Vierow, Karen; Aldemir, Tunc

    2009-09-10

    The project entitled, “Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors”, was conducted as a DOE NERI project collaboration between Texas A&M University and The Ohio State University between March 2006 and June 2009. The overall goal of the proposed project was to develop practical approaches and tools by which dynamic reliability and risk assessment techniques can be used to augment the uncertainty quantification process in probabilistic risk assessment (PRA) methods and PRA applications for Generation IV reactors. This report is the Final Scientific/Technical Report summarizing the project.

  14. SVBR-100 module-type fast reactor of the IV generation for regional power industry

    NASA Astrophysics Data System (ADS)

    Zrodnikov, A. V.; Toshinsky, G. I.; Komlev, O. G.; Stepanov, V. S.; Klimov, N. N.

    2011-08-01

    In the report the following is presented: basic conceptual provisions of the innovative nuclear power technology (NPT) based on modular fast reactors (FR) SVBR-100, summarized results of calculations of the reactor, analysis of the opportunities of multi-purpose application of such reactor facilities (RF) including export potentials with due account of nonproliferation requirements. The most important features of the proposed NPT analyzed in the report are as follows: (1) integral (monoblock) arrangement of the primary circuit equipment with entire elimination of the primary circuit pipelines and valves that considerably reduces the construction and assembly works period and coupling with high boiling point of lead-bismuth coolant (LBC) deterministically eliminates accidents of the LOCA type, (2) option for 100 MWe power and dimensions of the reactor provide: on the one hand, an opportunity to transport the reactor monoblock in factory-readiness by railway as well as other kinds of transport, on the other hand, core breeding ratio (CBR) exceeds 1 while MOX-fuel is used. The preferable area of application of RF SVBR-100 is regional and small power requiring power-units of electric power in a range of (100-600) MW, which could be used for cogeneration-based district heating while locating them nearby cities as well as for generation of electric power in a mode of load tracking in the regions with low network systems.

  15. Generation-IV Multi-Application Small Light Water Reactor (MASLWR)

    SciTech Connect

    Modro, Slawomir Michael; Fisher, James Ebberly; Weaver, Kevan Dean; Babka, P.; Reyes, Johnny Paul; Groome, J.; Wilson, Gary Edward

    2002-04-01

    The Idaho National Engineering and Environmental Laboratory (INEEL), Nexant Inc. and the Oregon State University (OSU) have developed a Multi-Application Small Light Water Reactor (MASLWR) concept. The MASLWR is a small, safe and economic natural circulation pressurized light water reactor. MASLWR reactor module consists of an integral reactor/steam generator located in a steel cylindrical containment. The entire module is to be entirely shop fabricated and transported to site on most railways or roads. Two or more modules are located in a reactor building, each being submersed in a common, below grade cavity filled with water. For the most severe postulated accident, the volume of water in the cavity provides a passive ultimate heat sink for 3 or more days allowing the restoration of lost normal active heat removal systems. MASLWR thermal power of a single module is 150 MWt, primary system pressure 10.5 MPa, steam pressure1.52 MPa and the net electrical output is 35 - 50 MWe.

  16. Generation-IV Multi-Application Small Light Water Reactor (MASLWR)

    SciTech Connect

    Modro, S. Michael; Fisher, James; Weaver, Kevan; Babka, Pierre; Reyes, Jose; Groome, John

    2002-07-01

    The Idaho National Engineering and Environmental Laboratory (INEEL), Nexant Inc. and the Oregon State University (OSU) have developed a Multi-Application Small Light Water Reactor (MASLWR) concept. The MASLWR is a small, safe and economic natural circulation pressurized light water reactor. MASLWR reactor module consists of an integral reactor/steam generator located in a steel cylindrical containment. The entire module is to be entirely shop fabricated and transported to site on most railways or roads. Two or more modules are located in a reactor building, each being submersed in a common, below grade cavity filled with water. For the most severe postulated accident, the volume of water in the cavity provides a passive ultimate heat sink for 3 or more days allowing the restoration of lost normal active heat removal systems. MASLWR thermal power of a single module is 150 MWt, primary system pressure 10.5 MPa, steam pressure 1.52 MPa and the net electrical output is 35 - 50 MWe. (authors)

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

    SciTech Connect

    Hongbin Zhang; Vincent Mousseau; Haihua Zhao

    2008-06-01

    Traditional nuclear reactor system analysis codes such as RELAP and TRAC employ an operator split methodology. In this approach, each of the physics (fluid flow, heat conduction and neutron diffusion) is solved separately and the coupling terms are done explicitly. This approach limits accuracy (first order in time at best) and makes the codes slow in running since the explicit coupling imposes stability restrictions on the time step size. These codes have been extensively tested and validated for the existing LWRs. However, for GEN IV nuclear reactor designs which tend to have long lasting transients resulting from passive safety systems, the performance is questionable and modern high fidelity simulation tools will be required. The requirement for accurate predictability is the motivation for a large scale overhaul of all of the models and assumptions in transient nuclear reactor safety simulation software. At INL we have launched an effort with the long term goal of developing a high fidelity system analysis code that employs modern physical models, numerical methods, and computer science for transient safety analysis of GEN IV nuclear reactors. Modern parallel solution algorithms will be employed through utilizing the nonlinear solution software package PETSc developed by Argonne National Laboratory. The physical models to be developed will have physically realistic length scales and time scales. The solution algorithm will be based on the physics-based preconditioned Jacobian-free Newton-Krylov solution methods. In this approach all of the physical models are solved implicitly and simultaneously in a single nonlinear system. This includes the coolant flow, nonlinear heat conduction, neutron kinetics, and thermal radiation, etc. Including modern physical models and accurate space and time discretizations will allow the simulation capability to be second order accurate in space and in time. This paper presents the current status of the development efforts as

  18. validation and Enhancement of Computational Fluid Dynamics and Heat Transfer Predictive Capabilities for Generation IV Reactor Systems

    SciTech Connect

    Robert E. Spall; Barton Smith; Thomas Hauser

    2008-12-08

    Nationwide, the demand for electricity due to population and industrial growth is on the rise. However, climate change and air quality issues raise serious questions about the wisdom of addressing these shortages through the construction of additional fossil fueled power plants. In 1997, the President's Committee of Advisors on Science and Technology Energy Research and Development Panel determined that restoring a viable nuclear energy option was essential and that the DOE should implement a R&D effort to address principal obstacles to achieving this option. This work has addressed the need for improved thermal/fluid analysis capabilities, through the use of computational fluid dynamics, which are necessary to support the design of generation IV gas-cooled and supercritical water reactors.

  19. Eugene P. Wigner's Visionary Contributions to Generations-I through IV Fission Reactors

    NASA Astrophysics Data System (ADS)

    Carré, Frank

    2014-09-01

    Among Europe's greatest scientists who fled to Britain and America in the 1930s, Eugene P. Wigner made instrumental advances in reactor physics, reactor design and technology, and spent nuclear fuel processing for both purposes of developing atomic weapons during world-war II and nuclear power afterwards. Wigner who had training in chemical engineering and self-education in physics first gained recognition for his remarkable articles and books on applications of Group theory to Quantum mechanics, Solid state physics and other topics that opened new branches of Physics.

  20. Analysis of supercritical CO{sub 2} cycle control strategies and dynamic response for Generation IV Reactors.

    SciTech Connect

    Moisseytsev, A.; Sienicki, J. J.

    2011-04-12

    The analysis of specific control strategies and dynamic behavior of the supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle has been extended to the two reactor types selected for continued development under the Generation IV Nuclear Energy Systems Initiative; namely, the Very High Temperature Reactor (VHTR) and the Sodium-Cooled Fast Reactor (SFR). Direct application of the standard S-CO{sub 2} recompression cycle to the VHTR was found to be challenging because of the mismatch in the temperature drop of the He gaseous reactor coolant through the He-to-CO{sub 2} reactor heat exchanger (RHX) versus the temperature rise of the CO{sub 2} through the RHX. The reference VHTR features a large temperature drop of 450 C between the assumed core outlet and inlet temperatures of 850 and 400 C, respectively. This large temperature difference is an essential feature of the VHTR enabling a lower He flow rate reducing the required core velocities and pressure drop. In contrast, the standard recompression S-CO{sub 2} cycle wants to operate with a temperature rise through the RHX of about 150 C reflecting the temperature drop as the CO{sub 2} expands from 20 MPa to 7.4 MPa in the turbine and the fact that the cycle is highly recuperated such that the CO{sub 2} entering the RHX is effectively preheated. Because of this mismatch, direct application of the standard recompression cycle results in a relatively poor cycle efficiency of 44.9%. However, two approaches have been identified by which the S-CO{sub 2} cycle can be successfully adapted to the VHTR and the benefits of the S-CO{sub 2} cycle, especially a significant gain in cycle efficiency, can be realized. The first approach involves the use of three separate cascaded S-CO{sub 2} cycles. Each S-CO{sub 2} cycle is coupled to the VHTR through its own He-to-CO{sub 2} RHX in which the He temperature is reduced by 150 C. The three respective cycles have efficiencies of 54, 50, and 44%, respectively, resulting in a net cycle

  1. Tritium permeation characterization of materials for fusion and generation IV very high temperature reactors

    SciTech Connect

    Thomson, S.; Pilatzke, K.; McCrimmon, K.; Castillo, I.; Suppiah, S.

    2015-03-15

    The objective of this work is to establish the tritium-permeation properties of structural alloys considered for Fusion systems and very high temperature reactors (VHTR). A description of the work performed to set up an apparatus to measure permeation rates of hydrogen and tritium in 304L stainless steel is presented. Following successful commissioning with hydrogen, the test apparatus was commissioned with tritium. Commissioning tests with tritium suggest the need for a reduction step that is capable of removing the oxide layer from the test sample surfaces before accurate tritium-permeation data can be obtained. Work is also on-going to clearly establish the temperature profile of the sample to correctly estimate the tritium-permeability data.

  2. Generation-IV Nuclear Energy Systems

    NASA Astrophysics Data System (ADS)

    McFarlane, Harold

    2008-05-01

    Nuclear power technology has evolved through roughly three generations of system designs: a first generation of prototypes and first-of-a-kind units implemented during the period 1950 to 1970; a second generation of industrial power plants built from 1970 to the turn of the century, most of which are still in operation today; and a third generation of evolutionary advanced reactors which began being built by the turn of the 20^th century, usually called Generation III or III+, which incorporate technical lessons learned through more than 12,000 reactor-years of operation. The Generation IV International Forum (GIF) is a cooperative international endeavor to develop advanced nuclear energy systems in response to the social, environmental and economic requirements of the 21^st century. Six Generation IV systems under development by GIF promise to enhance the future contribution and benefits of nuclear energy. All Generation IV systems aim at performance improvement, new applications of nuclear energy, and/or more sustainable approaches to the management of nuclear materials. High-temperature systems offer the possibility of efficient process heat applications and eventually hydrogen production. Enhanced sustainability is achieved primarily through adoption of a closed fuel cycle with reprocessing and recycling of plutonium, uranium and minor actinides using fast reactors. This approach provides significant reduction in waste generation and uranium resource requirements.

  3. Nuclear Energy Research Initiative Program (NERI) Quarterly Progress Report; New Design Equations for Swelling and Irradiation Creep in Generation IV Reactors

    SciTech Connect

    Wolfer, W G; Surh, M P; Garner, F A; Chrzan, D C; Schaldach, C; Sturgeon, J B

    2003-02-13

    The objectives of this research project are to significantly extend the theoretical foundation and the modeling of radiation-induced microstructural changes in structural materials used in Generation IV nuclear reactors, and to derive from these microstructure models the constitutive laws for void swelling, irradiation creep and stress-induced swelling, as well as changes in mechanical properties. The need for the proposed research is based on three major developments and advances over the past two decades. First, new experimental discoveries have been made on void swelling and irradiation creep which invalidate previous theoretical models and empirical constitutive laws for swelling and irradiation creep. Second, recent advances in computational methods and power make it now possible to model the complex processes of microstructure evolution over long-term neutron exposures. Third, it is now required that radiation-induced changes in structural materials over extended lifetimes be predicted and incorporated in the design of Generation IV reactors. Our approach to modeling and data analysis is a dual one in accord with both the objectives to simulate the evolution of the microstructure and to develop design equations for macroscopic properties. Validation of the models through data analysis is therefore carried out at both the microscopic and the macroscopic levels. For the microstructure models, we utilize the transmission electron microscopy results from steels irradiated in reactors and from model materials irradiated by neutrons as well as ion bombardments. The macroscopic constitutive laws will be tested and validated by analyzing density data, irradiation creep data, diameter changes of fuel elements, and post-irradiation tensile data. Validation of both microstructure models and macroscopic constitutive laws is a more stringent test of the internal consistency of the underlying science for radiation effects in structural materials for nuclear reactors.

  4. Incorporation of Passive Safety Systems in the Generation-IV Multi-Application Small Light Water Reactor (MASLWR)

    SciTech Connect

    Modro, S. Michael; Ougouag, Abderrafi M.; Fisher, James; Weaver, Kevan

    2002-07-01

    The Idaho National Engineering and Environmental Laboratory (INEEL), Nexant Inc. and the Oregon State University (OSU) developed an innovative Multi-Application Small Light Water Reactor (MASLWR) concept. The MASLWR is a small, modular, safe, and economic natural circulation light water reactor developed with the primary goal of producing electric power, but with the flexibility to be used for water desalination or district heating with deployment in a variety of locations. The MASLWR was developed, by design, to be a safe and economic reactor concept that can be deployed in the near term by utilizing current experience and capabilities of the industry. The key features of the MASLWR concept are the extreme simplicity of the design and its passive safety systems. This paper provides an overview of safety analyses performed for the MASLWR concept and explores potential for the increase in passive safety via the implementation of new features. The results of these safety studies demonstrate that the reactor core will be provided with a stable cooling source adequate to remove decay heat without significant cladding heatup under all credible scenarios. The response of the system to accident conditions is a controlled depressurization, whereby most of the primary system blowdown occurs via the submerged ADS blowdown pathway. (authors)

  5. Advanced Non-Destructive Assessment Technology to Determine the Aging of Silicon Containing Materials for Generation IV Nuclear Reactors

    NASA Astrophysics Data System (ADS)

    Koenig, T. W.; Olson, D. L.; Mishra, B.; King, J. C.; Fletcher, J.; Gerstenberger, L.; Lawrence, S.; Martin, A.; Mejia, C.; Meyer, M. K.; Kennedy, R.; Hu, L.; Kohse, G.; Terry, J.

    2011-06-01

    To create an in-situ, real-time method of monitoring neutron damage within a nuclear reactor core, irradiated silicon carbide samples are examined to correlate measurable variations in the material properties with neutron fluence levels experienced by the silicon carbide (SiC) during the irradiation process. The reaction by which phosphorus doping via thermal neutrons occurs in the silicon carbide samples is known to increase electron carrier density. A number of techniques are used to probe the properties of the SiC, including ultrasonic and Hall coefficient measurements, as well as high frequency impedance analysis. Gamma spectroscopy is also used to examine residual radioactivity resulting from irradiation activation of elements in the samples. Hall coefficient measurements produce the expected trend of increasing carrier concentration with higher fluence levels, while high frequency impedance analysis shows an increase in sample impedance with increasing fluence.

  6. Development and Validation of Temperature Dependent Thermal Neutron Scattering Laws for Applications and Safety Implications in Generation IV Reactor Designs

    SciTech Connect

    Ayman Hawari

    2008-06-20

    The overall obljectives of this project are to critically review the currently used thermal neutron scattering laws for various moderators as a function of temperature, select as well documented and representative set of experimental data sensitive to the neutron spectra to generate a data base of benchmarks, update models and models parameters by introducing new developments in thermalization theory and condensed matter physics into various computational approaches in establishing the scattering laws, benchmark the results against the experimentatl set. In the case of graphite, a validation experiment is performed by observing nutron slowing down as a function of temperatures equal to or greater than room temperature.

  7. Nuclear Data Needs for Generation IV Nuclear Energy Systems

    NASA Astrophysics Data System (ADS)

    Rullhusen, Peter

    2006-04-01

    Nuclear data needs for generation IV systems. Future of nuclear energy and the role of nuclear data / P. Finck. Nuclear data needs for generation IV nuclear energy systems-summary of U.S. workshop / T. A. Taiwo, H. S. Khalil. Nuclear data needs for the assessment of gen. IV systems / G. Rimpault. Nuclear data needs for generation IV-lessons from benchmarks / S. C. van der Marck, A. Hogenbirk, M. C. Duijvestijn. Core design issues of the supercritical water fast reactor / M. Mori ... [et al.]. GFR core neutronics studies at CEA / J. C. Bosq ... [et al]. Comparative study on different phonon frequency spectra of graphite in GCR / Young-Sik Cho ... [et al.]. Innovative fuel types for minor actinides transmutation / D. Haas, A. Fernandez, J. Somers. The importance of nuclear data in modeling and designing generation IV fast reactors / K. D. Weaver. The GIF and Mexico-"everything is possible" / C. Arrenondo Sánchez -- Benmarks, sensitivity calculations, uncertainties. Sensitivity of advanced reactor and fuel cycle performance parameters to nuclear data uncertainties / G. Aliberti ... [et al.]. Sensitivity and uncertainty study for thermal molten salt reactors / A. Biduad ... [et al.]. Integral reactor physics benchmarks- The International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPHEP) / J. B. Briggs, D. W. Nigg, E. Sartori. Computer model of an error propagation through micro-campaign of fast neutron gas cooled nuclear reactor / E. Ivanov. Combining differential and integral experiments on [symbol] for reducing uncertainties in nuclear data applications / T. Kawano ... [et al.]. Sensitivity of activation cross sections of the Hafnium, Tanatalum and Tungsten stable isotopes to nuclear reaction mechanisms / V. Avrigeanu ... [et al.]. Generating covariance data with nuclear models / A. J. Koning. Sensitivity of Candu-SCWR reactors physics calculations to nuclear data files / K. S

  8. IRIS Responsiveness to Generation IV Road-map Goals

    SciTech Connect

    Carelli, M.D.; Paramonov, D.V.; Petrovic, B.

    2002-07-01

    The DOE Generation IV road-map process is in its second and final year. Almost one hundred concepts submitted from all over the world have been reviewed against the Generation IV goals of resources sustainability; safety and reliability; and, economics. Advanced LWRs are taken as the reference point. IRIS (International Reactor Innovative and Secure), a 100-335 MWe integral light water reactor being developed by a vast international consortium led by Westinghouse, is one on the concepts being considered in the road-map and is perhaps the most visible representative of the concept set known as Integral Primary System Reactors (IPSR). This paper presents how IRIS satisfies the prescribed goals. The first goal of resource sustainability includes criteria like utilization of fuel resources, amount and toxicity of waste produced, environmental impact, proliferation and sabotage resistance. As a thermal reactor IRIS does not have the same fuel utilization as fast reactors. However, it has a significant flexibility in fuel cycles as it is designed to utilize either UO{sub 2} or MOX with straight burn cycles of 4 to 10 years, depending on the fissile content. High discharge burnup and Pu recycling result in good fuel utilization and lower waste; IRIS has also attractive proliferation resistance characteristics, due to the reduced accessibility of the fuel. The safety and reliability goal include reliability, workers' exposure, robust safety features, models with well characterized uncertainty, source term and mechanisms of energy release, robust mitigation of accidents. IRIS is significantly better than advanced LWRs because of its safety by design which eliminates a variety of accidents such as LOCAs, its containment vessel coupled design which maintains the core safely covered during the accident sequences, its design simplification features such as no (or reduced) soluble boron, internal shielding and four-year refueling/maintenance interval which significantly reduce

  9. STEAM GENERATOR FOR NUCLEAR REACTOR

    DOEpatents

    Kinyon, B.W.; Whitman, G.D.

    1963-07-16

    The steam generator described for use in reactor powergenerating systems employs a series of concentric tubes providing annular passage of steam and water and includes a unique arrangement for separating the steam from the water. (AEC)

  10. An Innovative Three-Dimensional Heterogeneous Coarse-Mesh Transport Method for Advanced and Generation IV Reactor Core Analysis and Design

    SciTech Connect

    Farzad Rahnema

    2009-11-12

    This project has resulted in a highly efficient method that has been shown to provide accurate solutions to a variety of 2D and 3D reactor problems. The goal of this project was to develop (1) an accurate and efficient three-dimensional whole-core neutronics method with the following features: based sollely on transport theory, does not require the use of cross-section homogenization, contains a highly accurate and self-consistent global flux reconstruction procedure, and is applicable to large, heterogeneous reactor models, and to (2) create new numerical benchmark problems for code cross-comparison.

  11. Steam Generator of the International Reactor Innovative and Secure

    SciTech Connect

    Cinotti, L.; Bruzzone, M.; Meda, N.; Corsini, G.; Lombardi, C.V.; Ricotti, M.; Conway, L.E.

    2002-07-01

    IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the main reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long-life core and enhanced safety to address the requirements defined by the US DOE for Generation IV reactors. The design of the steam generators, which are internally contained within the reactor vessel, is a major design effort in the development of the integral IRIS concept. The ongoing design activity about the steam generator is the subject of this paper. (authors)

  12. GEN IV reactors: Where we are, where we should go

    SciTech Connect

    Locatelli, G.; Mancini, M.; Todeschini, N.

    2012-07-01

    GEN IV power plants represent the mid-long term option of the nuclear sector. International literature proposes many papers and reports dealing with these reactors, but there is an evident difference of type and shape of information making impossible each kind of detailed comparison. Moreover, authors are often strongly involved in some particular design; this creates many difficulties in their super-partes position. Therefore it is necessary to put order in the most relevant information to understand strengths and weaknesses of each design and derive an overview useful for technicians and policy makers. This paper presents the state-of the art for GEN IV nuclear reactors providing a comprehensive literature review of the different designs with a relate taxonomy. It presents the more relevant references, data, advantages, disadvantages and barriers to the adoptions. In order to promote an efficient and wide adoption of GEN IV reactors the paper provides the pre-conditions that must be accomplished, enabling factors promoting the implementation and barriers limiting the extent and intensity of its implementation. It concludes outlying the state of the art of the most important R and D areas and the future achievements that must be accomplished for a wide adoption of these technologies. (authors)

  13. Structural materials for Gen-IV nuclear reactors: Challenges and opportunities

    NASA Astrophysics Data System (ADS)

    Murty, K. L.; Charit, I.

    2008-12-01

    Generation-IV reactor design concepts envisioned thus far cater toward a common goal of providing safer, longer lasting, proliferation-resistant and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-IV reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses and extremely corrosive environment, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This paper presents a summary of various Gen-IV reactor concepts, with emphasis on the structural materials issues depending on the specific application areas. This paper also discusses the challenges involved in using the existing materials under both service and off-normal conditions. Tasks become increasingly complex due to the operation of various fundamental phenomena like radiation-induced segregation, radiation-enhanced diffusion, precipitation, interactions between impurity elements and radiation-produced defects, swelling, helium generation and so forth. Further, high temperature capability (e.g. creep properties) of these materials is a critical, performance-limiting factor. It is demonstrated that novel alloy and microstructural design approaches coupled with new materials processing and fabrication techniques may mitigate the challenges, and the optimum system performance may be achieved under much demanding conditions.

  14. An Integrated Safety Assessment Methodology for Generation IV Nuclear Systems

    SciTech Connect

    Timothy J. Leahy

    2010-06-01

    The Generation IV International Forum (GIF) Risk and Safety Working Group (RSWG) was created to develop an effective approach for the safety of Generation IV advanced nuclear energy systems. Early work of the RSWG focused on defining a safety philosophy founded on lessons learned from current and prior generations of nuclear technologies, and on identifying technology characteristics that may help achieve Generation IV safety goals. More recent RSWG work has focused on the definition of an integrated safety assessment methodology for evaluating the safety of Generation IV systems. The methodology, tentatively called ISAM, is an integrated “toolkit” consisting of analytical techniques that are available and matched to appropriate stages of Generation IV system concept development. The integrated methodology is intended to yield safety-related insights that help actively drive the evolving design throughout the technology development cycle, potentially resulting in enhanced safety, reduced costs, and shortened development time.

  15. DEVELOPMENT OF RISK-BASED AND TECHNOLOGY-INDEPENDENT SAFETY CRITERIA FOR GENERATION IV SYSTEMS

    SciTech Connect

    William E. Kastenberg; Edward Blandford; Lance Kim

    2009-03-31

    This project has developed quantitative safety goals for Generation IV (Gen IV) nuclear energy systems. These safety goals are risk based and technology independent. The foundations for a new approach to risk analysis has been developed, along with a new operational definition of risk. This project has furthered the current state-of-the-art by developing quantitative safety goals for both Gen IV reactors and for the overall Gen IV nuclear fuel cycle. The risk analysis approach developed will quantify performance measures, characterize uncertainty, and address a more comprehensive view of safety as it relates to the overall system. Appropriate safety criteria are necessary to manage risk in a prudent and cost-effective manner. This study is also important for government agencies responsible for managing, reviewing, and for approving advanced reactor systems because they are charged with assuring the health and safety of the public.

  16. Structural materials issues for the next generation fission reactors

    NASA Astrophysics Data System (ADS)

    Chant, I.; Murty, K. L.

    2010-09-01

    Generation-IV reactor design concepts envisioned thus far cater to a common goal of providing safer, longer lasting, proliferation-resistant, and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-W reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses, and extremely corrosive environments, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This article addresses the material requirements for these advanced fission reactor types, specifically addressing structural materials issues depending on the specific application areas.

  17. Critical Issues on Materials for Gen-IV Reactors

    SciTech Connect

    Caro, M; Marian, J; Martinez, E; Erhart, P

    2009-02-27

    Within the LDRD on 'Critical Issues on Materials for Gen-IV Reactors' basic thermodynamics of the Fe-Cr alloy and accurate atomistic modeling were used to help develop the capability to predict hardening, swelling and embrittlement using the paradigm of Multiscale Materials Modeling. Approaches at atomistic and mesoscale levels were linked to build-up the first steps in an integrated modeling platform that seeks to relate in a near-term effort dislocation dynamics to polycrystal plasticity. The requirements originated in the reactor systems under consideration today for future sources of nuclear energy. These requirements are beyond the present day performance of nuclear materials and calls for the development of new, high temperature, radiation resistant materials. Fe-Cr alloys with 9-12% Cr content are the base matrix of advanced ferritic/martensitic (FM) steels envisaged as fuel cladding and structural components of Gen-IV reactors. Predictive tools are needed to calculate structural and mechanical properties of these steels. This project represents a contribution in that direction. The synergy between the continuous progress of parallel computing and the spectacular advances in the theoretical framework that describes materials have lead to a significant advance in our comprehension of materials properties and their mechanical behavior. We took this progress to our advantage and within this LDRD were able to provide a detailed physical understanding of iron-chromium alloys microstructural behavior. By combining ab-initio simulations, many-body interatomic potential development, and mesoscale dislocation dynamics we were able to describe their microstructure evolution. For the first time in the case of Fe-Cr alloys, atomistic and mesoscale were merged and the first steps taken towards incorporating ordering and precipitation effects into dislocation dynamics (DD) simulations. Molecular dynamics (MD) studies of the transport of self-interstitial, vacancy and

  18. High Temperature Irradiation Effects in Selected Generation IV Structural Alloys

    SciTech Connect

    Nanstad, Randy K; McClintock, David A; Hoelzer, David T; Tan, Lizhen; Allen, Todd R.

    2009-01-01

    In the Generation IV Materials Program cross-cutting task, irradiation and testing were carried out to address the issue of high temperature irradiation effects with selected current and potential candidate metallic alloys. The materials tested were (1) a high-nickel iron-base alloy (Alloy 800H); (2) a nickel-base alloy (Alloy 617); (3) two advanced nano-structured ferritic alloys (designated 14YWT and 14WT); and (4) a commercial ferritic-martensitic steel (annealed 9Cr-1MoV). Small tensile specimens were irradiated in rabbit capsules in the High-Flux Isotope Reactor at temperatures from about 550 to 700 C and to irradiation doses in the range 1.2 to 1.6 dpa. The Alloy 800H and Alloy 617 exhibited significant hardening after irradiation at 580 C; some hardening occurred at 660 C as well, but the 800H showed extremely low tensile elongations when tested at 700 C. Notably, the grain boundary engineered 800H exhibited even greater hardening at 580 C and retained a high amount of ductility. Irradiation effects on the two nano-structured ferritic alloys and the annealed 9Cr-1MoV were relatively slight at this low dose.

  19. Modelling of advanced structural materials for GEN IV reactors

    NASA Astrophysics Data System (ADS)

    Samaras, M.; Hoffelner, W.; Victoria, M.

    2007-09-01

    The choice of suitable materials and the assessment of long-term materials damage are key issues that need to be addressed for the safe and reliable performance of nuclear power plants. Operating conditions such as high temperatures, irradiation and a corrosive environment degrade materials properties, posing the risk of very expensive or even catastrophic plant damage. Materials scientists are faced with the scientific challenge to determine the long-term damage evolution of materials under service exposure in advanced plants. A higher confidence in life-time assessments of these materials requires an understanding of the related physical phenomena on a range of scales from the microscopic level of single defect damage effects all the way up to macroscopic effects. To overcome lengthy and expensive trial-and-error experiments, the multiscale modelling of materials behaviour is a promising tool, bringing new insights into the fundamental understanding of basic mechanisms. This paper presents the multiscale modelling methodology which is taking root internationally to address the issues of advanced structural materials for Gen IV reactors.

  20. On reactor type comparisons for the next generation of reactors

    SciTech Connect

    Alesso, H.P.; Majumdar, K.C.

    1991-08-22

    In this paper, we present a broad comparison of studies for a selected set of parameters for different nuclear reactor types including the next generation. This serves as an overview of key parameters which provide a semi-quantitative decision basis for selecting nuclear strategies. Out of a number of advanced reactor designs of the LWR type, gas cooled type, and FBR type, currently on the drawing board, the Advanced Light Water Reactors (ALWR) seem to have some edge over other types of the next generation of reactors for the near-term application. This is based on a number of attributes related to the benefit of the vast operating experience with LWRs coupled with an estimated low risk profile, economics of scale, degree of utilization of passive systems, simplification in the plant design and layout, modular fabrication and manufacturing. 32 refs., 1 fig., 3 tabs.

  1. Heat-generating nuclear reactor

    SciTech Connect

    Dupuy, G.; Fajeau, M.; Labrousse, M.; Lerouge, B.; Minguet, J.

    1981-01-20

    A reactor vessel filled with coolant fluid is divided by a wall into an upper region and a lower region which contains the reactor core, part of the coolant fluid in the upper region being injected into the lower region. The injection flow rate is regulated as a function of the variations in pressure in the lower region by means of a baffle-plate container which communicates with a leak-tight chamber and with a storage reservoir, a flow of fluid from the chamber to the reservoir being established only at the time of a reduction in the rate of injection into the container. The reactor can be employed for the production of hot water which is passed through a heat exchanger and supplied to a heating installation.

  2. POWER GENERATING NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Vernon, H.C.

    1958-03-01

    This patent relates to reactor systems of the type wherein the cooiing medium is a liquid which is converted by the heat of the reaction to steam which is conveyed directly to a pnime mover such as a steam turbine driving a generatore after which it is condensed and returred to the coolant circuit. In this design, the reactor core is disposed within a tank for containing either a slurry type fuel or an aggregation of solid fuel elements such as elongated rods submerged in a liquid moderator such as heavy water. The top of the tank is provided with a nozzle which extends into an expansion chamber connected with the upper end of the tank, the coolant being maintained in the expansion chamber at a level above the nozzle and the steam being formed in the expansion chamber.

  3. Automatic generation and analysis of solar cell IV curves

    DOEpatents

    Kraft, Steven M.; Jones, Jason C.

    2014-06-03

    A photovoltaic system includes multiple strings of solar panels and a device presenting a DC load to the strings of solar panels. Output currents of the strings of solar panels may be sensed and provided to a computer that generates current-voltage (IV) curves of the strings of solar panels. Output voltages of the string of solar panels may be sensed at the string or at the device presenting the DC load. The DC load may be varied. Output currents of the strings of solar panels responsive to the variation of the DC load are sensed to generate IV curves of the strings of solar panels. IV curves may be compared and analyzed to evaluate performance of and detect problems with a string of solar panels.

  4. Improving proliferation resistance of high breeding gain generation 4 reactors using blankets composed of light water reactor waste

    SciTech Connect

    Hellesen, C.; Grape, S.; Haakanson, A.; Jacobson Svaerd, S.; Jansson, P.

    2013-07-01

    Fertile blankets can be used in fast reactors to enhance the breeding gain as well as the passive safety characteristics. However, such blankets typically result in the production of weapons grade plutonium. For this reason they are often excluded from Generation IV reactor designs. In this paper we demonstrate that using blankets manufactured directly from spent light water (LWR) reactor fuel it is possible to produce a plutonium product with non-proliferation characteristics on a par with spent LWR fuel of 30-50 MWd/kg burnup. The beneficial breeding and safety characteristics are retained. (authors)

  5. Progress reports for Gen IV sodium fast reactor activities FY 2007.

    SciTech Connect

    Cahalan, J. E.; Tentner, A. M.; Nuclear Engineering Division

    2007-10-04

    An important goal of the US DOE Sodium Fast Reactor (SFR) program is to develop the technology necessary to increase safety margins in future fast reactor systems. Although no decision has been made yet about who will build the next demonstration fast reactor, it seems likely that the construction team will include a combination of international companies, and the safety design philosophy for the reactor will reflect a consensus of the participating countries. A significant amount of experience in the design and safety analysis of Sodium Fast Reactors (SFR) using oxide fuel has been developed in both Japan and France during last few decades. In the US, the traditional approach to reactor safety is based on the principle of defense-in-depth, which is usually expressed in physical terms as multiple barriers to release of radioactive material (e.g. cladding, reactor vessel, containment building), but it is understood that the 'barriers' may consist of active systems or even procedures. As implemented in a reactor design, defense-in-depth is classed in levels of safety. Level 1 includes measures to specify and build a reliable design with significant safety margins that will perform according to the intentions of the designers. Level 2 consists of additional design measures, usually active systems, to protect against unlikely accidental events that may occur during the life of the plant. Level 3 design measures are intended to protect the public in the event of an extremely unlikely accident not foreseen to occur during the plant's life. All of the design measures that make up the first three levels of safety are within the design basis of the plant. Beyond Level 3, and beyond the normal design basis, there are accidents that are not expected to occur in a whole generation of plants, and it is in this class that severe accidents, i.e. accidents involving core melting, are included. Beyond design basis measures to address severe accidents are usually identified as being

  6. A Virtual Reality Framework to Optimize Design, Operation and Refueling of GEN-IV Reactors.

    SciTech Connect

    Rizwan-uddin; Nick Karancevic; Stefano Markidis; Joel Dixon; Cheng Luo; Jared Reynolds

    2008-04-23

    many GEN-IV candidate designs are currently under investigation. Technical issues related to material, safety and economics are being addressed at research laboratories, industry and in academia. After safety, economic feasibility is likely to be the most important crterion in the success of GEN-IV design(s). Lessons learned from the designers and operators of GEN-II (and GEN-III) reactors must play a vital role in achieving both safety and economic feasibility goals.

  7. Flexible Fuel Cycle Initiative for the Harmonized Deployment of Gen-IV Reactors

    NASA Astrophysics Data System (ADS)

    Fukasawa, Tetsuo; Yamashita, Junichi; Hoshino, Kuniyoshi; Fujimura, Koji; Sasahira, Akira

    Generation IV type fast reactors (FR) are expected to be commercially deployed instead of light water reactors (LWR) from around 2050. Replacement of LWR to FR needs flexibility due to uncertain factors such as FR deployment rate which affects the FR fuel (Pu) supply amount from LWR spent fuel reprocessing and the capacity of related facilities. If the FR deployment rate is as currently planned, more Pu must be prepared by expanding LWR reprocessing. If the FR deployment rate decreases, LWR reprocessing must be reduced to avoid excess Pu. To cope with this issue we proposed the innovative system called Flexible Fuel Cycle Initiative (FFCI) that has integral reprocessing for LWR and FR spent fuels. LWR reprocessing in FFCI only carries out about 90% U recovery and residual material with Pu, U (˜5%), minor actinides (MA) and fission products (FP) goes to FR reprocessing for the planned FR deployment rate. For any decrease in the FR deployment rate temporary storage will be used. Coexistence of Pu/U with MA and FP until just before Pu/U usage in the FR provides high proliferation resistance. Preliminary evaluation revealed that FFCI can reduce the LWR reprocessing capacity and LWR spent fuel storage amount compared with current plan (reference system) if the FR deployment rate decreases. Several FR deployment scenarios and countermeasures such as FFCI were investigated.

  8. Electrical I-V Probes Calibration And Reactor Plasma Chamber Characterization

    SciTech Connect

    Tadjine, R.; Lahmar, H.

    2008-09-23

    We have constructed a high resolution I-V Probes including a high voltage and current probes to measure the current and the tension at the entry of the electrode excitation in reactor plasma RF argon. The I-V Probes is no intrusive, therefore it can be transposable in another surface treatment reactor, and it is what is required in the industrial applications. Because of capacitive aspect of the coupling and the presence of parasitic electrical elements in the chamber plasma reactor and the connection circuits, the measured signals are largely different from the signals at the RF electrode in contact with plasma. A phase angle error of 1 deg. between i(t) and v(t) introduce a large error in the calculated parameters. The objective of this work is the calibration of the I-V Probes and the whole system of measurement, and the characterization of the parasitic elements of the reactor. The characterization of the I-V Probes required the realization of several reactive loads (capacitive) measured with a network analyzer HP8753ES. The signals Vi(t) and Vv(t), resulting from the I-V Probes and collected by a DPO (TDS4054 numerical oscilloscope of very high performance), are transferred towards the computer for a digital processing. The elimination of the offsets voltage present at the entries of the DPO and the smoothing by a powerful method of least squares, make it possible to increase precision.

  9. Plasma generators, reactor systems and related methods

    DOEpatents

    Kong, Peter C.; Pink, Robert J.; Lee, James E.

    2007-06-19

    A plasma generator, reactor and associated systems and methods are provided in accordance with the present invention. A plasma reactor may include multiple sections or modules which are removably coupled together to form a chamber. Associated with each section is an electrode set including three electrodes with each electrode being coupled to a single phase of a three-phase alternating current (AC) power supply. The electrodes are disposed about a longitudinal centerline of the chamber and are arranged to provide and extended arc and generate an extended body of plasma. The electrodes are displaceable relative to the longitudinal centerline of the chamber. A control system may be utilized so as to automatically displace the electrodes and define an electrode gap responsive to measure voltage or current levels of the associated power supply.

  10. Safety of New Generation Concepts in Reference to Present Reactors

    SciTech Connect

    Rouyer, Jean-Loup; Vitton, Francis

    2004-07-01

    Present operational nuclear reactors have reached a high level of safety and reliability. Main safety principles which have allowed to obtain excellent present performances must remain the cornerstone for future projects, with adaptations on points where operating feedback shows needs for reinforcement or simplification. These basic principles are defense in depth concept and probabilistic quantification. Criteria for which specific emphasis should be put for the future are: Inertia of processes, important parameter contributing to stability, applying to neutronics as well as coolant fluids. Barriers, whose number must not be rigidly fixed, but determined by independence and margins considerations. Hazards, risk induced by external and internal hazards which should be at least, comparable to the one induced by internal accidents. On the basis of these safety criteria, the paper analyses the potentials of several reactor concepts for the future, as compared with advanced existing reactors. The six concepts selected by the Generation IV Forum (VHTR, GFR, SFR, LFR, SCWR, MSR) are appreciated. The MHTGR project which has been evaluated by the US-NRC is also considered as a reference for gas-cooled advanced concepts. (authors)

  11. Generation IV PR and PP Methods and Applications

    SciTech Connect

    Bari,R.A.

    2008-10-13

    This paper presents an evaluation methodology for proliferation resistance and physical protection (PR&PP) of Generation IV nuclear energy systems (NESs). For a proposed NES design, the methodology defines a set of challenges, analyzes system response to these challenges, and assesses outcomes. The challenges to the NES are the threats posed by potential actors (proliferant States or sub-national adversaries). The characteristics of Generation IV systems, both technical and institutional, are used to evaluate the response of the system and determine its resistance against proliferation threats and robustness against sabotage and terrorism threats. The outcomes of the system response are expressed in terms of six measures for PR and three measures for PP, which are the high-level PR&PP characteristics of the NES. The methodology is organized to allow evaluations to be performed at the earliest stages of system design and to become more detailed and more representative as design progresses. Uncertainty of results are recognized and incorporated into the evaluation at all stages. The results are intended for three types of users: system designers, program policy makers, and external stakeholders. Particular current relevant activities will be discussed in this regard. The methodology has been illustrated in a series of demonstration and case studies and these will be summarized in the paper.

  12. General Point-Depletion and Fission Product Code System and Four-Group Fission Product Neutron Absorption Chain Data Library Generated from ENDF/B-IV for Thermal Reactors

    1981-12-01

    EPRI-CINDER calculates, for any specified initial fuel (actinide) description and flux or power history, the fuel and fission-product nuclide concentrations and associated properties. Other nuclide chains can also be computed with user-supplied libraries. The EPRI-CINDER Data Library (incorporating ENDF/B-IV fission-product processed 4-group cross sections, decay constants, absorption and decay branching fractions, and effective fission yields) is used in each constant-flux time step calculation and in time step summaries of nuclide decay rates and macroscopic absorptionmore » and barns-per-fission (b/f) absorption cross sections (by neutron group). User-supplied nuclide decay energy and multigroup-spectra data libraries may be attached to permit decay heating and decay-spectra calculations. An additional 12-chain library, explicitly including 27 major fission-product neutron absorbers and 4 fictitious nuclides, may be used to accurately calculate the aggregate macroscopic absorption buildup in fission products.« less

  13. A Qualitative Assessment of Diversion Scenarios for an Example Sodium Fast Reactor Using the GEN IV PR&PP Methodology

    SciTech Connect

    Zentner, Michael D.; Coles, Garill A.; Therios, Ike

    2012-01-20

    FAST REACTORS;NUCLEAR ENERGY;NUCLEAR MATERIALS MANAGEMENT;PROLIFERATION;SAFEGUARDS;THEFT; A working group was created in 2002 by the Generation IV International Forum (GIF) for the purpose of developing an internationally accepted methodology for assessing the Proliferation Resistance of a nuclear energy system (NES) and its individual elements. A two year case study is being performed by the experts group using this methodology to assess the proliferation resistance of a hypothetical NES called the Example Sodium Fast Reactor (ESFR). This work demonstrates how the PR and PP methodology can be used to provide important information at various levels of details to NES designers, safeguard administrators and decision makers. The study analyzes the response of the complete ESFR nuclear energy system to different proliferation and theft strategies. The challenges considered include concealed diversion, concealed misuse and 'break out' strategies. This paper describes the work done in performing a qualitative assessment of concealed diversion scenarios from the ESFR.

  14. Fusion-fission hybrids for nuclear waste transmutation : a synergistic step between Gen-IV fission and fusion reactors.

    SciTech Connect

    Olson, Craig Lee; Mehlhorn, Thomas Alan; Cipiti, Benjamin B.; Rochau, Gary Eugene

    2007-09-01

    Energy demand and GDP per capita are strongly correlated, while public concern over the role of energy in climate change is growing. Nuclear power plants produce 16% of world electricity demands without greenhouse gases. Generation-IV advanced nuclear energy systems are being designed to be safe and economical. Minimizing the handling and storage of nuclear waste is important. NIF and ITER are bringing sustainable fusion energy closer, but a significant gap in fusion technology development remains. Fusion-fission hybrids could be a synergistic step to a pure fusion economy and act as a technology bridge. We discuss how a pulsed power-driven Z-pinch hybrid system producing only 20 MW of fusion yield can drive a sub-critical transuranic blanket that transmutes 1280 kg of actinide wastes per year and produces 3000 MW. These results are applicable to other inertial and magnetic fusion energy systems. A hybrid system could be introduced somewhat sooner because of the modest fusion yield requirements and can provide both a safe alternative to fast reactors for nuclear waste transmutation and a maturation path for fusion technology. The development and demonstration of advanced materials that withstand high-temperature, high-irradiation environments is a fundamental technology issue that is common to both fusion-fission hybrids and Generation-IV reactors.

  15. Selection of materials for sodium fast reactor steam generators

    SciTech Connect

    Dubiez-Le Goff, S.; Garnier, S.; Gelineau, O.; Dalle, F.; Blat-Yrieix, M.; Augem, J. M.

    2012-07-01

    Sodium Fast Reactor (SFR) is considered in France as the most mature technology of the different Generation IV systems. In the short-term the designing work is focused on the identification of the potential tracks to demonstrate licensing capability, availability, in-service inspection capability and economical performance. In that frame materials selection for the major components, as the steam generator, is a particularly key point managed within a French Research and Development program launched by AREVA, CEA and EDF. The choice of the material for the steam generator is indeed complex because various aspects shall be considered like mechanical and thermal properties at high temperature, interaction with sodium on one side and water and steam on the other side, resistance to wastage, procurement, fabrication, weldability and ability for inspection and in-situ intervention. The following relevant options are evaluated: the modified 9Cr1Mo ferritic-martensitic grade and the Alloy 800 austenitic grade. The objective of this paper is to assess for both candidates their abilities to reach the current SFR needs regarding material design data, from AFCEN RCC-MRx Code in particular, compatibility with environments and manufacturability. (authors)

  16. Generation IV nuclear energy systems and the need of accurate nuclear data

    NASA Astrophysics Data System (ADS)

    Colonna, N.

    2009-05-01

    To satisfy the world's demand of energy, constantly increasing over the years, a suitable mix of different energy sources has to be envisaged. In this scenario, an important role may be played by nuclear energy, provided that major safety, waste and proliferation issues affecting current nuclear reactors are satisfactorily addressed. In this respect, a large effort is under way since a few years towards the development of advanced nuclear systems that would use more efficiently the uranium resources, and produce a minimal amount of long-lived nuclear waste. The main activity concerns Generation IV reactors, with full or partial waste recycling capability. Their design requires R&D in numerous fields. Among the different needs, it is of fundamental importance to improve the knowledge of basic nuclear data, such as cross-sections for neutron-induced reactions on actinides. The main characteristics and principle of operation of the new generation nuclear systems are here described, together with the related needs of new and accurate nuclear data. Finally, an example of activity currently undergoing in the field is shown, with the recent experimental results obtained at the neutron facility n_TOF at CERN.

  17. Generating unstructured nuclear reactor core meshes in parallel

    SciTech Connect

    Jain, Rajeev; Tautges, Timothy J.

    2014-10-24

    Recent advances in supercomputers and parallel solver techniques have enabled users to run large simulations problems using millions of processors. Techniques for multiphysics nuclear reactor core simulations are under active development in several countries. Most of these techniques require large unstructured meshes that can be hard to generate in a standalone desktop computers because of high memory requirements, limited processing power, and other complexities. We have previously reported on a hierarchical lattice-based approach for generating reactor core meshes. Here, we describe efforts to exploit coarse-grained parallelism during reactor assembly and reactor core mesh generation processes. We highlight several reactor core examples including a very high temperature reactor, a full-core model of the Korean MONJU reactor, a ¼ pressurized water reactor core, the fast reactor Experimental Breeder Reactor-II core with a XX09 assembly, and an advanced breeder test reactor core. The times required to generate large mesh models, along with speedups obtained from running these problems in parallel, are reported. A graphical user interface to the tools described here has also been developed.

  18. Generating unstructured nuclear reactor core meshes in parallel

    DOE PAGES

    Jain, Rajeev; Tautges, Timothy J.

    2014-10-24

    Recent advances in supercomputers and parallel solver techniques have enabled users to run large simulations problems using millions of processors. Techniques for multiphysics nuclear reactor core simulations are under active development in several countries. Most of these techniques require large unstructured meshes that can be hard to generate in a standalone desktop computers because of high memory requirements, limited processing power, and other complexities. We have previously reported on a hierarchical lattice-based approach for generating reactor core meshes. Here, we describe efforts to exploit coarse-grained parallelism during reactor assembly and reactor core mesh generation processes. We highlight several reactor coremore » examples including a very high temperature reactor, a full-core model of the Korean MONJU reactor, a ¼ pressurized water reactor core, the fast reactor Experimental Breeder Reactor-II core with a XX09 assembly, and an advanced breeder test reactor core. The times required to generate large mesh models, along with speedups obtained from running these problems in parallel, are reported. A graphical user interface to the tools described here has also been developed.« less

  19. Thermal stability study for candidate stainless steels of GEN IV reactors

    NASA Astrophysics Data System (ADS)

    Simeg Veternikova, J.; Degmova, J.; Pekarcikova, M.; Simko, F.; Petriska, M.; Skarba, M.; Mikula, P.; Pupala, M.

    2016-11-01

    Candidate stainless steels for GEN IV reactors were investigated in term of thermal and corrosion stability at high temperatures. New austenitic steel (NF 709), austenitic ODS steel (ODS 316) and two ferritic ODS steels (MA 956 and MA 957) were exposed to around 1000 °C in inert argon atmosphere at pressure of ∼8 MPa. The steels were further studied in a light of vacancy defects presence by positron annihilation spectroscopy and their thermal resistance was confronted to classic AISI steels. The thermal strain supported a creation of oxide layers observed by scanning electron microscopy (SEM).

  20. Irradiation effects in oxide dispersion strengthened (ODS) Ni-base alloys for Gen. IV nuclear reactors

    NASA Astrophysics Data System (ADS)

    Oono, Naoko; Ukai, Shigeharu; Kondo, Sosuke; Hashitomi, Okinobu; Kimura, Akihiko

    2015-10-01

    Oxide particle dispersion strengthened (ODS) Ni-base alloys are irradiated by using simulation technique (Fe/He dual-ion irradiation) to investigate the reliability to Gen. IV high-temperature reactors. The fine oxide particles with less than 10 nm in average size and approximately 8.0 × 1022 m-3 in number density remained after 101 dpa irradiation. The tiny helium bubbles were inside grains, not at grain-boundaries; it is advantageous effect of oxide particles which trap the helium atoms at the particle-matrix interface. Ni-base ODS alloys demonstrated their great ability to overcome He embrittlement.

  1. Emergency Decay Heat Removal in a GEN-IV Gas-Cooled Fast Reactor

    SciTech Connect

    Cheng, Lap Y.; Ludewig, Hans; Jo, Jae

    2006-07-01

    A series of transient analyses using the system code RELAP5-3d has been performed to confirm the efficacy of a proposed hybrid active/passive combination approach to the decay heat removal for an advanced 2400 MWt GEN-IV gas-cooled fast reactor. The accident sequence of interest is a station blackout simultaneous with a small break (10 sq.inch/0.645 m{sup 2}) in the reactor vessel. The analyses cover the three phases of decay heat removal in a depressurization accident: (1) forced flow cooling by the power conversion unit (PCU) coast down, (2) active forced flow cooling by a battery powered blower, and (3) passive cooling by natural circulation. The blower is part of an emergency cooling system (ECS) that by design is to sustain passive decay heat removal via natural circulation cooling 24 hours after shutdown. The RELAP5 model includes the helium-cooled reactor, the ECS (primary and secondary side), the PCU with all the rotating machinery (turbine and compressors) and the heat transfer components (recuperator, pre-cooler and inter-cooler), and the guard containment that surrounds the reactor and the PCU. The transient analysis has demonstrated the effectiveness of passive decay heat removal by natural circulation cooling when the guard containment pressure is maintained at or above 800 kPa. (authors)

  2. Evaluation of Thermoelectric Generators by I-V Curves

    NASA Astrophysics Data System (ADS)

    Min, Gao; Singh, Tanuj; Garcia-Canadas, Jorge; Ellor, Robert

    2016-03-01

    A recent theoretical study proposes a new way to evaluate thermoelectric devices by measuring two I-V curves—one obtained under a constant temperature difference and the other obtained for a constant thermal input. We report an experimental demonstration of the feasibility of this novel technique. A measurement system was designed and constructed, which enables both types of I-V curves to be obtained automatically. The effective ZT values of a thermoelectric module were determined using this system and compared with those measured by an impedance spectroscopy technique. The results confirm the validity of the proposed technique. In addition, the capability of measuring ZT under a large temperature difference was also investigated. The results show that the ZTs obtained for a large temperature difference are significantly smaller than those for a small temperature difference, providing insights into the design and operation of thermoelectric modules in realistic applications.

  3. Lead-cooled system design and challenges in the frame of Generation IV International Forum

    NASA Astrophysics Data System (ADS)

    Cinotti, Luciano; Smith, Craig F.; Sekimoto, Hiroshi; Mansani, Luigi; Reale, Marco; Sienicki, James J.

    2011-08-01

    The Generation IV International Forum (GIF) Technology Roadmap identified the Lead-cooled Fast Reactor (LFR) as a technology well suited for electricity generation, hydrogen production and actinide management in a closed fuel cycle. One of the most important features of the LFR is the fact that lead is a relatively inert coolant, a feature that conveys significant advantages in terms of safety, system simplification, and the consequent potential for economic performance. In 2004, the GIF LFR Provisional System Steering Committee was organized and began to develop the LFR System Research Plan. The committee selected two pool-type reactor concepts as candidates for international cooperation and joint development in the GIF framework: these are the Small Secure Transportable Autonomous Reactor (SSTAR); and the European Lead-cooled System (ELSY). The high boiling point (1745 °C) of lead has a beneficial impact to the safety of the system, whereas its high melting point (327.4 °C) requires new engineering strategies, especially for In-Service-Inspection and refuelling. Lead, especially at high temperatures, is also relatively corrosive towards structural materials. This necessitates that coolant purity and the level of dissolved oxygen be carefully controlled, in addition to the proper selection of structural materials. For the GIF LFR concepts, lead has been chosen as the coolant rather than Lead-Bismuth Eutectic primarily because of its greatly reduced generation of the alpha-emitting 210Po isotope formed in the coolant. This results in significantly reduced levels of radioactive contamination of the coolant while minimizing the effect of decay power in the coolant from such contaminants; an additional consideration is the desire to eliminate dependence on bismuth which might be a limited resource. This paper provides an overview of the historical development of the LFR, a summary of the advantages and challenges associated with heavy liquid metal coolants, and an

  4. Millimeter-Wave Thermal Analysis Development and Application to GEN IV Reactor Materials

    SciTech Connect

    Wosko, Paul; Sundram, S. K.

    2012-10-16

    New millimeter-wave thermal analysis instrumentation has been developed and studied for characterization of materials required for diverse fuel and structural needs in high temperature reactor environments such as the Next Generation Nuclear Plant (NGNP). A two-receiver 137 GHz system with orthogonal polarizations for anisotropic resolution of material properties has been implemented at MIT. The system was tested with graphite and silicon carbide specimens at temperatures up to 1300 ºC inside an electric furnace. The analytic and hardware basis for active millimeter-wave radiometry of reactor materials at high temperature has been established. Real-time, non contact measurement sensitivity to anisotropic surface emissivity and submillimeter surface displacement was demonstrated. The 137 GHz emissivity of reactor grade graphite (NBG17) from SGL Group was found to be low, ~ 5 %, in the 500 – 1200 °C range and increases by a factor of 2 to 4 with small linear grooves simulating fracturing. The low graphite emissivity would make millimeter-wave active radiometry a sensitive diagnostic of graphite changes due to environmentally induced stress fracturing, swelling, or corrosion. The silicon carbide tested from Ortek, Inc. was found to have a much higher emissivity at 137 GHz of ~90% Thin coatings of silicon carbide on reactor grade graphite supplied by SGL Group were found to be mostly transparent to millimeter-waves, increasing the 137 GHz emissivity of the coated reactor grade graphite to about ~14% at 1250 ºC.

  5. Uranium droplet nuclear reactor core with MHD generator

    NASA Astrophysics Data System (ADS)

    Anghaie, Samim; Kumar, Ratan

    An innovative concept employing liquid uranium droplets as fuel in an ultrahigh-temperature vapor core reactor (UTVR) magnetohydrodynamic (MHD) generator power system for space power generation has been studied. Metallic vapor in superheated form acts as a working fluid for a closed-Rankine-type thermodynamic cycle. Usage of fuel and working fluid in this form assures certain advantages. The major technical issues emerging as a result involve a method for droplet generation, droplet transport in the reactor core, heat generation in the fuel and transport to the metallic vapor, and materials compatibility. A qualitative and quantitative attempt to resolve these issues has indicated the promise and tentative feasibility of the system.

  6. Group IV nanotube transistors for next generation ubiquitous computing

    NASA Astrophysics Data System (ADS)

    Fahad, Hossain M.; Hussain, Aftab M.; Sevilla Torres, Galo A.; Banerjee, Sanjay K.; Hussain, Muhammad M.

    2014-06-01

    Evolution in transistor technology from increasingly large power consuming single gate planar devices to energy efficient multiple gate non-planar ultra-narrow (< 20 nm) fins has enhanced the scaling trend to facilitate doubling performance. However, this performance gain happens at the expense of arraying multiple devices (fins) per operation bit, due to their ultra-narrow dimensions (width) originated limited number of charges to induce appreciable amount of drive current. Additionally arraying degrades device off-state leakage and increases short channel characteristics, resulting in reduced chip level energy-efficiency. In this paper, a novel nanotube device (NTFET) topology based on conventional group IV (Si, SiGe) channel materials is discussed. This device utilizes a core/shell dual gate strategy to capitalize on the volume-inversion properties of an ultra-thin (< 10 nm) group IV nanotube channel to minimize leakage and short channel effects while maximizing performance in an area-efficient manner. It is also shown that the NTFET is capable of providing a higher output drive performance per unit chip area than an array of gate-all-around nanowires, while maintaining the leakage and short channel characteristics similar to that of a single gate-all-around nanowire, the latter being the most superior in terms of electrostatic gate control. In the age of big data and the multitude of devices contributing to the internet of things, the NTFET offers a new transistor topology alternative with maximum benefits from performance-energy efficiency-functionality perspective.

  7. From NDE to Prognostics: A Revolution in Asset Management for Generation IV Nuclear Power Plants

    SciTech Connect

    Bond, Leonard J.; Doctor, Steven R.

    2007-06-01

    For Generation IV nuclear power plants (NPP) to achieve operational goals it is necessary to adopt new on-line monitoring and prognostic methodologies, giving operators better plant situational awareness and reliable predictions of remaining service life. Such techniques can improve plant economics, reduce unplanned outages, improve safety and provide probabilistic risk assessments. This paper reviews the state of the art and the potential impact from monitoring, diagnostics and prognostics on advanced NPP, with a focus on the needs of Generation IV systems.

  8. Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

    DOE PAGES

    Cheng, Lap-Yan; Wei, Thomas Y. C.

    2009-01-01

    The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct-cycle gas-cooled fast reactor (GFR) which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow weremore » evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.« less

  9. Consistent Multigroup Theory Enabling Accurate Course-Group Simulation of Gen IV Reactors

    SciTech Connect

    Rahnema, Farzad; Haghighat, Alireza; Ougouag, Abderrafi

    2013-11-29

    The objective of this proposal is the development of a consistent multi-group theory that accurately accounts for the energy-angle coupling associated with collapsed-group cross sections. This will allow for coarse-group transport and diffusion theory calculations that exhibit continuous energy accuracy and implicitly treat cross- section resonances. This is of particular importance when considering the highly heterogeneous and optically thin reactor designs within the Next Generation Nuclear Plant (NGNP) framework. In such reactors, ignoring the influence of anisotropy in the angular flux on the collapsed cross section, especially at the interface between core and reflector near which control rods are located, results in inaccurate estimates of the rod worth, a serious safety concern. The scope of this project will include the development and verification of a new multi-group theory enabling high-fidelity transport and diffusion calculations in coarse groups, as well as a methodology for the implementation of this method in existing codes. This will allow for a higher accuracy solution of reactor problems while using fewer groups and will reduce the computational expense. The proposed research represents a fundamental advancement in the understanding and improvement of multi- group theory for reactor analysis.

  10. A Technology Roadmap for Generation IV Nuclear Energy Systems Executive Summary

    SciTech Connect

    2003-03-01

    To meet future energy needs, ten countries--Argentina, Brazil, Canada, France, Japan, the Republic of Korea, the Republic of South Africa, Switzerland, the United Kingdom, and the United States--have agreed on a framework for international cooperation in research for an advanced generation of nuclear energy systems, known as Generation IV. These ten countries have joined together to form the Generation IV International Forum (GIF) to develop future-generation nuclear energy systems that can be licensed, constructed, and operated in a manner that will provide competitively priced and reliable energy products while satisfactorily addressing nuclear safety, waste, proliferation, and public perception concerns. The objective for Generation IV nuclear energy systems is to be available for international deployment before the year 2030, when many of the world's currently operating nuclear power plants will be at or near the end of their operating licenses.

  11. Steam generator for liquid metal fast breeder reactor

    DOEpatents

    Gillett, James E.; Garner, Daniel C.; Wineman, Arthur L.; Robey, Robert M.

    1985-01-01

    Improvements in the design of internal components of J-shaped steam generators for liquid metal fast breeder reactors. Complex design improvements have been made to the internals of J-shaped steam generators which improvements are intended to reduce tube vibration, tube jamming, flow problems in the upper portion of the steam generator, manufacturing complexities in tube spacer attachments, thermal stripping potentials and difficulties in the weld fabrication of certain components.

  12. Dry phase reactor for generating medical isotopes

    DOEpatents

    Mackie, Thomas Rockwell; Heltemes, Thad Alexander

    2016-05-03

    An apparatus for generating medical isotopes provides for the irradiation of dry-phase, granular uranium compounds which are then dissolved in a solvent for separation of the medical isotope from the irradiated compound. Once the medical isotope is removed, the dissolved compound may be reconstituted in dry granular form for repeated irradiation.

  13. An EBSD investigation on flow localization and microstructure evolution of 316L stainless steel for Gen IV reactor applications

    NASA Astrophysics Data System (ADS)

    Wu, Xianglin; Pan, Xiao; Mabon, James C.; Li, Meimei; Stubbins, James F.

    2007-09-01

    Type 316L stainless steel has been selected as a candidate structural material in a series of current accelerator driven systems and Generation IV reactor conceptual designs. The material is sensitive to irradiation damage in the temperature range of 150-400 °C: even low levels of irradiation exposure, as small as 0.1 dpa, can cause severe loss of ductility during tensile loading. This process, where the plastic flow becomes highly localized resulting in extremely low overall ductility, is referred as flow localization. The process controlling this confined flow is related to the difference between the yield and ultimate tensile strengths such that large irradiation-induced increases in the yield strength result in very limited plastic flow leading to necking after very small levels of uniform elongation. In this study, the microstructural evolution controlling flow localization is examined. It is found that twinning is an important deformation mechanism at lower temperatures since it promotes the strain hardening process. At higher temperatures, twinning becomes energetically impossible since the activation of twinning is determined by the critical twinning stress, which increases rapidly with temperature. Mechanical twinning and dislocation-based planar slip are competing mechanisms for plastic deformation.

  14. Non-destructive research methods applied on materials for the new generation of nuclear reactors

    NASA Astrophysics Data System (ADS)

    Bartošová, I.; Slugeň, V.; Veterníková, J.; Sojak, S.; Petriska, M.; Bouhaddane, A.

    2014-06-01

    The paper is aimed on non-destructive experimental techniques applied on materials for the new generation of nuclear reactors (GEN IV). With the development of these reactors, also materials have to be developed in order to guarantee high standard properties needed for construction. These properties are high temperature resistance, radiation resistance and resistance to other negative effects. Nevertheless the changes in their mechanical properties should be only minimal. Materials, that fulfil these requirements, are analysed in this work. The ferritic-martensitic (FM) steels and ODS steels are studied in details. Microstructural defects, which can occur in structural materials and can be also accumulated during irradiation due to neutron flux or alpha, beta and gamma radiation, were analysed using different spectroscopic methods as positron annihilation spectroscopy and Barkhausen noise, which were applied for measurements of three different FM steels (T91, P91 and E97) as well as one ODS steel (ODS Eurofer).

  15. Generation III reactors safety requirements and the design solutions

    NASA Astrophysics Data System (ADS)

    Felten, P.

    2009-03-01

    Nuclear energy's public acceptance, and hence its development, depends on its safety. As a reactor designer, we will first briefly remind the basic safety principles of nuclear reactors' design. We will then show how the industry, and in particular Areva with its EPR, made design evolution in the wake of the Three Miles Island accident in 1979. In particular, for this new generation of reactors, severe accidents are taken into account beyond the standard design basis accidents. Today, Areva's EPR meets all so-called "generation III" safety requirements and was licensed by several nuclear safety authorities in the world. Many innovative solutions are integrated in the EPR, some of which will be introduced here.

  16. Design of Radiation-Tolerant Structural Alloys for Generation IV Nuclear Energy Systems

    SciTech Connect

    Allen, T.R.; Was, G.S.; Bruemmer, S.M.; Gan, J.; Ukai, S.

    2005-12-28

    The objective of this program is to improve the radiation tolerance of both austenitic and ferritic-martensitic (F-M) alloys projected for use in Generation IV systems. The expected materials limitations of Generation IV components include: creep strength, dimensional stability, and corrosion/stress corrosion compatibility. The material design strategies to be tested fall into three main categories: (1) engineering grain boundaries; (2) alloying, by adding oversized elements to the matrix; and (3) microstructural/nanostructural design, such as adding matrix precipitates. These three design strategies were tested across both austenitic and ferritic-martensitic alloy classes

  17. Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency

    SciTech Connect

    R. Wigeland; K. Hamman

    2009-09-01

    Suggested for Track 7: Advances in Reactor Core Design and In-Core Management _____________________________________________________________________________________ Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency R. Wigeland and K. Hamman Idaho National Laboratory Given the ability of fast reactors to effectively transmute the transuranic elements as are present in spent nuclear fuel, fast reactors are being considered as one element of future nuclear power systems to enable continued use and growth of nuclear power by limiting high-level waste generation. However, a key issue for fast reactors is higher electricity cost relative to other forms of nuclear energy generation. The economics of the fast reactor are affected by the amount of electric power that can be produced from a reactor, i.e., the thermal efficiency for electricity generation. The present study is examining the potential for fast reactor subassembly design changes to improve the thermal efficiency by increasing the average coolant outlet temperature without increasing peak temperatures within the subassembly, i.e., to make better use of current technology. Sodium-cooled fast reactors operate at temperatures far below the coolant boiling point, so that the maximum coolant outlet temperature is limited by the acceptable peak temperatures for the reactor fuel and cladding. Fast reactor fuel subassemblies have historically been constructed using a large number of small diameter fuel pins contained within a tube of hexagonal cross-section, or hexcan. Due to this design, there is a larger coolant flow area next to the hexcan wall as compared to flow area in the interior of the subassembly. This results in a higher flow rate near the hexcan wall, overcooling the fuel pins next to the wall, and a non-uniform coolant temperature distribution. It has been recognized for many years that this difference in sodium coolant temperature was detrimental to achieving

  18. Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

    SciTech Connect

    Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

    2007-03-21

    In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for

  19. Mechanisms Governing the Creep Behavior of High Temperature Alloys for Generation IV Nuclear Energy Systems

    SciTech Connect

    Vasudevan, Vijay; Carroll, Laura; Sham, Sam

    2015-04-06

    This research project, which includes collaborators from INL and ORNL, focuses on the study of alloy 617 and alloy 800H that are candidates for applications as intermediate heat exchangers in GEN IV nuclear reactors, with an emphasis on the effects of grain size, grain boundaries and second phases on the creep properties; the mechanisms of dislocation creep, diffusional creep and cavitation; the onset of tertiary creep; and theoretical modeling for long-term predictions of materials behavior and for high temperature alloy design.

  20. Managing Model Data Introduced Uncertainties in Simulator Predictions for Generation IV Systems via Optimum Experimental Design

    SciTech Connect

    Turinsky, Paul J; Abdel-Khalik, Hany S; Stover, Tracy E

    2011-03-31

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

  1. Generation IV Nuclear Energy Systems Ten-Year Program Plan Fiscal Year 2005, Volume 1

    SciTech Connect

    2005-03-01

    As reflected in the U.S. ''National Energy Policy'', nuclear energy has a strong role to play in satisfying our nation's future energy security and environmental quality needs. The desirable environmental, economic, and sustainability attributes of nuclear energy give it a cornerstone position, not only in the U.S. energy portfolio, but also in the world's future energy portfolio. Accordingly, on September 20, 2002, U.S. Energy Secretary Spencer Abraham announced that, ''The United States and nine other countries have agreed to develop six Generation IV nuclear energy concepts''. The Secretary also noted that the systems are expected to ''represent significant advances in economics, safety, reliability, proliferation resistance, and waste minimization''. The six systems and their broad, worldwide research and development (R&D) needs are described in ''A Technology Roadmap for Generation IV Nuclear Energy Systems'' (hereafter referred to as the Generation IV Roadmap). The first 10 years of required U.S. R&D contributions to achieve the goals described in the Generation IV Roadmap are outlined in this Program Plan.

  2. The Generation in Between: A Perspective from the Keystone IV Conference.

    PubMed

    Chen, Frederick M; Bliss, Erika; Dunn, Aaron; Edgoose, Jennifer; Elliott, Tricia C; Maxwell, Lisa C; Morris, Carl G; Phillips, Robert L

    2016-01-01

    Keystone IV affirmed the value of relationships in family medicine, but each generation of family physicians took away different impressions and lessons. "Generation III," between the Baby Boomers and Millennials, reported conflict between their professional ideal of family medicine and the realities of current practice. But the Keystone conference also helped them appreciate core values of family medicine, their shared experience, and new opportunities for leadership. PMID:27387165

  3. Physics of reactor safety. Quarterly report, October-December 1980. Volume IV

    SciTech Connect

    Not Available

    1981-02-01

    The work in the Applied Physics Division includes reports on reactor safety modeling and assessment by members of the Reactor Safety Appraisals Section. Work on reactor core thermal-hydraulics is performed in ANL's Components Technology Division, emphasizing 3-dimensional code development for LMFBR accidents under natural convection conditions.

  4. Technological Transfer from Research Nuclear Reactors to New Generation Nuclear Power Reactors

    NASA Astrophysics Data System (ADS)

    Radulescu, Laura; Pavelescu, Margarit

    2010-01-01

    The goal of this paper is the analysis of the technological transfer role in the nuclear field, with particular emphasis on nuclear reactors domain. The presentation is sustained by historical arguments. In this frame, it is very important to start with the achievements of the first nuclear systems, for instant those with natural uranium as fuel and heavy water as moderator, following in time through the history until the New Generation Nuclear Power Reactors. Starting with 1940, the accelerated development of the industry has implied the increase of the global demand for energy. In this respect, the nuclear energy could play an important role, being essentially an unlimited source of energy. However, the nuclear option faces the challenges of increasingly demanding safety requirements, economic competitiveness and public acceptance. Worldwide, a significant amount of experience has been accumulated during development, licensing, construction, and operation of nuclear power reactors. The experience gained is a strong basis for further improvements. Actually, the nuclear programs of many countries are addressing the development of advanced reactors, which are intended to have better economics, higher reliability, improved safety, and proliferation-resistant characteristics in order to overcome the current concerns about nuclear power. Advanced reactors, now under development, may help to meet the demand for energy power of both developed and developing countries as well as for district heating, desalination and for process heat. The paper gives historical examples that illustrate the steps pursued from first research nuclear reactors to present advanced power reactors. Emphasis was laid upon the fact that the progress is due to the great discoveries of the nuclear scientists using the technological transfer.

  5. Technological Transfer from Research Nuclear Reactors to New Generation Nuclear Power Reactors

    SciTech Connect

    Radulescu, Laura; Pavelescu, Margarit

    2010-01-21

    The goal of this paper is the analysis of the technological transfer role in the nuclear field, with particular emphasis on nuclear reactors domain. The presentation is sustained by historical arguments. In this frame, it is very important to start with the achievements of the first nuclear systems, for instant those with natural uranium as fuel and heavy water as moderator, following in time through the history until the New Generation Nuclear Power Reactors.Starting with 1940, the accelerated development of the industry has implied the increase of the global demand for energy. In this respect, the nuclear energy could play an important role, being essentially an unlimited source of energy. However, the nuclear option faces the challenges of increasingly demanding safety requirements, economic competitiveness and public acceptance. Worldwide, a significant amount of experience has been accumulated during development, licensing, construction, and operation of nuclear power reactors. The experience gained is a strong basis for further improvements. Actually, the nuclear programs of many countries are addressing the development of advanced reactors, which are intended to have better economics, higher reliability, improved safety, and proliferation-resistant characteristics in order to overcome the current concerns about nuclear power. Advanced reactors, now under development, may help to meet the demand for energy power of both developed and developing countries as well as for district heating, desalination and for process heat.The paper gives historical examples that illustrate the steps pursued from first research nuclear reactors to present advanced power reactors. Emphasis was laid upon the fact that the progress is due to the great discoveries of the nuclear scientists using the technological transfer.

  6. Creep Behavior of High Temperature Alloys for Generation IV Nuclear Energy Systems

    NASA Astrophysics Data System (ADS)

    Wen, Xingshuo

    The Very High Temperature Reactor (VHTR) is one of the leading concepts of the Generation IV nuclear reactor development, which is the core component of Next Generation Nuclear Plant (NGNP). The major challenge in the research and development of NGNP is the performance and reliability of structure materials at high temperature. Alloy 617, with an exceptional combination of high temperature strength and oxidation resistance, has been selected as a primary candidate material for structural use, particularly in Intermediate Heat Exchanger (IHX) which has an outlet temperature in the range of 850 to 950°C and an inner pressure from 5 to 20MPa. In order to qualify the material to be used at the operation condition for a designed service life of 60 years, a comprehensive scientific understanding of creep behavior at high temperature and low stress regime is necessary. In addition, the creep mechanism and the impact factors such as precipitates, grain size, and grain boundary characters need to be evaluated for the purpose of alloy design and development. In this study, thermomechanically processed specimens of alloy 617 with different grain sizes were fabricated, and creep tests with a systematic test matrix covering the temperatures of 850 to 1050°C and stress levels from 5 to 100MPa were conducted. Creep data was analyzed, and the creep curves were found to be unconventional without a well-defined steady-state creep. Very good linear relationships were determined for minimum creep rate versus stress levels with the stress exponents determined around 3-5 depending on the grain size and test condition. Activation energies were also calculated for different stress levels, and the values are close to 400kJ/mol, which is higher than that for self-diffusion in nickel. Power law dislocation climb-glide mechanism was proposed as the dominant creep mechanism in the test condition regime. Dynamic recrystallization happening at high strain range enhanced dislocation climb and

  7. Steam generator tubing development for commercial fast breeder reactors

    SciTech Connect

    Sessions, C.E.; Uber, C.F.

    1981-11-01

    The development work to design, manufacture, and evaluate pre-stressed double-wall 2/one quarter/ Cr-1 Mo steel tubing for commercial fast breeder reactor steam generator application is discussed. The Westinghouse plan for qualifying tubing vendors to produce this tubing is described. The results achieved to date show that a long length pre-stressed double-wall tube is both feasible and commercially available. The evaluation included structural analysis and experimental measurement of the pre-stress within tubes, as well as dimensional, metallurgical, and interface wear tests of tube samples produced. This work is summarized and found to meet the steam generator design requirements. 10 refs.

  8. Documentation for MeshKit - Reactor Geometry (&mesh) Generator

    SciTech Connect

    Jain, Rajeev; Mahadevan, Vijay

    2015-09-30

    This report gives documentation for using MeshKit’s Reactor Geometry (and mesh) Generator (RGG) GUI and also briefly documents other algorithms and tools available in MeshKit. RGG is a program designed to aid in modeling and meshing of complex/large hexagonal and rectilinear reactor cores. RGG uses Argonne’s SIGMA interfaces, Qt and VTK to produce an intuitive user interface. By integrating a 3D view of the reactor with the meshing tools and combining them into one user interface, RGG streamlines the task of preparing a simulation mesh and enables real-time feedback that reduces accidental scripting mistakes that could waste hours of meshing. RGG interfaces with MeshKit tools to consolidate the meshing process, meaning that going from model to mesh is as easy as a button click. This report is designed to explain RGG v 2.0 interface and provide users with the knowledge and skills to pilot RGG successfully. Brief documentation of MeshKit source code, tools and other algorithms available are also presented for developers to extend and add new algorithms to MeshKit. RGG tools work in serial and parallel and have been used to model complex reactor core models consisting of conical pins, load pads, several thousands of axially varying material properties of instrumentation pins and other interstices meshes.

  9. IGORR-IV -- Proceedings of the fourth meeting of the International Group on Research Reactors

    SciTech Connect

    Rosenbalm, K.F.

    1995-12-31

    The International Group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Twenty-nine papers were presented in five sessions and written versions of the papers or hard copies of the vugraphs used are published in these proceedings. The five sessions were: (1) Operating Research Reactors and Facility Upgrades; (2) Research Reactors in Design and Construction; (3) ANS Closeout Activities; (4) and (5) Research, Development, and Analysis Results.

  10. Generation IV Nuclear Energy Systems Construction Cost Reductions through the Use of Virtual Environments - Final Report

    SciTech Connect

    Timothy Shaw; Anthony Baratta; Vaughn Whisker

    2005-02-28

    Final report of 3 year DOE NERI-sponsored effort evaluating immersive virtual reality (CAVE) technology for design review, construction planning, and maintenance planning and training for next generation nuclear power plants. Program covers development of full-scale virtual mockups generated from 3D CAD data presented in a CAVE visualization facility. Mockups applied to design review of AP600/1000, Construction planning for AP 600, and AP 1000 maintenance evaluation. Proof of concept study also performed for GenIV PBMR models.

  11. A nonheme manganese(IV)-oxo species generated in photocatalytic reaction using water as an oxygen source.

    PubMed

    Wu, Xiujuan; Yang, Xiaonan; Lee, Yong-Min; Nam, Wonwoo; Sun, Licheng

    2015-03-01

    A nonheme manganese(IV)-oxo complex, [Mn(IV)(O)(BQCN)](2+), was generated in the photochemical and chemical oxidation of [Mn(II)(BQCN)](2+) with water as an oxygen source, respectively. The photocatalytic oxidation of organic substrates, such as alcohol and sulfide, by [Mn(II)(BQCN)](2+) has been demonstrated in both neutral and acidic media. PMID:25658677

  12. "A New Class od Functionally Graded Cearamic-Metal Composites for Next Generation Very High Temperature Reactors"

    SciTech Connect

    Dr. Mohit Jain; Dr. Ganesh Skandan; Dr. Gordon E. Khose; Mrs. Judith Maro, Nuclear Reactor Laboratory, MIT

    2008-05-01

    Generation IV Very High Temperature power generating nuclear reactors will operate at temperatures greater than 900 oC. At these temperatures, the components operating in these reactors need to be fabricated from materials with excellent thermo-mechanical properties. Conventional pure or composite materials have fallen short in delivering the desired performance. New materials, or conventional materials with new microstructures, and associated processing technologies are needed to meet these materials challenges. Using the concept of functionally graded materials, we have fabricated a composite material which has taken advantages of the mechanical and thermal properties of ceramic and metals. Functionally-graded composite samples with various microstructures were fabricated. It was demonstrated that the composition and spatial variation in the composition of the composite can be controlled. Some of the samples were tested for irradiation resistance to neutrons. The samples did not degrade during initial neutron irradiation testing.

  13. High-temperature gas-cooled reactors: preliminary safety and environmental information document. Volume IV

    SciTech Connect

    Not Available

    1980-01-01

    Information is presented concerning medium-enriched uranium/thorium once-through fuel cycle; medium-enrichment uranium-233/thorium recycle fuel; high-enrichment uranium-235/thorium recycle (spiked) fuel cycle; high-enrichment uranium-233/thorium recycle (spiked) fuel cycle; and gas-turbine high-temperature gas-cooled reactor.

  14. Evaluation Methodology For Proliferation Resistance And Physical Protection Of Generation IV Nuclear Energy Systems: An Overview

    SciTech Connect

    T. Bjornard; R. Bari; R. Nishimura; P. Peterson; J. Roglans; D. Bley; J. Cazalet; G.G.M. Cojazzi; P. Delaune; M. Golay; G. Rendad; G. Rochau; M. Senzaki; I. Therios; M. Zentner

    2006-05-01

    This paper provides an overview of the methodology approach developed by the Generation IV International Forum Expert Group on Proliferation Resistance & Physical Protection for evaluation of Proliferation Resistance and Physical Protection robustness of Generation IV nuclear energy systems options. The methodology considers a set of alternative systems and evaluates their resistance or robustness to a collection of potential threats. For the challenges considered, the response of the system to these challenges is assessed and expressed in terms of outcomes. The challenges to the system are given by the threats posed by potential proliferant States and sub-national adversaries on the nuclear systems. The characteristics of the Generation IV systems, both technical and institutional, are used to evaluate their response to the threats and determine their resistance against the proliferation threats and robustness against sabotage and theft threats. System response encompasses three main elements: 1.System Element Identification. The nuclear energy system is decomposed into smaller elements (subsystems) at a level amenable to further analysis. 2.Target Identification and Categorization. A systematic process is used to identify and select representative targets for different categories of pathways, within each system element, that actors (proliferant States or adversaries) might choose to use or attack. 3.Pathway Identification and Refinement. Pathways are defined as potential sequences of events and actions followed by the proliferant State or adversary to achieve its objectives (proliferation, theft or sabotage). For each target, individual pathway segments are developed through a systematic process, analyzed at a high level, and screened where possible. Segments are connected into full pathways and analyzed in detail. The outcomes of the system response are expressed in terms of PR&PP measures. Measures are high-level characteristics of a pathway that include

  15. EVALUATION METHODOLOGY FOR PROLIFERATION RESISTANCE AND PHYSICAL PROTECTION OF GENERATION IV NUCLEAR ENERGY SYSTEMS: AN OVERVIEW.

    SciTech Connect

    BARI, R.; ET AL.

    2006-03-01

    This paper provides an overview of the methodology approach developed by the Generation IV International Forum Expert Group on Proliferation Resistance & Physical Protection for evaluation of Proliferation Resistance and Physical Protection robustness of Generation IV nuclear energy systems options. The methodology considers a set of alternative systems and evaluates their resistance or robustness to a collection of potential threats. For the challenges considered, the response of the system to these challenges is assessed and expressed in terms of outcomes. The challenges to the system are given by the threats posed by potential proliferant States and sub-national adversaries on the nuclear systems. The characteristics of the Generation IV systems, both technical and institutional, are used to evaluate their response to the threats and determine their resistance against the proliferation threats and robustness against sabotage and theft threats. System response encompasses three main elements: (1) System Element Identification. The nuclear energy system is decomposed into smaller elements (subsystems) at a level amenable to further analysis. (2) Target Identification and Categorization. A systematic process is used to identify and select representative targets for different categories of pathways, within each system element, that actors (proliferant States or adversaries) might choose to use or attack. (3) Pathway Identification and Refinement. Pathways are defined as potential sequences of events and actions followed by the proliferant State or adversary to achieve its objectives (proliferation, theft or sabotage). For each target, individual pathway segments are developed through a systematic process, analyzed at a high level, and screened where possible. Segments are connected into full pathways and analyzed in detail. The outcomes of the system response are expressed in terms of PR&PP measures. Measures are high-level characteristics of a pathway that include

  16. Design reliability assurance program for Korean next generation reactor

    SciTech Connect

    Lee, Beom-Su; Han, Jin-Kyu; Na, Jang Hwan; Yoo, Kyung Yeong

    1997-12-01

    The Korean Next Generation Reactor (KNGR) project is to develop standardized nuclear power plant design for the construction of future nuclear power plants in Korea. The main purpose of the KNGR project is to develop the advanced nuclear power plants, which enhance safety and economics significantly through the incorporation of design concepts for severe accident prevention and mitigation, supplementary passive safety concept, simplification and application of modularization and so on. For those, Probabilistic Safety Assessment (PSA) and availability study will be performed at the early stage of the design, and the Design Reliability Assurance Program (D-RAP) is applied in the development of the KNGR to ensure that the safety and availability evaluated in the PSA and availability study at the early phase of the design is maintained through the detailed design, construction, procurement and operation of the plants. This paper presents the D-RAP concept that could be applied at the stage of the basic design of the nuclear power plants, based on the models for the reference plants and/or similar plants. 4 refs., 1 fig.

  17. Calculation of ozone generation by positive dc corona discharge in coaxial wire-cylinder reactors

    SciTech Connect

    Yehia, Ashraf

    2007-01-15

    A theoretical equation has been derived in this paper for calculation of ozone generation by positive dc corona discharge in coaxial wire-cylinder reactors. The derived equation has been based on the theories of the positive dc coronas reported in the literature and extended to account the ozone destruction within the corona discharge plasma generated in the reactor. The equation has been investigated with experimental results for ozone generated in a coaxial wire-cylinder reactor under different discharge conditions. The reactor was stressed with a positive dc voltage and fed by dry air flowing with constant rates at atmospheric pressure and room temperature. The theoretical results calculated by the derived equation have shown good agreement with the experimental results over the whole range of the investigated parameters. Subsequently, the derived equation is valid to predict the ozone concentration generated in the investigated reactor under any discharge conditions.

  18. The utility industry response to Title IV: generation mix, fuel choice, emissions and costs.

    PubMed

    Molburg, J C

    1993-02-01

    The Clean Air Act Amendments of 1990 incorporate, for the first time, provisions aimed specifically at the control of acid rain. These provisions restrict emissions of sulfur dioxide and oxides of nitrogen from electric power generating stations. The restrictions on sulfur dioxide take the form of an overall cap on the aggregate emissions from major generating plants, allowing substantial flexibility in the industry's response to those restrictions. This report describes one response scenario through the year 2030, which was examined by simulation of the utility industry under assumptions consistent with a reference case that was used for analysis of the National Energy Strategy. Emissions that would result from the use of existing and new capacity and the associated additional costs of meeting demand subject to the emission limitations imposed by the Clean Air Act are projected. Fuel use effects, including coal market shifts, consistent with the response scenario are also described. These results, while dependent on specific assumptions for this scenario, provide insight into the general character of the likely utility industry response to Title IV.

  19. Raytheon explores thorium for next generation nuclear reactor

    SciTech Connect

    Crawford, M.

    1994-03-08

    Few new orders for nuclear power plants have been placed anywhere in the world in the last 20 years, but that is not discouraging Raytheon Engineers Constructors from making plans to explore new light water reactor technologies for commercial markets. The Lexington, Mass.-based company, which has extensive experience in nuclear power engineering and construction, has a vision for the light water reactor of the future - one that is based on the use of thorium-232, an element that decays over several steps to uranium-233. The use of thorium and a small amount of uranium that is 20 percent enriched is seen as providing operational, environmental, and safety advantages over reactors using the standard fuel mixture of uranium-238 and enriched uranium-235. According to Raytheon, the system could improve the economics of some reactors' operations by reducing fuel costs and lowering related waste volumes. At the same time, reactor safety could be improved by simpler control rod systems and the absence from reactor coolant of corrosive boric acid, which is used to slow neutrons in order to enhance reactions. Using thorium is also attractive because more of the fuel is burned up by the reactor, an estimated 12 percent as compared to about 4 percent for U-235. However, the technology's greatest attraction may well be its implications for nuclear proliferation. Growing plutonium inventories embedded in spent fuel rods from light water reactors have sparked concern worldwide. But according to Raytheon, using a thorium-based fuel core would alleviate this concern because it would produce only small quantities of plutonium. A thorium-based fuel system would produce 12 kilograms of plutonium over a decade versus 2,235 kilograms for an equivalent reactor operating with conventional uranium fuel.

  20. Generation of dipeptidyl peptidase-IV-inhibiting peptides from β-lactoglobulin secreted by Lactococcus lactis.

    PubMed

    Shigemori, Suguru; Oshiro, Kazushi; Wang, Pengfei; Yamamoto, Yoshinari; Wang, Yeqin; Sato, Takashi; Uyeno, Yutaka; Shimosato, Takeshi

    2014-01-01

    Previous studies showed that hydrolysates of β-lactoglobulin (BLG) prepared using gastrointestinal proteases strongly inhibit dipeptidyl peptidase-IV (DPP-IV) activity in vitro. In this study, we developed a BLG-secreting Lactococcus lactis strain as a delivery vehicle and in situ expression system. Interestingly, trypsin-digested recombinant BLG from L. lactis inhibited DPP-IV activity, suggesting that BLG-secreting L. lactis may be useful in the treatment of type 2 diabetes mellitus. PMID:25157356

  1. Dipeptidyl peptidase IV inhibitory peptides generated in Spanish dry-cured ham.

    PubMed

    Gallego, Marta; Aristoy, María-Concepción; Toldrá, Fidel

    2014-02-01

    Dipeptidyl peptidase IV (DPP-IV) inhibitors are promising new therapies for type 2 diabetes. The aim of this study was to assay DPP-IV inhibitory peptides that can be present in a water soluble extract of Spanish dry-cured ham. Such an extract was fractionated by size-exclusion chromatography and the in vitro DPP-IV inhibitory activity determined in each collected fraction. Then, several peptides previously identified in dry-cured ham extracts or known to be products of DPP IV action were synthesised and assayed for DPP-IV inhibition. Peptides KA and AAATP showed the strongest DPP-IV inhibitory activity, with IC50 values of 6.27 mM and 6.47 mM, respectively. Dipeptides AA, GP, PL, and carnosine, as well as peptides AAAAG, ALGGA, and LVSGM were also DPP-IV inhibitors, although at a lower degree. These findings suggest the potential of Spanish dry-cured ham as a natural precursor of DPP-IV inhibitory peptides. These biopeptides could also be used as ingredients for functional foods or pharmaceutical products against type 2 diabetes. PMID:24200567

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

    SciTech Connect

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

    2005-10-01

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

  3. What went Right: Resilience of Existing Reactors to - for Generation III+ Reactor Design

    NASA Astrophysics Data System (ADS)

    Garwin, Richard L.

    2014-07-01

    To quote Tolstoy's Anna Karenina, "All happy families are alike; each unhappy family is unhappy in its own way." So the reactors that have been working well in the world don't get a lot of attention...

  4. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    DOEpatents

    McDermott, Daniel J.; Schrader, Kenneth J.; Schulz, Terry L.

    1994-01-01

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  5. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    DOEpatents

    McDermott, D.J.; Schrader, K.J.; Schulz, T.L.

    1994-05-03

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  6. An experimental study of a VVER reactor's steam generator model operating in the condensing mode

    NASA Astrophysics Data System (ADS)

    Morozov, A. V.; Remizov, O. V.

    2012-05-01

    Results obtained from an experimental study of a VVER reactor's steam generator model operating in the condensing mode are presented. The obtained empirical dependence for calculating the power of heat exchangers operating in the steam condensation mode is presented.

  7. DOS-HEATING6: A general conduction code with nuclear heat generation derived from DOT-IV transport calculations

    SciTech Connect

    Williams, M.L.; Yuecel, A.; Nadkarny, S.

    1988-05-01

    The HEATING6 heat conduction code is modified to (a) read the multigroup particle fluxes from a two-dimensional DOT-IV neutron- photon transport calculation, (b) interpolate the fluxes from the DOT-IV variable (optional) mesh to the HEATING6 control volume mesh, and (c) fold the interpolated fluxes with kerma factors to obtain a nuclear heating source for the heat conduction equation. The modified HEATING6 is placed as a module in the ORNL discrete ordinates system (DOS), and has been renamed DOS-HEATING6. DOS-HEATING6 provides the capability for determining temperature distributions due to nuclear heating in complex, multi-dimensional systems. All of the original capabilities of HEATING6 are retained for the nuclear heating calculation; e.g., generalized boundary conditions (convective, radiative, finned, fixed temperature or heat flux), temperature and space dependent thermal properties, steady-state or transient analysis, general geometry description, etc. The numerical techniques used in the code are reviewed and the user input instructions and JCL to perform DOS-HEATING6 calculations are presented. Finally a sample problem involving coupled DOT-IV and DOS-HEATING6 calculations of a complex space-reactor configurations described, and the input and output of the calculations are listed. 10 refs., 11 figs., 6 tabs.

  8. UF6 breeder reactor power plants for electric power generation

    NASA Technical Reports Server (NTRS)

    Rust, J. H.; Clement, J. D.; Hohl, F.

    1976-01-01

    The reactor concept analyzed is a U-233F6 core surrounded by a molten salt (Li(7)F, BeF2, ThF4) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. Advantages of the gas core breeder reactor (GCBR) are as follows: (1) high efficiency; (2) simplified on-line reprocessing; (3) inherent safety considerations; (4) high breeding ratio; (5) possibility of burning all or most of the long-lived nuclear waste actinides; and (6) possibility of extrapolating the technology to higher temperatures and MHD direct conversion.

  9. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    SciTech Connect

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

    The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

  10. Roadmap to NRC Approval of Ceramic Matrix Composites in Generation IV Reactors

    SciTech Connect

    M. G. Jenkins; E. Lara-Curzio; W. Windes

    2006-05-01

    This report provides an initial roadmap to obtain Nuclear Regulatory Commission (NRC) approval for using these material systems in a nuclear application. The possible paths taken to achieving NRC approval are necessarily subject to change as this is an on-going process that shifts as more data and a clearer understanding of the nuclear regulations are gathered.

  11. Investigation of a Novel NDE Method for Monitoring Thermomechanical Damage and Microstructure Evolution in Ferritic-Martensitic Steels for Generation IV Nuclear Energy Systems

    SciTech Connect

    Nagy, Peter

    2013-09-30

    The main goal of the proposed project is the development of validated nondestructive evaluation (NDE) techniques for in situ monitoring of ferritic-martensitic steels like Grade 91 9Cr-1Mo, which are candidate materials for Generation IV nuclear energy structural components operating at temperatures up to ~650{degree}C and for steam-generator tubing for sodium-cooled fast reactors. Full assessment of thermomechanical damage requires a clear separation between thermally activated microstructural evolution and creep damage caused by simultaneous mechanical stress. Creep damage can be classified as "negligible" creep without significant plastic strain and "ordinary" creep of the primary, secondary, and tertiary kind that is accompanied by significant plastic deformation and/or cavity nucleation and growth. Under negligible creep conditions of interest in this project, minimal or no plastic strain occurs, and the accumulation of creep damage does not significantly reduce the fatigue life of a structural component so that low-temperature design rules, such as the ASME Section III, Subsection NB, can be applied with confidence. The proposed research project will utilize a multifaceted approach in which the feasibility of electrical conductivity and thermo-electric monitoring methods is researched and coupled with detailed post-thermal/creep exposure characterization of microstructural changes and damage processes using state-of-the-art electron microscopy techniques, with the aim of establishing the most effective nondestructive materials evaluation technique for particular degradation modes in high-temperature alloys that are candidates for use in the Next Generation Nuclear Plant (NGNP) as well as providing the necessary mechanism-based underpinnings for relating the two. Only techniques suitable for practical application in situ will be considered. As the project evolves and results accumulate, we will also study the use of this technique for monitoring other GEN IV

  12. Composite Materials under Extreme Radiation and Temperature Environments of the Next Generation Nuclear Reactors

    SciTech Connect

    Simos, N.

    2011-05-01

    In the nuclear energy renaissance, driven by fission reactor concepts utilizing very high temperatures and fast neutron spectra, materials with enhanced performance that exceeds are expected to play a central role. With the operating temperatures of the Generation III reactors bringing the classical reactor materials close to their performance limits there is an urgent need to develop and qualify new alloys and composites. Efforts have been focused on the intricate relations and the high demands placed on materials at the anticipated extreme states within the next generation fusion and fission reactors which combine high radiation fluxes, elevated temperatures and aggressive environments. While nuclear reactors have been in operation for several decades, the structural materials associated with the next generation options need to endure much higher temperatures (1200 C), higher neutron doses (tens of displacements per atom, dpa), and extremely corrosive environments, which are beyond the experience on materials accumulated to-date. The most important consideration is the performance and reliability of structural materials for both in-core and out-of-core functions. While there exists a great body of nuclear materials research and operating experience/performance from fission reactors where epithermal and thermal neutrons interact with materials and alter their physio-mechanical properties, a process that is well understood by now, there are no operating or even experimental facilities that will facilitate the extreme conditions of flux and temperature anticipated and thus provide insights into the behaviour of these well understood materials. Materials, however, still need to be developed and their interaction and damage potential or lifetime to be quantified for the next generation nuclear energy. Based on material development advances, composites, and in particular ceramic composites, seem to inherently possess properties suitable for key functions within the

  13. Power generation from nuclear reactors in aerospace applications

    SciTech Connect

    English, R.E.

    1982-01-01

    Power generation in nuclear powerplants in space is addressed. In particular, the states of technology of the principal competitive concepts for power generation are assessed. The possible impact of power conditioning on power generation is also discussed. For aircraft nuclear propulsion, the suitability of various technologies is cursorily assessed for flight in the Earth's atmosphere. A program path is suggested to ease the conditions of first use of aircraft nuclear propulsion.

  14. Power Generation from Nuclear Reactors in Aerospace Applications

    NASA Technical Reports Server (NTRS)

    English, Robert E.

    1982-01-01

    Power generation in nuclear powerplants in space is addressed. In particular, the states of technology of the principal competitive concepts for power generation are assessed. The possible impact of power conditioning on power generation is also discussed. For aircraft nuclear propulsion, the suitability of various technologies is cursorily assessed for flight in the Earth's atmosphere; a program path is suggested to ease the conditions of first use of aircraft nuclear propulsion.

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

    SciTech Connect

    S. Blaine Grover; Raymond V. Furstenau

    2005-10-01

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

  16. How an 8 path chordal ultrasonic flowmeter solves steam generator and reactor water level control problems

    SciTech Connect

    Estrada, H.; Hauser, E.

    2006-07-01

    The application of a chordal ultrasonic transit time flowmeter to reactor and steam generator water level control is described. Analyses show that the extremely high turndown of the ultrasonic instrument, when coupled with a linear steam demand signal, leads to water level control system performance far superior to that achieved with conventional flow instruments, particularly at very low power. The proposed control system is readily justified by reduced startup times, avoidance of reactor trips (scrams), and reduced feedwater pumping power requirements. (authors)

  17. Generation of a novel mouse model that recapitulates early and adult onset glycogenosis type IV.

    PubMed

    Akman, H Orhan; Sheiko, Tatiana; Tay, Stacey K H; Finegold, Milton J; Dimauro, Salvatore; Craigen, William J

    2011-11-15

    Glycogen storage disease type IV (GSD IV) is a rare autosomal recessive disorder caused by deficiency of the glycogen branching enzyme (GBE). The diagnostic feature of the disease is the accumulation of a poorly branched form of glycogen known as polyglucosan (PG). The disease is clinically heterogeneous, with variable tissue involvement and age of disease onset. Absence of enzyme activity is lethal in utero or in infancy affecting primarily muscle and liver. However, residual enzyme activity (5-20%) leads to juvenile or adult onset of a disorder that primarily affects muscle as well as central and peripheral nervous system. Here, we describe two mouse models of GSD IV that reflect this spectrum of disease. Homologous recombination was used to insert flippase recognition target recombination sites around exon 7 of the Gbe1 gene and a phosphoglycerate kinase-Neomycin cassette within intron 7, leading to a reduced synthesis of GBE. Mice bearing this mutation (Gbe1(neo/neo)) exhibit a phenotype similar to juvenile onset GSD IV, with wide spread accumulation of PG. Meanwhile, FLPe-mediated homozygous deletion of exon 7 completely eliminated GBE activity (Gbe1(-/-)), leading to a phenotype of lethal early onset GSD IV, with significant in utero accumulation of PG. Adult mice with residual GBE exhibit progressive neuromuscular dysfunction and die prematurely. Differently from muscle, PG in liver is a degradable source of glucose and readily depleted by fasting, emphasizing that there are structural and regulatory differences in glycogen metabolism among tissues. Both mouse models recapitulate typical histological and physiological features of two human variants of branching enzyme deficiency. PMID:21856731

  18. Foreign Trip Report MATGEN-IV Sep 24- Oct 26, 2007

    SciTech Connect

    de Caro, M S

    2007-10-30

    Gen-IV activities in France, Japan and US focus on the development of new structural materials for Gen-IV nuclear reactors. Oxide dispersion strengthened (ODS) F/M steels have raised considerable interest in nuclear applications. Promising collaborations can be established seeking fundamental knowledge of relevant Gen-IV ODS steel properties (see attached travel report on MATGEN- IV 'Materials for Generation IV Nuclear Reactors'). Major highlights refer to results on future Ferritic/Martensitic steel cladding candidates (relevant to Gen-IV materials properties for LFR Materials Program) and on thermodynamic and mechanic behavior of metallic FeCr binary alloys, base matrix for future candidate steels (for the LLNL-LDRD project on Critical Issues on Materials for Gen-IV Reactors).

  19. Multiscale Modeling of the Deformation of Advanced Ferritic Steels for Generation IV Nuclear Energy

    SciTech Connect

    Nasr M. Ghoniem; Nick Kioussis

    2009-04-18

    The objective of this project is to use the multi-scale modeling of materials (MMM) approach to develop an improved understanding of the effects of neutron irradiation on the mechanical properties of high-temperature structural materials that are being developed or proposed for Gen IV applications. In particular, the research focuses on advanced ferritic/ martensitic steels to enable operation up to 650-700°C, compared to the current 550°C limit on high-temperature steels.

  20. Considerations Associated with Reactor Technology Selection for the Next Generation Nuclear Plant Project

    SciTech Connect

    L.E. Demick

    2010-09-01

    At the inception of the Next Generation Nuclear Plant Project and during predecessor activities, alternative reactor technologies have been evaluated to determine the technology that best fulfills the functional and performance requirements of the targeted energy applications and market. Unlike the case of electric power generation where the reactor performance is primarily expressed in terms of economics, the targeted energy applications involve industrial applications that have specific needs in terms of acceptable heat transport fluids and the associated thermodynamic conditions. Hence, to be of interest to these industrial energy applications, the alternative reactor technologies are weighed in terms of the reactor coolant/heat transport fluid, achievable reactor outlet temperature, and practicality of operations to achieve the very high reliability demands associated with the petrochemical, petroleum, metals and related industries. These evaluations have concluded that the high temperature gas-cooled reactor (HTGR) can uniquely provide the required ranges of energy needs for these target applications, do so with promising economics, and can be commercialized with reasonable development risk in the time frames of current industry interest – i.e., within the next 10-15 years.

  1. Decomposition of chlorinated ethylenes and ethanes in an electron beam generated plasma reactor

    SciTech Connect

    Vitale, S.A.

    1996-02-01

    An electron beam generated plasma reactor (EBGPR) is used to determine the plasma chemistry kinetics, energetics and decomposition pathways of six chlorinated ethylenes and ethanes: 1,1,1-trichloroethane, 1,1-dichloroethane, ethyl chloride, trichloroethylene, 1,1-dichloroethylene, and vinyl chloride. A traditional chemical kinetic and chemical engineering analysis of the data from the EBGPR is performed, and the following hypothesis was verified: The specific energy required for chlorinated VOC decomposition in the electron beam generated plasma reactor is determined by the electron attachment coefficient of the VOC and the susceptibility of the molecule to radical attack. The technology was demonstrated at the Hanford Reservation to remove VOCs from soils.

  2. Acoustical gas core reactor with MHD power generation for burst power in a bimodal system

    NASA Astrophysics Data System (ADS)

    Dugan, E. T.; Jacobs, A. M.; Oliver, C. C.; Lear, W. E., Jr.

    Research is being conducted on gas core reactors for space nuclear power to establish the scientific feasibility and engineering validation of a reactor and energy conversion system that can significantly improve specific power, dynamic performance and system efficiency. Rapid achievement of burst mode (GWe) operation at core power densities of 1 kW/mL and reactor masses of a kg/MWt are research objectives; coupled with MHD conversion, system efficiencies of 40 percent for open cycle operation and heat rejection temperatures of 1500 K or higher for closed cycle operation are anticipated. The design of the gas core reactor/MHD generator configuration to directly produce pulsed electrical power, thereby alleviating external power conditioning requirements, is also a research objective.

  3. The Next Generation of Platinum Drugs: Targeted Pt(II) Agents, Nanoparticle Delivery, and Pt(IV) Prodrugs

    PubMed Central

    Johnstone, Timothy C.; Suntharalingam, Kogularamanan; Lippard, Stephen J.

    2016-01-01

    The platinum drugs, cisplatin, carboplatin, and oxaliplatin, prevail in the treatment of cancer,, but new platinum agents have been very slow to enter the clinic. Recently, however, there has been a surge of activity, based on a great deal of mechanistic information, aimed at developing non-classical platinum complexes that operate via mechanisms of action distinct from those of the approved drugs. The use of nanodelivery devices has also grown and many different strategies have been explored to incorporate platinum warheads into nanomedicine constructs. In this review, we discuss these efforts to create the next generation of platinum anticancer drugs. The introduction provides the reader with a brief overview of the use, development, and mechanism of action of the approved platinum drugs to provide the context in which more recent research has flourished. We then describe approaches that explore non-classical platinum(II) complexes with trans geometry and with a monofunctional coordination mode, polynuclear platinum(II) compounds, platinum(IV) prodrugs, dual-treat agents, and photoactivatable platinum(IV) complexes. Nanodelivery particles designed to deliver platinum(IV) complexes will also be discussed, including carbon nanotubes, carbon nanoparticles, gold nanoparticles, quantum dots, upconversion nanoparticles, and polymeric micelles. Additional nanoformulations including supramolecular self-assembled structures, proteins, peptides, metal-organic frameworks, and coordination polymers will then be described. Finally, the significant clinical progress made by nanoparticle formulations of platinum(II) agents will be reviewed. We anticipate that such a synthesis of disparate research efforts will not only help to generate new drug development ideas and strategies, but also reflect our optimism that the next generation of platinum cancer drugs is about to arrive. PMID:26865551

  4. Challenges in the Development of Advanced Reactors

    SciTech Connect

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

    2012-08-01

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

  5. Generation mechanism of the slowly drifting narrowband structure in the type IV solar radio bursts observed by AMATERAS

    SciTech Connect

    Katoh, Y.; Nishimura, Y.; Kumamoto, A.; Ono, T.; Iwai, K.; Misawa, H.; Tsuchiya, F.

    2014-05-20

    We investigate the type IV burst event observed by AMATERAS on 2011 June 7, and reveal that the main component of the burst was emitted from the plasmoid eruption identified in the EUV images of the Solar Dynamics Observatory (SDO)/AIA. We show that a slowly drifting narrowband structure (SDNS) appeared in the burst's spectra. Using statistical analysis, we reveal that the SDNS appeared for a duration of tens to hundreds of milliseconds and had a typical bandwidth of 3 MHz. To explain the mechanism generating the SDNS, we propose wave-wave coupling between Langmuir waves and whistler-mode chorus emissions generated in a post-flare loop, which were inferred from the similarities in the plasma environments of a post-flare loop and the equatorial region of Earth's inner magnetosphere. We assume that a chorus element with a rising tone is generated at the top of a post-flare loop. Using the magnetic field and plasma density models, we quantitatively estimate the expected duration of radio emissions generated from coupling between Langmuir waves and chorus emissions during their propagation in the post-flare loop, and we find that the observed duration and bandwidth properties of the SDNS are consistently explained by the proposed generation mechanism. While observations in the terrestrial magnetosphere show that the chorus emissions are a group of large-amplitude wave elements generated naturally and intermittently, the mechanism proposed in the present study can explain both the intermittency and the frequency drift in the observed spectra.

  6. Tying the knot with next-generation reactors: Can the industry afford a second marriage

    SciTech Connect

    Not Available

    1993-01-01

    This article examines the future of nuclear power beyond the year 2000. The nuclear industry just celebrated 50 years of nuclear technology, but no new plants have been ordered in the US since 1978 and some European countries are giving up on the nuclear option. This article discusses the four US advanced light-water reactor design and safety features, specific design features and parameters for the advanced designs, advanced designs from Europe, features utilities look for in a reactor, evolutionary versus passive designs, gaining public acceptance for new designs, and what alternatives are there to installing next-generation nuclear systems

  7. Study of hydrogen generation plant coupled to high temperature gas cooled reactor

    NASA Astrophysics Data System (ADS)

    Brown, Nicholas Robert

    Hydrogen generation using a high temperature nuclear reactor as a thermal driving vector is a promising future option for energy carrier production. In this scheme, the heat from the nuclear reactor drives an endothermic water-splitting plant, via coupling, through an intermediate heat exchanger. While both high temperature nuclear reactors and hydrogen generation plants have high individual degrees of development, study of the coupled plant is lacking. Particularly absent are considerations of the transient behavior of the coupled plant, as well as studies of the safety of the overall plant. The aim of this document is to contribute knowledge to the effort of nuclear hydrogen generation. In particular, this study regards identification of safety issues in the coupled plant and the transient modeling of some leading candidates for implementation in the Nuclear Hydrogen Initiative (NHI). The Sulfur Iodine (SI) and Hybrid Sulfur (HyS) cycles are considered as candidate hydrogen generation schemes. Several thermodynamically derived chemical reaction chamber models are coupled to a well-known reference design of a high temperature nuclear reactor. These chemical reaction chamber models have several dimensions of validation, including detailed steady state flowsheets, integrated loop test data, and bench scale chemical kinetics. Eight unique case studies are performed based on a thorough literature review of possible events. The case studies are: (1) feed flow failure from one section of the chemical plant to another, (2) product flow failure (recycle) within the chemical plant, (3) rupture or explosion within the chemical plant, (4) nuclear reactor helium inlet overcooling due to a process holding tank failure, (5) helium inlet overcooling as an anticipated transient without SCRAM, (6) total failure of the chemical plant, (7) parametric study of the temperature in an individual reaction chamber, and (8) control rod insertion in the nuclear reactor. Various parametric

  8. Feasibility of Burning First- and Second-Generation Plutonium in Pebble Bed High-Temperature Reactors

    SciTech Connect

    Haas, J.B.M. de; Kuijper, J.C

    2005-08-15

    The core physics investigations at the Nuclear Research Consultancy Group in the Netherlands, as part of the activities within the HTR-N project of the European Fifth Framework Program, are focused on the incineration of pure (first- and second-generation) Pu fuels in the reference pebble bed high-temperature gas-cooled reactor (HTR) HTR-MODUL with a continuous reload [MEDUL, (MEhrfach DUrchLauf, multipass)] fueling strategy in which the spherical fuel elements, or pebbles, pass through the core a number of times before being permanently discharged. For pebbles fueled with different loadings of plutonium, the feasibility of a sustained fuel cycle under nominal reactor conditions was investigated by means of the reactivity and temperature coefficients of the reactor. The HTR-MODUL was found to be a very effective reactor to reduce the stockpile of first-generation plutonium. It reduces the amount of plutonium to about one-sixth of the original and reduces the risk of proliferation by denaturing the plutonium vector. For second-generation plutonium the incineration is less favorable, as the amount of plutonium is only halved.

  9. Group Constants Generation of the Pseudo Fission Products for Fast Reactor Burnup Calculations

    SciTech Connect

    Gil, Choong-Sup; Kim, Do Heon; Chang, Jonghwa

    2005-05-24

    The pseudo fission products for the burnup calculations of the liquid metal fast reactor were generated. The cross-section data and fission product yield data of ENDF/B-VI were used for the pseudo fission product data of U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242. The pseudo fission product data can be used with the KAFAX-F22 or -E66, which are the MATXS-format libraries for analyses of the liquid metal fast reactor at KAERI and were distributed through the OECD/NEA. The 80-group MATXS-format libraries of the 172 fission products were generated and the burnup chains for generation of the pseudo fission products were prepared.

  10. Group Constants Generation of the Pseudo Fission Products for Fast Reactor Burnup Calculations

    NASA Astrophysics Data System (ADS)

    Gil, Choong-Sup; Kim, Do Heon; Chang, Jonghwa

    2005-05-01

    The pseudo fission products for the burnup calculations of the liquid metal fast reactor were generated. The cross-section data and fission product yield data of ENDF/B-VI were used for the pseudo fission product data of U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242. The pseudo fission product data can be used with the KAFAX-F22 or -E66, which are the MATXS-format libraries for analyses of the liquid metal fast reactor at KAERI and were distributed through the OECD/NEA. The 80-group MATXS-format libraries of the 172 fission products were generated and the burnup chains for generation of the pseudo fission products were prepared.

  11. New Materials for NGNP/Gen IV

    SciTech Connect

    Robert W. Swindeman; Douglas L. Marriott

    2009-12-18

    The bounding conditions were briefly summarized for the Next Generation Nuclear Plant (NGNP) that is the leading candidate in the Department of Energy Generation IV reactor program. Metallic materials essential to the successful development and proof of concept for the NGNP were identified. The literature bearing on the materials technology for high-temperature gas-cooled reactors was reviewed with emphasis on the needs identified for the NGNP. Several materials were identified for a more thorough study of their databases and behavioral features relative to the requirements ASME Boiler and Pressure Vessel Code, Section III, Division 1, Subsection NH.

  12. The Next Generation Nuclear Plant Graphite Creep Experiment Irradiation in the Advanced Test Reactor

    SciTech Connect

    Blaine Grover

    2010-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will have differing compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during initial start-up of

  13. Selection of the reference steam generator for the advanced liquid metal reactor

    SciTech Connect

    Loewen, Eric P.; Boardman, Chuck

    2007-07-01

    In February 2006 President Bush announced the Advanced Energy Initiative, which included the Department of Energy's (DOE) Global Nuclear Energy Partnership (GNEP). GNEP has seven broad goals; one of the major elements being to develop and deploy advanced nuclear fuel recycling technology that includes consuming spent nuclear fuel in an Advanced Recycling Reactor (ARR). DOE is contemplating accelerating the deployment of these technologies to achieve the construction of a commercial scale application of these technologies. DOE now defines this approach as 'two simultaneous tracks: (1) deployment of commercial scale facilities for which advanced technologies are available now or in the near future, and (2) further research and development of transmutation fuels technologies'. GEHitachi Nuclear Energy Americas LLC (GHNEA) believes an integrated technical solution is achievable in the near term to accelerate the commercial demonstration of GNEP infrastructure. The GHNEA ARR concept involves a single integrated recycling facility sized to service a single reactor module ARR capable of destroying light water and fast reactor sourced actinides. This paper describes the bases and rationale behind the selection of the helical coil steam generator (HCSG) as the reference steam generator concept for the ALMR and S-PRISM reactor concepts. (authors)

  14. REACTOR

    DOEpatents

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  15. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  16. Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants

    SciTech Connect

    Woo, H.H.; Lu, S.C.

    1981-09-15

    Objective is to assess the reliability of steam generators of pressurized water reactor (PWR) power plants in the United States and abroad. The assessment is based on operation experience of both domestic and foreign PWR plants. The approach taken is to collect and review papers and reports available from the literature as well as information obtained by contacting research institutes both here and abroad. This report presents the results of the assessment. It contains a general background of PWR plant operations, plant types, and materials used in PWR plants. A review of the worldwide distribution of PWR plants is also given. The report describes in detail the degradation problems discovered in PWR steam generators: their causes, their impacts on the performance of steam generators, and the actions to mitigate and avoid them. One chapter is devoted to operating experience of PWR steam generators in foreign countries. Another discusses the improvements in future steam generator design.

  17. A Comparison of the Safety Analysis Process and the Generation IV Proliferation Resistance/Physical Protection Assessment Methodology

    SciTech Connect

    T. A. Bjornard; M. D. Zentner

    2006-05-01

    The Generation IV International Forum (GIF) is a vehicle for the cooperative international development of future nuclear energy systems. The Generation IV program has established primary objectives in the areas of sustainability, economics, safety and reliability, and Proliferation Resistance and Physical Protection (PR&PP). In order to help meet the latter objective a program was launched in December 2002 to develop a rigorous means to assess nuclear energy systems with respect to PR&PP. The study of Physical Protection of a facility is a relatively well established methodology, but an approach to evaluate the Proliferation Resistance of a nuclear fuel cycle is not. This paper will examine the Proliferation Resistance (PR) evaluation methodology being developed by the PR group, which is largely a new approach and compare it to generally accepted nuclear facility safety evaluation methodologies. Safety evaluation methods have been the subjects of decades of development and use. Further, safety design and analysis is fairly broadly understood, as well as being the subject of federally mandated procedures and requirements. It is therefore extremely instructive to compare and contrast the proposed new PR evaluation methodology process with that used in safety analysis. By so doing, instructive and useful conclusions can be derived from the comparison that will help to strengthen the PR methodological approach as it is developed further. From the comparison made in this paper it is evident that there are very strong parallels between the two processes. Most importantly, it is clear that the proliferation resistance aspects of nuclear energy systems are best considered beginning at the very outset of the design process. Only in this way can the designer identify and cost effectively incorporate intrinsic features that might be difficult to implement at some later stage. Also, just like safety, the process to implement proliferation resistance should be a dynamic

  18. Evaluation of an Integrated Gas-Cooled Reactor Simulator and Brayton Turbine-Generator

    NASA Technical Reports Server (NTRS)

    Hissam, David Andy; Stewart, Eric T.

    2006-01-01

    A closed-loop brayton cycle, powered by a fission reactor, offers an attractive option for generating both planetary and in-space electric power. Non-nuclear testing of this type of system provides the opportunity to safely work out integration and system control challenges for a modest investment. Recognizing this potential, a team at Marshall Space Flight Center has evaluated the viability of integrating and testing an existing gas-cooled reactor simulator and a modified commercially available, off-the-shelf, brayton turbine-generator. Since these two systems were developed independently of one another, this evaluation had to determine if they could operate together at acceptable power levels, temperatures, and pressures. Thermal, fluid, and structural analyses show that this combined system can operate at acceptable power levels and temperatures. In addition, pressure drops across the reactor simulator, although higher than desired, are also viewed as acceptable. Three potential working fluids for the system were evaluated: N2, He/Ar, and He/Xe. Other potential issues, such as electrical breakdown in the generator and the operation of the brayton foil bearings using various gas mixtures, were also investigated.

  19. Experimental and computational studies of thermal mixing in next generation nuclear reactors

    NASA Astrophysics Data System (ADS)

    Landfried, Douglas Tyler

    The Very High Temperature Reactor (VHTR) is a proposed next generation nuclear power plant. The VHTR utilizes helium as a coolant in the primary loop of the reactor. Helium traveling through the reactor mixes below the reactor in a region known as the lower plenum. In this region there exists large temperature and velocity gradients due to non-uniform heat generation in the reactor core. Due to these large gradients, concern should be given to reducing thermal striping in the lower plenum. Thermal striping is the phenomena by which temperature fluctuations in the fluid and transferred to and attenuated by surrounding structures. Thermal striping is a known cause of long term material failure. To better understand and predict thermal striping in the lower plenum two separate bodies of work have been conducted. First, an experimental facility capable of predictably recreating some aspects of flow in the lower plenum is designed according to scaling analysis of the VHTR. Namely the facility reproduces jets issuing into a crossflow past a tube bundle. Secondly, extensive studies investigate the mixing of a non-isothermal parallel round triple-jet at two jet-to-jet spacings was conducted. Experimental results were validation with an open source computational fluid dynamics package, OpenFOAMRTM. Additional care is given to understanding the implementation of the realizable k-a and Launder Gibson RSM turbulence Models in OpenFOAMRTM. In order to measure velocity and temperature in the triple-jet experiment a detailed investigation of temperature compensated hotwire anemometry is carried out with special concern being given to quantify the error with the measurements. Finally qualitative comparisons of trends in the experimental results and the computational results is conducted. A new and unexpected physical behavior was observed in the center jet as it appeared to spread unexpectedly for close spacings (S/Djet = 1.41).

  20. The key-role of instrumentation for the new generation of research reactors

    SciTech Connect

    Bignan, G.; Villard, J. F.; Destouches, C.; Baeten, P.; Vermeeren, L.; Michiels, S.

    2011-07-01

    Experimental reactors have been indispensable since the beginning of the use of nuclear energy to support many important fields of industry and research: safety, lifetime management and operation optimisation of nuclear power plants, development of new types of reactors with improved resources and fuel cycle management, medical applications, material development for fusion... Over the last decade, modifications of the operational needs and the ageing of the nuclear facilities have led to several closures and time is coming for new key European Experimental Reactors (EER) within a European and International Framework. Projects like MYRRHA and JHR are underway to define and implement a new consistent EER policy: - Meeting industry and public needs, keeping a high level of scientific expertise; - With a limited number of EER, specified within a rational compromise between specialisation, complementarities and back-up capacities; - To be put into effective operation in this or the next decade. These new projects will give to the scientific community high performances allowing innovative fields of R and D. A new generation of instrumentation to address new phenomena and that allows better on-line investigation of some key physical parameters is necessary to achieve these challenges. One initiative to progress in this direction is the Joint Instrumentation Laboratory between CEA and SCK.CEN which has already given significant results and patents. Major scientific challenges to achieve in the field of instrumentation for this new generation of European Research Reactors have to be investigated and are described in this paper as well as a short description of the JHR and MYRRHA reactors that will be serving as flexible irradiation facilities for testing them. (authors)

  1. Generation IV Nuclear Energy Systems Construction Cost Reductions Through the Use of Virtual Environments

    SciTech Connect

    Timothy Shaw; Vaugh Whisker

    2004-02-28

    The objective of this multi-phase project is to demonstrate the feasibility and effectiveness of using full-scale virtual reality simulation in the design, construction, and maintenance of future nuclear power plants. The project will test the suitability of immersive virtual reality technology to aid engineers in the design of the next generation nuclear power plant and to evaluate potential cost reductions that can be realized by optimization of installation and construction sequences. The intent is to see if this type of information technology can be used in capacities similar to those currently filled by full-scale physical mockups. This report presents the results of the completed project.

  2. Evaluation Metrics for Intermediate Heat Exchangers for Next Generation Nuclear Reactors

    SciTech Connect

    Piyush Sabharwall; Eung Soo Kim; Nolan Anderson

    2011-06-01

    The Department of Energy (DOE) is working with industry to develop a next generation, high-temperature gas-cooled reactor (HTGR) as a part of the effort to supply the United States with abundant, clean, and secure energy as initiated by the Energy Policy Act of 2005 (EPAct; Public Law 109-58,2005). The NGNP Project, led by the Idaho National Laboratory (INL), will demonstrate the ability of the HTGR to generate hydrogen, electricity, and/or high-quality process heat for a wide range of industrial applications.

  3. Development and Transient Analysis of a Helical-coil Steam Generator for High Temperature Reactors

    SciTech Connect

    Nathan V. Hoffer; Nolan A. Anderson; Piyush Sabharwall

    2011-08-01

    A high temperature gas-cooled reactor (HTGR) is under development by the Next Generation Nuclear Plant (NGNP) Project at the Idaho National Laboratory (INL). Its design emphasizes electrical power production which may potentially be coupled with process heat for hydrogen production and other industrial applications. NGNP is considering a helical-coil steam generator for the primary heat transport loop heat exchanger based on its increased heat transfer and compactness when compared to other steam generators. The safety and reliability of the helical-coil steam generator is currently under evaluation as part of the development of NGNP. Transients, such as loss of coolant accidents (LOCA), are of interest in evaluating the safety of steam generators. In this study, a complete steam generator inlet pipe break (double ended pipe break) LOCA was simulated by an exponential loss of primary side pressure. For this analysis, a model of the helical-coil steam generator was developed using RELAP5-3D, an INL inhouse systems analysis code. The steam generator model behaved normally during the transient simulating the complete steam generator inlet pipe break LOCA. Further analysis is required to comprehensively evaluate the safety and reliability of the helical-coil steam generator design in the NGNP setting.

  4. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  5. THE NEXT GENERATION VIRGO CLUSTER SURVEY. IV. NGC 4216: A BOMBARDED SPIRAL IN THE VIRGO CLUSTER

    SciTech Connect

    Paudel, Sanjaya; Duc, Pierre-Alain; Ferriere, Etienne; Cuillandre, Jean-Charles; Mihos, J. Christopher; Vollmer, Bernd; Balogh, Michael L.; Carlberg, Ray G.; Boissier, Samuel; Boselli, Alessandro; Durrell, Patrick R.; Emsellem, Eric; Michel-Dansac, Leo; Mei, Simona; Van Driel, Wim

    2013-04-20

    The final stages of mass assembly of present-day massive galaxies are expected to occur through the accretion of multiple satellites. Cosmological simulations thus predict a high frequency of stellar streams resulting from this mass accretion around the massive galaxies in the Local Volume. Such tidal streams are difficult to observe, especially in dense cluster environments, where they are readily destroyed. We present an investigation into the origins of a series of interlaced narrow filamentary stellar structures, loops and plumes in the vicinity of the Virgo Cluster, edge-on spiral galaxy, NGC 4216 that were previously identified by the Blackbird telescope. Using the deeper, higher-resolution, and precisely calibrated optical CFHT/MegaCam images obtained as part of the Next Generation Virgo Cluster Survey (NGVS), we confirm the previously identified features and identify a few additional structures. The NGVS data allowed us to make a physical study of these low surface brightness features and investigate their origin. The likely progenitors of the structures were identified as either already cataloged Virgo Cluster Catalog dwarfs or newly discovered satellites caught in the act of being destroyed. They have the same g - i color index and likely contain similar stellar populations. The alignment of three dwarfs along an apparently single stream is intriguing, and we cannot totally exclude that these are second-generation dwarf galaxies being born inside the filament from the debris of an original dwarf. The observed complex structures, including in particular a stream apparently emanating from a satellite of a satellite, point to a high rate of ongoing dwarf destruction/accretion in the region of the Virgo Cluster where NGC 4216 is located. We discuss the age of the interactions and whether they occurred in a group that is just falling into the cluster and shows signs of the so-called pre-processing before it gets affected by the cluster environment, or in a

  6. Anomalous independence of multiple exciton generation on different group IV-VI quantum dot architectures.

    PubMed

    Trinh, M Tuan; Polak, Leo; Schins, Juleon M; Houtepen, Arjan J; Vaxenburg, Roman; Maikov, Georgy I; Grinbom, Gal; Midgett, Aaron G; Luther, Joseph M; Beard, Matthew C; Nozik, Arthur J; Bonn, Mischa; Lifshitz, Efrat; Siebbeles, Laurens D A

    2011-04-13

    Multiple exciton generation (MEG) in PbSe quantum dots (QDs), PbSe(x)S(1-x) alloy QDs, PbSe/PbS core/shell QDs, and PbSe/PbSe(y)S(1-y) core/alloy-shell QDs was studied with time-resolved optical pump and probe spectroscopy. The optical absorption exhibits a red-shift upon the introduction of a shell around a PbSe core, which increases with the thickness of the shell. According to electronic structure calculations this can be attributed to charge delocalization into the shell. Remarkably, the measured quantum yield of MEG, the hot exciton cooling rate, and the Auger recombination rate of biexcitons are similar for pure PbSe QDs and core/shell QDs with the same core size and varying shell thickness. The higher density of states in the alloy and core/shell QDs provide a faster exciton cooling channel that likely competes with the fast MEG process due to a higher biexciton density of states. Calculations reveal only a minor asymmetric delocalization of holes and electrons over the entire core/shell volume, which may partially explain why the Auger recombination rate does not depend on the presence of a shell. PMID:21348493

  7. The Next Generation Virgo Cluster Survey. IV. NGC 4216: A Bombarded Spiral in the Virgo Cluster

    NASA Astrophysics Data System (ADS)

    Paudel, Sanjaya; Duc, Pierre-Alain; Côté, Patrick; Cuillandre, Jean-Charles; Ferrarese, Laura; Ferriere, Etienne; Gwyn, Stephen D. J.; Mihos, J. Christopher; Vollmer, Bernd; Balogh, Michael L.; Carlberg, Ray G.; Boissier, Samuel; Boselli, Alessandro; Durrell, Patrick R.; Emsellem, Eric; MacArthur, Lauren A.; Mei, Simona; Michel-Dansac, Leo; van Driel, Wim

    2013-04-01

    The final stages of mass assembly of present-day massive galaxies are expected to occur through the accretion of multiple satellites. Cosmological simulations thus predict a high frequency of stellar streams resulting from this mass accretion around the massive galaxies in the Local Volume. Such tidal streams are difficult to observe, especially in dense cluster environments, where they are readily destroyed. We present an investigation into the origins of a series of interlaced narrow filamentary stellar structures, loops and plumes in the vicinity of the Virgo Cluster, edge-on spiral galaxy, NGC 4216 that were previously identified by the Blackbird telescope. Using the deeper, higher-resolution, and precisely calibrated optical CFHT/MegaCam images obtained as part of the Next Generation Virgo Cluster Survey (NGVS), we confirm the previously identified features and identify a few additional structures. The NGVS data allowed us to make a physical study of these low surface brightness features and investigate their origin. The likely progenitors of the structures were identified as either already cataloged Virgo Cluster Catalog dwarfs or newly discovered satellites caught in the act of being destroyed. They have the same g - i color index and likely contain similar stellar populations. The alignment of three dwarfs along an apparently single stream is intriguing, and we cannot totally exclude that these are second-generation dwarf galaxies being born inside the filament from the debris of an original dwarf. The observed complex structures, including in particular a stream apparently emanating from a satellite of a satellite, point to a high rate of ongoing dwarf destruction/accretion in the region of the Virgo Cluster where NGC 4216 is located. We discuss the age of the interactions and whether they occurred in a group that is just falling into the cluster and shows signs of the so-called pre-processing before it gets affected by the cluster environment, or in a

  8. Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression

    DOEpatents

    Lasche, G.P.

    1983-09-29

    The invention is a laser or particle-beam-driven fusion reactor system which takes maximum advantage of both the very short pulsed nature of the energy release of inertial confinement fusion (ICF) and the very small volumes within which the thermonuclear burn takes place. The pulsed nature of ICF permits dynamic direct energy conversion schemes such as magnetohydrodynamic (MHD) generation and magnetic flux compression; the small volumes permit very compact blanket geometries. By fully exploiting these characteristics of ICF, it is possible to design a fusion reactor with exceptionally high power density, high net electric efficiency, and low neutron-induced radioactivity. The invention includes a compact blanket design and method and apparatus for obtaining energy utilizing the compact blanket.

  9. Development of Liquid-Vapor Core Reactors with MHD Generator for Space Power and Propulsion Applications

    SciTech Connect

    Samim Anghaie

    2002-08-13

    Any reactor that utilizes fuel consisting of a fissile material in a gaseous state may be referred to as a gaseous core reactor (GCR). Studies on GCRs have primarily been limited to the conceptual phase, mostly due to budget cuts and program cancellations in the early 1970's. A few scientific experiments have been conducted on candidate concepts, primarily of static pressure fissile gas filling a cylindrical or spherical cavity surrounded by a moderating shell, such as beryllium, heavy water, or graphite. The main interest in this area of nuclear power generation is for space applications. The interest in space applications has developed due to the promise of significant enhancement in fuel utilization, safety, plant efficiency, special high-performance features, load-following capabilities, power conversion optimization, and other key aspects of nuclear power generation. The design of a successful GCR adapted for use in space is complicated. The fissile material studied in the pa st has been in a fluorine compound, either a tetrafluoride or a hexafluoride. Both of these molecules have an impact on the structural material used in the making of a GCR. Uranium hexafluoride as a fuel allows for a lower operating temperature, but at temperatures greater than 900K becomes essentially impossible to contain. This difficulty with the use of UF6 has caused engineers and scientists to use uranium tetrafluoride, which is a more stable molecule but has the disadvantage of requiring significantly higher operating temperatures. Gas core reactors have traditionally been studied in a steady state configuration. In this manner a fissile gas and working fluid are introduced into the core, called a cavity, that is surrounded by a reflector constructed of materials such as Be or BeO. These reactors have often been described as cavity reactors because the density of the fissile gas is low and criticality is achieved only by means of the reflector to reduce neutron leakage from the core

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

    SciTech Connect

    Impellezzeri, J.R.; Camaret, T.L.; Friske, W.H.

    1981-03-11

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

  11. Multi-scale approach to the modeling of fission gas discharge during hypothetical loss-of-flow accident in gen-IV sodium fast reactor

    SciTech Connect

    Behafarid, F.; Shaver, D. R.; Bolotnov, I. A.; Jansen, K. E.; Antal, S. P.; Podowski, M. Z.

    2012-07-01

    The required technological and safety standards for future Gen IV Reactors can only be achieved if advanced simulation capabilities become available, which combine high performance computing with the necessary level of modeling detail and high accuracy of predictions. The purpose of this paper is to present new results of multi-scale three-dimensional (3D) simulations of the inter-related phenomena, which occur as a result of fuel element heat-up and cladding failure, including the injection of a jet of gaseous fission products into a partially blocked Sodium Fast Reactor (SFR) coolant channel, and gas/molten sodium transport along the coolant channels. The computational approach to the analysis of the overall accident scenario is based on using two different inter-communicating computational multiphase fluid dynamics (CMFD) codes: a CFD code, PHASTA, and a RANS code, NPHASE-CMFD. Using the geometry and time history of cladding failure and the gas injection rate, direct numerical simulations (DNS), combined with the Level Set method, of two-phase turbulent flow have been performed by the PHASTA code. The model allows one to track the evolution of gas/liquid interfaces at a centimeter scale. The simulated phenomena include the formation and breakup of the jet of fission products injected into the liquid sodium coolant. The PHASTA outflow has been averaged over time to obtain mean phasic velocities and volumetric concentrations, as well as the liquid turbulent kinetic energy and turbulence dissipation rate, all of which have served as the input to the core-scale simulations using the NPHASE-CMFD code. A sliding window time averaging has been used to capture mean flow parameters for transient cases. The results presented in the paper include testing and validation of the proposed models, as well the predictions of fission-gas/liquid-sodium transport along a multi-rod fuel assembly of SFR during a partial loss-of-flow accident. (authors)

  12. System and method for generating steady state confining current for a toroidal plasma fusion reactor

    DOEpatents

    Bers, Abraham

    1981-01-01

    A system for generating steady state confining current for a toroidal plasma fusion reactor providing steady-state generation of the thermonuclear power. A dense, hot toroidal plasma is initially prepared with a confining magnetic field with toroidal and poloidal components. Continuous wave RF energy is injected into said plasma to estalish a spectrum of traveling waves in the plasma, where the traveling waves have momentum components substantially either all parallel, or all anti-parallel to the confining magnetic field. The injected RF energy is phased to couple to said traveling waves with both a phase velocity component and a wave momentum component in the direction of the plasma traveling wave components. The injected RF energy has a predetermined spectrum selected so that said traveling waves couple to plasma electrons having velocities in a predetermined range .DELTA.. The velocities in the range are substantially greater than the thermal electron velocity of the plasma. In addition, the range is sufficiently broad to produce a raised plateau having width .DELTA. in the plasma electron velocity distribution so that the plateau electrons provide steady-state current to generate a poloidal magnetic field component sufficient for confining the plasma. In steady state operation of the fusion reactor, the fusion power density in the plasma exceeds the power dissipated inthe plasma.

  13. System and method for generating steady state confining current for a toroidal plasma fusion reactor

    DOEpatents

    Fisch, Nathaniel J.

    1981-01-01

    A system for generating steady state confining current for a toroidal plasma fusion reactor providing steady-state generation of the thermonuclear power. A dense, hot toroidal plasma is initially prepared with a confining magnetic field with toroidal and poloidal components. Continuous wave RF energy is injected into said plasma to establish a spectrum of traveling waves in the plasma, where the traveling waves have momentum components substantially either all parallel, or all anti-parallel to the confining magnetic field. The injected RF energy is phased to couple to said traveling waves with both a phase velocity component and a wave momentum component in the direction of the plasma traveling wave components. The injected RF energy has a predetermined spectrum selected so that said traveling waves couple to plasma electrons having velocities in a predetermined range .DELTA.. The velocities in the range are substantially greater than the thermal electron velocity of the plasma. In addition, the range is sufficiently broad to produce a raised plateau having width .DELTA. in the plasma electron velocity distribution so that the plateau electrons provide steady-state current to generate a poloidal magnetic field component sufficient for confining the plasma. In steady state operation of the fusion reactor, the fusion power density in the plasma exceeds the power dissipated in the plasma.

  14. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    SciTech Connect

    J. K. Wright; R. N. Wright

    2008-04-01

    The U.S. Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic, or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development Program is responsible for performing research and development on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. Studies of potential Reactor Pressure Vessel (RPV) steels have been carried out as part of the pre-conceptual design studies. These design studies generally focus on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Three realistic candidate materials have been identified by this process: conventional light water reactor RPV steels A508/533, 2¼Cr-1Mo in the annealed condition, and modified 9Cr 1Mo ferritic martenistic steel. Based on superior strength and higher temperature limits, the modified 9Cr-1Mo steel has been identified by the majority of design engineers as the preferred choice for the RPV. All of the vendors have

  15. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    SciTech Connect

    J. K. Wright; R. N. Wright

    2010-07-01

    The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750°C, and a design service life of 60 years. The reactor design will be a graphite-moderated, helium-cooled, prismatic, or pebble bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.

  16. Laboratory-Scale Membrane Reactor for the Generation of Anhydrous Diazomethane.

    PubMed

    Dallinger, Doris; Pinho, Vagner D; Gutmann, Bernhard; Kappe, C Oliver

    2016-07-15

    A configurationally simple and robust semibatch apparatus for the in situ on-demand generation of anhydrous solutions of diazomethane (CH2N2) avoiding distillation methods is presented. Diazomethane is produced by base-mediated decomposition of commercially available Diazald within a semipermeable Teflon AF-2400 tubing and subsequently selectively separated from the tubing into a solvent- and substrate-filled flask (tube-in-flask reactor). Reactions with CH2N2 can therefore be performed directly in the flask without dangerous and labor-intensive purification operations or exposure of the operator to CH2N2. The reactor has been employed for the methylation of carboxylic acids, the synthesis of α-chloro ketones and pyrazoles, and palladium-catalyzed cyclopropanation reactions on laboratory scale. The implementation of in-line FTIR technology allowed monitoring of the CH2N2 generation and its consumption. In addition, larger scales (1.8 g diazomethane per hour) could be obtained via parallelization (numbering up) by simply wrapping several membrane tubings into the flask.

  17. Laboratory-Scale Membrane Reactor for the Generation of Anhydrous Diazomethane.

    PubMed

    Dallinger, Doris; Pinho, Vagner D; Gutmann, Bernhard; Kappe, C Oliver

    2016-07-15

    A configurationally simple and robust semibatch apparatus for the in situ on-demand generation of anhydrous solutions of diazomethane (CH2N2) avoiding distillation methods is presented. Diazomethane is produced by base-mediated decomposition of commercially available Diazald within a semipermeable Teflon AF-2400 tubing and subsequently selectively separated from the tubing into a solvent- and substrate-filled flask (tube-in-flask reactor). Reactions with CH2N2 can therefore be performed directly in the flask without dangerous and labor-intensive purification operations or exposure of the operator to CH2N2. The reactor has been employed for the methylation of carboxylic acids, the synthesis of α-chloro ketones and pyrazoles, and palladium-catalyzed cyclopropanation reactions on laboratory scale. The implementation of in-line FTIR technology allowed monitoring of the CH2N2 generation and its consumption. In addition, larger scales (1.8 g diazomethane per hour) could be obtained via parallelization (numbering up) by simply wrapping several membrane tubings into the flask. PMID:27359257

  18. CO2 emission free co-generation of energy and ethylene in hydrocarbon SOFC reactors with a dehydrogenation anode.

    PubMed

    Fu, Xian-Zhu; Lin, Jie-Yuan; Xu, Shihong; Luo, Jing-Li; Chuang, Karl T; Sanger, Alan R; Krzywicki, Andrzej

    2011-11-21

    A dehydrogenation anode is reported for hydrocarbon proton conducting solid oxide fuel cells (SOFCs). A Cu-Cr(2)O(3) nanocomposite is obtained from CuCrO(2) nanoparticles as an inexpensive, efficient, carbon deposition and sintering tolerant anode catalyst. A SOFC reactor is fabricated using a Cu-Cr(2)O(3) composite as a dehydrogenation anode and a doped barium cerate as a proton conducting electrolyte. The protonic membrane SOFC reactor can selectively convert ethane to valuable ethylene, and electricity is simultaneously generated in the electrochemical oxidative dehydrogenation process. While there are no CO(2) emissions, traces of CO are present in the anode exhaust when the SOFC reactor is operated at over 700 °C. A mechanism is proposed for ethane electro-catalytic dehydrogenation over the Cu-Cr(2)O(3) catalyst. The SOFC reactor also has good stability for co-generation of electricity and ethylene at 700 °C. PMID:21984357

  19. Modular Pebble-Bed Reactor Project: Laboratory-Directed Research and Development Program FY 2002 Annual Report

    SciTech Connect

    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.

  20. FAST TRACK COMMUNICATION Generation of stable multi-jets by flow-limited field-injection electrostatic spraying and their control via I-V characteristics

    NASA Astrophysics Data System (ADS)

    Gu, W.; Heil, P. E.; Choi, H.; Kim, K.

    2010-12-01

    The I-V characteristics of flow-limited field-injection electrostatic spraying (FFESS) were investigated, exposing a new way to predict and control the specific spraying modes from single-jet to multi-jet. Monitoring the I-V characteristics revealed characteristic drops in the current upon formation of an additional jet in the multi-jet spraying mode. For fixed jet numbers, space-charge-limited current behaviour was measured which was attributed to space charge in the dielectric liquids between the needle electrode and the nozzle opening. The present work establishes that FFESS can, in particular, generate stable multiple jets and that their control is possible through monitoring the I-V characteristics. This can allow for automatic control of the FFESS process and expedite its future scientific and industrial applications.

  1. ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING DEACTIVATION AND DECOMMISSIONING OF REACTOR VESSELS AT THE SAVANNAH RIVER SITE

    SciTech Connect

    Wiersma, B.; Serrato, M.; Langton, C.

    2010-11-10

    The R- and P-reactor vessels at the Savannah River Site (SRS) are being prepared for deactivation and decommissioning (D&D). D&D activities will consist primarily of physically isolating and stabilizing the reactor vessel by filling it with a grout material. The reactor vessels contain aluminum alloy materials, which pose a concern in that aluminum corrodes rapidly when it comes in contact with the alkaline grout. A product of the corrosion reaction is hydrogen gas and therefore potential flammability issues were assessed. A model was developed to calculate the hydrogen generation rate as the reactor is being filled with the grout material. Three options existed for the type of grout material for D&D of the reactor vessels. The grout formulation options included ceramicrete (pH 6-8), a calcium aluminate sulfate (CAS) based cement (pH 10), or Portland cement grout (pH 12.4). Corrosion data for aluminum in concrete were utilized as input for the model. The calculations considered such factors as the surface area of the aluminum components, the open cross-sectional area of the reactor vessel, the rate at which the grout is added to the reactor vessel, and temperature. Given the hydrogen generation rate, the hydrogen concentration in the vapor space of the reactor vessel above the grout was calculated. This concentration was compared to the lower flammability limit for hydrogen. The assessment concluded that either ceramicrete or the CAS grout may be used to safely grout the P-reactor vessel. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Portland cement grout, on the other hand, for the same range of process parameters did not provide a margin of safety against the accumulation of flammable gas in the reactor vessel during grouting operations in the P-reactor vessel. Therefore, it was recommended that this grout not be utilized for this task. On the other hand, the R-reactor vessel

  2. Reactor

    DOEpatents

    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.

  3. Gen IV Materials Handbook Implementation Plan

    SciTech Connect

    Rittenhouse, P.; Ren, W.

    2005-03-29

    A Gen IV Materials Handbook is being developed to provide an authoritative single source of highly qualified structural materials information and materials properties data for use in design and analyses of all Generation IV Reactor Systems. The Handbook will be responsive to the needs expressed by all of the principal government, national laboratory, and private company stakeholders of Gen IV Reactor Systems. The Gen IV Materials Handbook Implementation Plan provided here addresses the purpose, rationale, attributes, and benefits of the Handbook and will detail its content, format, quality assurance, applicability, and access. Structural materials, both metallic and ceramic, for all Gen IV reactor types currently supported by the Department of Energy (DOE) will be included in the Gen IV Materials Handbook. However, initial emphasis will be on materials for the Very High Temperature Reactor (VHTR). Descriptive information (e.g., chemical composition and applicable technical specifications and codes) will be provided for each material along with an extensive presentation of mechanical and physical property data including consideration of temperature, irradiation, environment, etc. effects on properties. Access to the Gen IV Materials Handbook will be internet-based with appropriate levels of control. Information and data in the Handbook will be configured to allow search by material classes, specific materials, specific information or property class, specific property, data parameters, and individual data points identified with materials parameters, test conditions, and data source. Details on all of these as well as proposed applicability and consideration of data quality classes are provided in the Implementation Plan. Website development for the Handbook is divided into six phases including (1) detailed product analysis and specification, (2) simulation and design, (3) implementation and testing, (4) product release, (5) project/product evaluation, and (6) product

  4. OH radical generation in a photocatalytic reactor using TiO2 nanotube plates.

    PubMed

    Lee, Kangpyung; Ku, Haemin; Pak, Daewon

    2016-04-01

    In order to use TiO2 nanotubes grown on a Ti plate as a photocatalyst, self-organized oxide nanotube layers were grown by anodization in a glycerol based electrolyte. The ultimate conditions for the synthesis of the TiO2 nanotube array on the Ti plate were investigated by comparing the morphology, length, and inner diameter of the nanotubes. They were significantly affected by the applied anodic voltage, anodization time, and composition of the electrolyte such as the water and fluoride ion concentration. The crystallographic structures of TiO2 nanotubes before and after annealing were compared. The photocatalytic reactor used in this study consisted of two parallel and closely spaced TiO2 nanotube plates. The plates were squares while a UV lamp was inserted perpendicularly to them. OH radical generation in the photocatalytic reactor was monitored by using a probe compound, parachlorobenzoate (pCBA). The steady state OH radical concentration was compared depending on the length of nanotubes and crystallographic structure. The longer the nanotubes, the higher the steady state OH radical concentration.

  5. OH radical generation in a photocatalytic reactor using TiO2 nanotube plates.

    PubMed

    Lee, Kangpyung; Ku, Haemin; Pak, Daewon

    2016-04-01

    In order to use TiO2 nanotubes grown on a Ti plate as a photocatalyst, self-organized oxide nanotube layers were grown by anodization in a glycerol based electrolyte. The ultimate conditions for the synthesis of the TiO2 nanotube array on the Ti plate were investigated by comparing the morphology, length, and inner diameter of the nanotubes. They were significantly affected by the applied anodic voltage, anodization time, and composition of the electrolyte such as the water and fluoride ion concentration. The crystallographic structures of TiO2 nanotubes before and after annealing were compared. The photocatalytic reactor used in this study consisted of two parallel and closely spaced TiO2 nanotube plates. The plates were squares while a UV lamp was inserted perpendicularly to them. OH radical generation in the photocatalytic reactor was monitored by using a probe compound, parachlorobenzoate (pCBA). The steady state OH radical concentration was compared depending on the length of nanotubes and crystallographic structure. The longer the nanotubes, the higher the steady state OH radical concentration. PMID:26855214

  6. Evaluation of the Safety Systems in the Next Generation Boiling Water Reactor

    NASA Astrophysics Data System (ADS)

    Cheng, Ling

    The thesis evaluates the safety systems in the next generation boiling water reactor by analyzing the main steam line break loss of coolant accident performed in the Purdue university multi-dimensional test assembly (PUMA). RELAP5 code simulations, both for the PUMA main steam line break (MSLB) case and for the simplified boiling water reactor (SBWR) MSLB case have been utilized to compare with the experiment data. The comparison shows that RELAP5 is capable to perform the safety analysis for SBWR. The comparison also validates the three-level scaling methodology applied to the design of the PUMA facility. The PUMA suppression pool mixing and condensation test data have been studied to give the detailed understanding on this important local phenomenon. A simple one dimensional integral model, which can reasonably simulate the mixing process inside suppression pool have been developed and the comparison between the model prediction and the experiment data demonstrates the model can be utilized for analyzing the suppression pool mixing process.

  7. Synergetic action of domain II and IV underlies persistent current generation in Nav1.3 as revealed by a tarantula toxin.

    PubMed

    Tang, Cheng; Zhou, Xi; Zhang, Yunxiao; Xiao, Zhaohua; Hu, Zhaotun; Zhang, Changxin; Huang, Ying; Chen, Bo; Liu, Zhonghua; Liang, Songping

    2015-01-01

    The persistent current (INaP) through voltage-gated sodium channels enhances neuronal excitability by causing prolonged depolarization of membranes. Nav1.3 intrinsically generates a small INaP, although the mechanism underlying its generation remains unclear. In this study, the involvement of the four domains of Nav1.3 in INaP generation was investigated using the tarantula toxin α-hexatoxin-MrVII (RTX-VII). RTX-VII activated Nav1.3 and induced a large INaP. A pre-activated state binding model was proposed to explain the kinetics of toxin-channel interaction. Of the four domains of Nav1.3, both domain II and IV might play important roles in the toxin-induced INaP. Domain IV constructed the binding site for RTX-VII, while domain II might not participate in interacting with RTX-VII but could determine the efficacy of RTX-VII. Our results based on the use of RTX-VII as a probe suggest that domain II and IV cooperatively contribute to the generation of INaP in Nav1.3. PMID:25784299

  8. Ozone generation by negative direct current corona discharges in dry air fed coaxial wire-cylinder reactors

    SciTech Connect

    Yehia, Ashraf; Mizuno, Akira

    2013-05-14

    An analytical study was made in this paper for calculating the ozone generation by negative dc corona discharges. The corona discharges were formed in a coaxial wire-cylinder reactor. The reactor was fed by dry air flowing with constant rates at atmospheric pressure and room temperature, and stressed by a negative dc voltage. The current-voltage characteristics of the negative dc corona discharges formed inside the reactor were measured in parallel with concentration of the generated ozone under different operating conditions. An empirical equation was derived from the experimental results for calculating the ozone concentration generated inside the reactor. The results, that have been recalculated by using the derived equation, have agreed with the experimental results over the whole range of the investigated parameters, except in the saturation range for the ozone concentration. Therefore, the derived equation represents a suitable criterion for expecting the ozone concentration generated by negative dc corona discharges in dry air fed coaxial wire-cylinder reactors under any operating conditions in range of the investigated parameters.

  9. Dipeptidyl peptidase-IV inhibitory peptides generated by tryptic hydrolysis of a whey protein concentrate rich in β-lactoglobulin.

    PubMed

    Silveira, Silvana T; Martínez-Maqueda, Daniel; Recio, Isidra; Hernández-Ledesma, Blanca

    2013-11-15

    Dipeptidyl peptidase-IV (DPP-IV) is a serine protease involved in the degradation and inactivation of incretin hormones that act by stimulating glucose-dependent insulin secretion after meal ingestion. DPP-IV inhibitors have emerged as new and promising oral agents for the treatment of type 2 diabetes. The purpose of this study was to investigate the potential of β-lactoglobulin as natural source of DPP-IV inhibitory peptides. A whey protein concentrate rich in β-lactoglobulin was hydrolysed with trypsin and fractionated using a chromatographic separation at semipreparative scale. Two of the six collected fractions showed notable DPP-IV inhibitory activity. These fractions were analysed by HPLC coupled to tandem mass spectrometry (HPLC-MS/MS) to identify peptides responsible for the observed activity. The most potent fragment (IPAVF) corresponded to β-lactoglobulin f(78-82) which IC50 value was 44.7μM. The results suggest that peptides derived from β-lactoglobulin would be beneficial ingredients of foods against type 2 diabetes.

  10. The IRIS Spool-Type Reactor Coolant Pump

    SciTech Connect

    Kujawski, J.M.; Kitch, D.M.; Conway, L.E.

    2002-07-01

    IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the major reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long life core and enhanced safety to address the requirements defined by the US DOE for Generation IV reactors. One of the innovative features of the IRIS design is the adoption of a reactor coolant pump (called 'spool' pump) which is completely contained inside the reactor vessel. Background, status and future developments of the IRIS spool pump are presented in this paper. (authors)

  11. Divertor conditions relevant for fusion reactors achieved with linear plasma generator

    SciTech Connect

    Eck, H. J. N. van; Lof, A.; Meiden, H. J. van der; Rooij, G. J. van; Scholten, J.; Zeijlmans van Emmichoven, P. A.; Kleyn, A. W.

    2012-11-26

    Intense magnetized hydrogen and deuterium plasmas have been produced with electron densities up to 3.6 Multiplication-Sign 10{sup 20} m{sup -3} and electron temperatures up to 3.7 eV with a linear plasma generator. Exposure of a W target has led to average heat and particle flux densities well in excess of 4 MW m{sup -2} and 10{sup 24} m{sup -2} s{sup -1}, respectively. We have shown that the plasma surface interactions are dominated by the incoming ions. The achieved conditions correspond very well to the projected conditions at the divertor strike zones of fusion reactors such as ITER. In addition, the machine has an unprecedented high gas efficiency.

  12. Regulatory Concerns on the In-Containment Water Storage System of the Korean Next Generation Reactor

    SciTech Connect

    Ahn, Hyung-Joon; Lee, Jae-Hun; Bang, Young-Seok; Kim, Hho-Jung

    2002-07-15

    The in-containment water storage system (IWSS) is a newly adopted system in the design of the Korean Next Generation Reactor (KNGR). It consists of the in-containment refueling water storage tank, holdup volume tank, and cavity flooding system (CFS). The IWSS has the function of steam condensation and heat sink for the steam release from the pressurizer and provides cooling water to the safety injection system and containment spray system in an accident condition and to the CFS in a severe accident condition. With the progress of the KNGR design, the Korea Institute of Nuclear Safety has been developing Safety and Regulatory Requirements and Guidances for safety review of the KNGR. In this paper, regarding the IWSS of the KNGR, the major contents of the General Safety Criteria, Specific Safety Requirements, Safety Regulatory Guides, and Safety Review Procedures were introduced, and the safety review items that have to be reviewed in-depth from the regulatory viewpoint were also identified.

  13. Neutron Dosimetry on the Full-Core First Generation VVER-440 Aimed at Reactor Support Structure Load Evaluation

    NASA Astrophysics Data System (ADS)

    Borodkin, P.; Borodkin, G.; Khrennikov, N.; Konheiser, J.; Noack, K.

    2009-08-01

    Reactor support structures (RSS), especially the ferritic steel wall of the water tank, of first-generation VVER-440 are non-restorable reactor equipment, and their lifetime may restrict plant-life. All operated Russian first generation VVER-440 have a reduced core with dummy assemblies except Unit 4 of Novovoronezh nuclear power plant (NPP). In comparison with other reactors, the full-core loading scheme of this reactor provides the highest neutron fluence on the reactor pressure vessel (RPV) and RSS accumulated over design service-life and its prolongation. The radiation load parameters on the RPV and RSS that have resulted from this core loading scheme should be evaluated by means of precise calculations and validated by ex-vessel neutron dosimetry to provide the reliable assessment of embrittlement parameters of these reactor components. The results of different types of calculations and their comparison with measured data have been analyzed in this paper. The calculational analysis of RSS fluence rate variation in dependence on the core loading scheme, including the standard and low leakage core as well as the introduction of dummy assemblies, is presented in this paper.

  14. Waste Stream Generated and Waste Disposal Plans for Molten Salt Reactor Experiment at Oak Ridge National Laboratory

    SciTech Connect

    Haghighi, M. H.; Szozda, R. M.; Jugan, M. R.

    2002-02-26

    The Molten Salt Reactor Experiment (MSRE) site is located in Tennessee, on the U.S. Department of Energy (DOE) Oak Ridge Reservation (ORR), south of the Oak Ridge National Laboratory (ORNL) main plant across Haw Ridge in Melton Valley. The MSRE was run by ORNL to demonstrate the desirable features of the molten-salt concept in a practical reactor that could be operated safely and reliably. It introduced the idea of a homogeneous reactor using fuel salt media and graphite moderation for power and breeder reactors. The MSRE reactor and associated components are located in cells beneath the floor in the high-bay area of Building 7503 (Figure 1). The reactor was operated from June 1965 to December 1969. When the reactor was shut down, fuel salt was drained from the reactor circuit to two drain tanks. A ''clean'' salt was then circulated through the reactor as a decontamination measure and drained to a third drain tank. When operations ceased, the fuel and flush salts were allowed t o cool and solidify in the drain tanks. At shutdown, the MSRE facility complex was placed in a surveillance and maintenance program. As a result of the S&M program, it was discovered in 1994 that gaseous uranium (233U/232U) hexafluoride (UF6) had moved throughout the MSRE process systems. The UF6 was generated when radiolysis of the fluorine salts caused the individual constituents to dissociate to their component atoms, including free fluorine.Some of the free fluorine combined with uranium fluorides (UF4) in the salt to form UF6. UF6 is gaseous at slightly above ambient temperatures; thus, periodic heating of the fuel salts (which was intended to remedy the radiolysis problems) and simple diffusion had allowed the UF6 to move out of the salt and into the process systems of MSRE.

  15. Evaluation of the Destruction of the Harmful Cyanobacteria, Microcystis aeruginosa, with a Cavitation and Superoxide Generating Water Treatment Reactor.

    PubMed

    Medina, Victor F; Griggs, Chris S; Thomas, Catherine

    2016-06-01

    Cyanobacterial/Harmful Algal Blooms are a major issue for lakes and reservoirs throughout the U.S.A. An effective destructive technology could be useful to protect sensitive areas, such as areas near water intakes. The study presented in this article explored the use of a reactor called the KRIA Water Treatment System. The reactor focuses on the injection of superoxide (O2 (-)), which is generated electrochemically from the atmosphere, into the water body. In addition, the injection process generates a significant amount of cavitation. The treatment process was tested in 190-L reactors spiked with water from cyanobacterial contaminated lakes. The treatment was very effective at destroying the predominant species of cyanobacteria, Microcystis aeruginosa, organic matter, and decreasing chlorophyll concentration. Microcystin toxin concentrations were also reduced. Data suggest that cavitation alone was an effective treatment, but the addition of superoxide improved performance, particularly regarding removal of cyanobacteria and reduction of microcystin concentration. PMID:26846314

  16. Evaluation of the Destruction of the Harmful Cyanobacteria, Microcystis aeruginosa, with a Cavitation and Superoxide Generating Water Treatment Reactor.

    PubMed

    Medina, Victor F; Griggs, Chris S; Thomas, Catherine

    2016-06-01

    Cyanobacterial/Harmful Algal Blooms are a major issue for lakes and reservoirs throughout the U.S.A. An effective destructive technology could be useful to protect sensitive areas, such as areas near water intakes. The study presented in this article explored the use of a reactor called the KRIA Water Treatment System. The reactor focuses on the injection of superoxide (O2 (-)), which is generated electrochemically from the atmosphere, into the water body. In addition, the injection process generates a significant amount of cavitation. The treatment process was tested in 190-L reactors spiked with water from cyanobacterial contaminated lakes. The treatment was very effective at destroying the predominant species of cyanobacteria, Microcystis aeruginosa, organic matter, and decreasing chlorophyll concentration. Microcystin toxin concentrations were also reduced. Data suggest that cavitation alone was an effective treatment, but the addition of superoxide improved performance, particularly regarding removal of cyanobacteria and reduction of microcystin concentration.

  17. The mechanism of metals precipitation by biologically generated alkalinity in biofilm reactors.

    PubMed

    Remoudaki, E; Hatzikioseyian, A; Kousi, P; Tsezos, M

    2003-09-01

    Microbial strains, characterized by increased tolerance and ability to grow in metal bearing wastewaters as well as by effective metal sequestering capability by both active (bioaccumulative) and passive (biosorptive) processes, were tested as inoculum for metal laden wastewater treatment systems. Their capacity to grow in metal bearing wastewater, using an easily available and inexpensive carbon source such as acetate, was studied in batch experiments. Two principal conclusions were drawn: (1). Growth was observed for all the strains examined suggesting that the strains can be acclimated to metals bearing wastewaters. (2). Solution pH increased from neutral to alkaline values during growth (pH(initial)=7, pH(final)=10). The later was observed systematically for all strains. Metal precipitation, due to the metabolically generated alkalinity is expected as a result.Supporting evidence for this hypothesis was provided during the operation of two pilot moving bed sand filters treating two different metal bearing wastewaters. Acetate was used as carbon source to support the growth and maintenance of microbial biomass on the sand grains of the filters. The characteristics of the sludge produced from the operation of the pilot plants were subsequently studied in the laboratory. Both sludges were significantly loaded with large amounts of metals.A mechanism of metal precipitation induced by the metabolically generated alkalinity, when acetate is used as carbon source, could be proposed as the main process responsible for the metals sequestering inside the moving bed sandfilter reactor. PMID:12909102

  18. Technology Perspectives on the Management of Spent-Resin Wastes Generated From Nuclear Power Reactor Operations

    SciTech Connect

    Shiv Vijayan; Makoto Kikuchi; Akihiro Komatsu

    2002-07-01

    Organic-resin wastes (spent resins) are generated by different purification systems employed in all types of nuclear power reactors during routine and non-routine operations. The quantities of such resin wastes, and their inventories of contaminants vary depend on the operational goals of the individual power plant. Depending on the regulatory target in the particular jurisdiction where the reactor is located, the type and amounts of radionuclides, metals and other chemical contaminants in the resin waste determine the extent of treatment required for interim storage or final disposal of the waste. Resin-waste treatment comprises different operations such as pretreatment, conditioning/stabilization and containerisation that produce a waste package suitable for handling, transport, storage and disposal. One aspect of the contaminants that has significant impact on waste conditioning and the overall cost of managing such wastes are the concentrations of short half-life (arbitrarily less than approximately 30 years) radionuclides, and long half-life radionuclides, in particular carbon-14, and toxic metals present in the waste. A spectrum of resin-waste conditioning methods is available. Some methods have been applied to specific situations while others are being developed for future applications to meet the need for reducing worker dose, environmental releases, and waste-storage and disposal costs. This paper describes waste treatment options for low-level radioactive resin wastes and potential options of resin wastes containing appreciable amounts of carbon-14. Indications are that drying of the resin waste containing long half-life radionuclides such as carbon-14 and compaction or pelletizing can be favourable to allow interim dry-storage of the waste and to provide sufficient flexibility in the preparation of a suitable waste form to meet applicable waste acceptance criteria for the eventual disposal of such wastes. (authors)

  19. Post-scram Liquid Metal cooled Fast Breeder Reactor (LMFBR) neat transport system dynamics and steam generator control

    NASA Astrophysics Data System (ADS)

    Brukx, J. F. L. M.

    1982-06-01

    Loop type LMFBR heat transport system dynamics after reactor shutdown and during subsequent decay heat removal are considered with emphasis on steam generator dynamics including the development and evaluation of various post-scram steam generator control systems, and natural circulation of the sodium coolant, including the influence of superimposed free convection on forced convection heat transfer and pressure drop. The normal operating and decay heat removal functions of the overall heat transport system are described.

  20. Efficient generation of volatile cadmium species using Ti(III) and Ti(IV) and application to determination of cadmium by cold vapor generation inductively coupled plasma mass spectrometry (CVG-ICP-MS)†

    PubMed Central

    Arslan, Zikri; Yilmaz, Vedat; Rose, LaKeysha

    2015-01-01

    In this study, a highly efficient chemical vapor generation (CVG) approach is reported for determination of cadmium (Cd). Titanium (III) and titanium (IV) were investigated for the first time as catalytic additives along with thiourea, L-cysteine and potassium cyanide (KCN) for generation of volatile Cd species. Both Ti(III) and Ti(IV) provided the highest enhancement with KCN. The improvement with thiourea was marginal (ca. 2-fold), while L-cysteine enhanced signal slightly only with Ti(III) in H2SO4. Optimum CVG conditions were 4% (v/v) HCl + 0.03 M Ti(III) + 0.16 M KCN and 2% (v/v) HNO3 + 0.03 M Ti(IV) + 0.16 M KCN with a 3% (m/v) NaBH4 solution. The sensitivity was improved about 40-fold with Ti(III) and 35-fold with Ti(IV). A limit of detection (LOD) of 3.2 ng L−1 was achieved with Ti(III) by CVG-ICP-MS. The LOD with Ti(IV) was 6.4 ng L−1 which was limited by the blank signals in Ti(IV) solution. Experimental evidence indicated that Ti(III) and Ti(IV) enhanced Cd vapor generation catalytically; for best efficiency mixing prior to reaction with NaBH4 was critical. The method was highly robust against the effects of transition metal ions. No significant suppression was observed in the presence of Co(II), Cr(III), Cu(II), Fe(III), Mn(II), Ni(II) and Zn(II) up to 1.0 μg mL−1. Among the hydride forming elements, no interference was observed from As(III) and Se(IV) at 0.5 μg mL−1 level. The depressive effects from Pb(II) and Sb(III) were not significant at 0.1 μg mL−1 while those from Bi(III) and Sn(II) were marginal. The procedures were validated with determination of Cd by CVG-ICP-MS in a number certified reference materials, including Nearshore seawater (CASS-4), Bone ash (SRM 1400), Dogfish liver (DOLT-4), Mussel tissue (SRM 2976) and Domestic Sludge (SRM 2781). PMID:26251554

  1. CHLORINE ABSORPTION IN S(IV) SOLUTIONS

    EPA Science Inventory

    The report gives results of measurements of the rate of Chlorine (Cl2) absorption into aqueous sulfite/bisulfite -- S(IV) -- solutions at ambient temperature using a highly characterized stirred-cell reactor. The reactor media were 0 to 10 mM S(IV) with pHs of 3.5-8.5. Experiment...

  2. Small and medium-sized high-temperature reactors for generation of electricity, process steam and district heat

    SciTech Connect

    Schoening, J.

    1988-01-01

    The HTR reactor line of BBC/HRB has been designed to cover the requirements in the market of nuclear power energy of the time being and in the future. Cornerstones of the group's future HTR line are the HTR 500 (550 MWe) and the HTR 100 (100 MWe) for the generation of electricity and process steam, with the possibility of heat extraction for district heating. The HTR 500 design characteristics, reasons for choice of a 500 MW design, economics of the HTR 500, the HTR 100 design characteristics, process heat application, and small heating reactors are discussed in the paper.

  3. Development of new generation reduced activation ferritic-martenstic steels for advanced fusion reactors

    DOE PAGES

    Tan, Lizhen; Snead, Lance Lewis; Katoh, Yutai

    2016-05-26

    International development of reduced activation ferritic-martensitic (RAFM) steels has focused on 9 wt percentage Cr, which primarily contain M23C6 (M = Cr-rich) and small amounts of MX (M = Ta/V, X = C/N) precipitates, not adequate to maintain strength and creep resistance above ~500 °C. To enable applications at higher temperatures for better thermal efficiency of fusion reactors, computational alloy thermodynamics coupled with strength modeling have been employed to explore a new generation RAFM steels. The new alloys are designed to significantly increase the amount of MX nanoprecipitates, which are manufacturable through standard and scalable industrial steelmaking methods. Preliminary experimentalmore » results of the developed new alloys demonstrated noticeably increased amount of MX, favoring significantly improved strength, creep resistance, and Charpy impact toughness as compared to current RAFM steels. Furthermore, the strength and creep resistance were comparable or approaching to the lower bound of, but impact toughness was noticeably superior to 9–20Cr oxide dispersion-strengthened ferritic alloys.« less

  4. Development of new generation reduced activation ferritic-martensitic steels for advanced fusion reactors

    NASA Astrophysics Data System (ADS)

    Tan, L.; Snead, L. L.; Katoh, Y.

    2016-09-01

    International development of reduced activation ferritic-martensitic (RAFM) steels has focused on 9 wt percentage Cr, which primarily contain M23C6 (M = Cr-rich) and small amounts of MX (M = Ta/V, X = C/N) precipitates, not adequate to maintain strength and creep resistance above ∼500 °C. To enable applications at higher temperatures for better thermal efficiency of fusion reactors, computational alloy thermodynamics coupled with strength modeling have been employed to explore a new generation RAFM steels. The new alloys are designed to significantly increase the amount of MX nanoprecipitates, which are manufacturable through standard and scalable industrial steelmaking methods. Preliminary experimental results of the developed new alloys demonstrated noticeably increased amount of MX, favoring significantly improved strength, creep resistance, and Charpy impact toughness as compared to current RAFM steels. The strength and creep resistance were comparable or approaching to the lower bound of, but impact toughness was noticeably superior to 9-20Cr oxide dispersion-strengthened ferritic alloys.

  5. Sodium fast reactor evaluation: Core materials

    NASA Astrophysics Data System (ADS)

    Cheon, Jin Sik; Lee, Chan Bock; Lee, Byoung Oon; Raison, J. P.; Mizuno, T.; Delage, F.; Carmack, J.

    2009-07-01

    In the framework of the Generation IV Sodium Fast Reactor (SFR) Program the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. In this paper the status of available and developmental materials for SFR core cladding and duct applications is reviewed. To satisfy the Generation IV SFR fuel requirements, an advanced cladding needs to be developed. The candidate cladding materials are austenitic steels, ferritic/martensitic (F/M) steels, and oxide dispersion strengthened (ODS) steels. A large amount of irradiation testing is required, and the compatibility of cladding with TRU-loaded fuel at high temperatures and high burnup must be investigated. The more promising F/M steels (compared to HT9) might be able to meet the dose requirements of over 200 dpa for ducts in the GEN-IV SFR systems.

  6. Technology Options for a Fast Spectrum Test Reactor

    SciTech Connect

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

    2006-06-01

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

  7. Code System to Process WIMSD4 Interface Output Files and Generate Two-Group Data for Reactor Calculations.

    1992-12-03

    Version 00 The code processes the WIMS-D/4 binary output files for producing two-group microscopic cross sections and macroscopic lattice cell constants (zone and cell macroscopic cross sections, D, M, and K-infinity) in a more flexible format needed for reactor burnup codes like CITATION, for reactor dynamics codes like NADYP-W and for other reactor codes. The purpose of the WIMSCORE-ENEA code is to facilitate the automation of data transfer between the cell calculation code WIMS andmore » the diffusion-burnup codes. Use is made of the VARY storage manipulation package. WIMSCORE generates output files to be used by the codes TDB, TRITON, CITATION.« less

  8. Design of the Next Generation Nuclear Plant Graphite Creep Experiments for Irradiation in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2009-05-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain seven separate stacks of graphite specimens. Six of the specimen stacks will have half of their graphite specimens under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will be organized into pairs with a different compressive load being applied to the top half of each pair of specimen stacks. The seventh stack will not have a compressive load on the graphite specimens during irradiation. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any

  9. Steam Generator Component Model in a Combined Cycle of Power Conversion Unit for Very High Temperature Gas-Cooled Reactor

    SciTech Connect

    Oh, Chang H; Han, James; Barner, Robert; Sherman, Steven R

    2007-06-01

    The Department of Energy and the Idaho National Laboratory are developing a Next Generation Nuclear Plant (NGNP), Very High Temperature Gas-Cooled Reactor (VHTR) to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is two fold 1) efficient low cost energy generation and 2) hydrogen production. Although a next generation plant could be developed as a single-purpose facility, early designs are expected to be dual-purpose. While hydrogen production and advanced energy cycles are still in its early stages of development, research towards coupling a high temperature reactor, electrical generation and hydrogen production is under way. A combined cycle is considered as one of the power conversion units to be coupled to the very high-temperature gas-cooled reactor (VHTR). The combined cycle configuration consists of a Brayton top cycle coupled to a Rankine bottoming cycle by means of a steam generator. A detailed sizing and pressure drop model of a steam generator is not available in the HYSYS processes code. Therefore a four region model was developed for implementation into HYSYS. The focus of this study was the validation of a HYSYS steam generator model of two phase flow correlations. The correlations calculated the size and heat exchange of the steam generator. To assess the model, those calculations were input into a RELAP5 model and its results were compared with HYSYS results. The comparison showed many differences in parameters such as the heat transfer coefficients and revealed the different methods used by the codes. Despite differences in approach, the overall results of heat transfer were in good agreement.

  10. Fuel development for gas-cooled fast reactors

    NASA Astrophysics Data System (ADS)

    Meyer, M. K.; Fielding, R.; Gan, J.

    2007-09-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High-Temperature Reactor (VHTR), as well as actinide burning concepts [A Technology Roadmap for Generation IV Nuclear Energy Systems, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the US and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic 'honeycomb' structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  11. FORIG: a computer code for calculating radionuclide generation and depletion in fusion and fission reactors. User's manual

    SciTech Connect

    Blink, J.A.

    1985-03-01

    In this manual we describe the use of the FORIG computer code to solve isotope-generation and depletion problems in fusion and fission reactors. FORIG runs on a Cray-1 computer and accepts more extensive activation cross sections than ORIGEN2 from which it was adapted. This report is an updated and a combined version of the previous ORIGEN2 and FORIG manuals. 7 refs., 15 figs., 13 tabs.

  12. Dynamic Strain Aging in New Generation Cr-Mo-V Steel for Reactor Pressure Vessel Applications

    NASA Astrophysics Data System (ADS)

    Gupta, C.; Chakravartty, J. K.; Banerjee, S.

    2010-12-01

    A new generation nuclear reactor pressure vessel steel (CrMoV type) having compositional similarities with thick section 3Cr-Mo class of low alloy steels and adapted for nuclear applications was investigated for various manifestations of dynamic strain aging (DSA) using uniaxial tests. The steel investigated herein has undergone quenched and tempered treatment such that a tempered bainite microstructure with Cr-rich carbides was formed. The scope of the uniaxial experiments included tensile tests over a temperature range of 298 K to 873 K (25 °C to 600 °C) at two strain rates (10-3 and 10-4 s-1), as well as suitably designed transient strain rate change tests. The flow behavior displayed serrated flow, negative strain rate sensitivity, plateau behavior of yield, negative temperature ( T), and strain rate left( {dot{\\varepsilon }} right) dependence of flow stress over the temperature range of 523 K to 673 K (250 °C to 400 °C) and strain rate range of 5 × 10-3 s-1 to 3 × 10-6 s-1, respectively. While these trends attested to the presence of DSA, a lack of work hardening and near negligible impairment of ductility point to the fact that manifestations of embrittling features of DSA were significantly enervated in the new generation pressure vessel steel. In order to provide a mechanistic understanding of these unique combinations of manifestations of DSA in the steel, a new approach for evaluation of responsible solutes from strain rate change tests was adopted. From these experiments and calculation of activation energy by application of vacancy-based models, the solutes responsible for DSA were identified as carbon/nitrogen. The lack of embrittling features of DSA in the steel was rationalized as being due to the beneficial effects arising from the presence of dynamic recovery effects, presence of alloy carbides in the tempered bainitic structure, and formation of solute clusters, all of which hinder the possibilities for strong aging of dislocations.

  13. Accelerated development of Zr-containing new generation ferritic steels for advanced nuclear reactors

    SciTech Connect

    Tan, Lizhen; Yang, Ying; Sridharan, K.

    2015-12-01

    The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as the sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computational tools) is an important path to more efficient alloy development and process optimization. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of Zr-bearing ferritic alloys that can be fabricated using conventional steelmaking methods. The new alloys are expected to have superior high-temperature creep performance and excellent radiation resistance as compared to Grade 91. The designed alloys were fabricated using arc-melting and drop-casting, followed by hot rolling and conventional heat treatments. Comprehensive experimental studies have been conducted on the developed alloys to evaluate their hardness, tensile properties, creep resistance, Charpy impact toughness, and aging resistance, as well as resistance to proton and heavy ion (Fe2+) irradiation.

  14. Design Options to Reduce Development Cost of First Generation Surface Reactors

    SciTech Connect

    Poston, David I.; Marcille, Thomas F.

    2006-01-20

    Low-power surface reactors have the potential to have the lowest development cost of any space reactor application, primarily because system alpha (mass/kg) is not of utmost importance and mission lifetimes do not have to be a decade or more. Even then, the development cost of a surface reactor can vary substantially depending on the performance requirements (e.g. mass, power, lifetime, reliability) and technical development risk deemed acceptable by the end-user. It is important for potential users to be aware of these relationships before they determine their future architecture (i.e. decide what they need). Generally, the greatest potential costs of a space reactor program are a nuclear-powered ground test and extensive material development campaigns, so it is important to consider options that can minimize the need for or complexity of such tasks. The intended goal of this paper is to inform potential surface reactor users of the potential sensitivities of surface reactor development cost to design requirements, and areas where technical risk can be traded with development cost.

  15. Batch slurry photocatalytic reactors for the generation of hydrogen from sulfide and sulfite waste streams under solar irradiation

    SciTech Connect

    Priya, R.; Kanmani, S.

    2009-10-15

    In this study, two solar slurry photocatalytic reactors i.e., batch reactor (BR) and batch recycle reactor with continuous supply of inert gas (BRRwCG) were developed for comparing their performance. The performance of the photocatalytic reactors were evaluated based on the generation of hydrogen (H{sub 2}) from water containing sodium sulfide (Na{sub 2}S) and sodium sulfite (Na{sub 2}SO{sub 3}) ions. The photoreactor of capacity 300 mL was developed with UV-vis transparent walls. The catalytic powders ((CdS/ZnS)/Ag{sub 2}S + (RuO{sub 2}/TiO{sub 2})) were kept suspended by means of magnetic stirrer in the BR and gas bubbling and recycling of the suspension in the BRRwCG. The rate constant was found to be 120.86 (einstein{sup -1}) for the BRRwCG whereas, for the BR it was found to be only 10.92 (einstein{sup -1}). The higher rate constant was due to the fast desorption of products and suppression of e{sup -}/h{sup +} recombination. (author)

  16. An experimental study of catalytic and non-catalytic reaction in heat recirculating reactors and applications to power generation

    NASA Astrophysics Data System (ADS)

    Ahn, Jeongmin

    An experimental study of the performance of a Swiss roll heat exchanger and reactor was conducted, with emphasis on the extinction limits and comparison of results with and without Pt catalyst. At Re<40, the catalyst was required to sustain reaction; with the catalyst self-sustaining reaction could be obtained at Re less than 1. Both lean and rich extinction limits were extended with the catalyst, though rich limits were extended much further. At low Re, the lean extinction limit was rich of stoichiometric and rich limit had equivalence ratios 80 in some cases. Non-catalytic reaction generally occurred in a flameless mode near the center of the reactor. With or without catalyst, for sufficiently robust conditions, a visible flame would propagate out of the center, but this flame could only be re-centered with catalyst. Gas chromatography indicated that at low Re, CO and non-C3 H8 hydrocarbons did not form. For higher Re, catalytic limits were slightly broader but had much lower limit temperatures. At sufficiently high Re, catalytic and gas-phase limits merged. Experiments with titanium Swiss rolls have demonstrated reducing wall thermal conductivity and thickness leads to lower heat losses and therefore increases operating temperatures and extends flammability limits. By use of Pt catalysts, reaction of propane-air mixtures at temperatures 54°C was sustained. Such low temperatures suggest that polymers may be employed as a reactor material. A polyimide reactor was built and survived prolonged testing at temperatures up to 500°C. Polymer reactors may prove more practical for microscale devices due to their lower thermal conductivity and ease of manufacturing. Since the ultimate goal of current efforts is to develop combustion driven power generation devices at MEMS like scales, a thermally self-sustaining miniature power generation device was developed utilizing a single-chamber solid-oxide-fuel-cell (SOFC) placed in a Swiss roll. With the single-chamber design

  17. Properties of bio-oil generated by a pyrolysis of forest cedar residuals with the movable Auger-type reactor

    NASA Astrophysics Data System (ADS)

    Nishimura, Shun; Miyazato, Akio; Ebitani, Kohki

    2016-02-01

    Our research project has developed the new movable reactor for bio-oil production in 2013 on the basis of Auger-type system. This package would be a great impact due to the concept of local production for local consumption in the hilly and mountainous area in not only Japan but also in the world. Herein, we would like to report the properties of the bio-oil generated by the developing Auger-type movable reactor. The synthesized bio-oil possessed C: 46.2 wt%, H: 6.5 wt%, N: wt%, S: <0.1 wt%, O: 46.8 wt% and H2O: 18.4 wt%, and served a good calorific value of 18.1 MJ/kg. The spectroscopic and mass analyses such as FT-IR, GC-MS, 13C-NMR and FT-ICR MS supported that the bio-oil was composed by the fine mixtures of methoxy phenols and variety of alcohol or carboxylic acid functional groups. Thus, it is suggested that the bio-oil generated by the new movable Auger-type reactor has a significant potential as well as the existing bio-oil reported previously.

  18. Core Design Issues of the Supercritcal Water Fast Reactor

    NASA Astrophysics Data System (ADS)

    Mori, Magnus; Rineiski, Andrei; Maschek, Werner; Sinitsa, Valentin

    2006-04-01

    The Super Critical water Fast Reactor is a Generation IV reactor concept, which presents new and challenging design issues. A correct estimation of the void effect for this water-cooled pressurized system is of fundamental importance to assess its theoretical feasibility. Hence, in this work an overview of the void effect analysis is shown together with the resulting core design issues. The effect of the application of different cross section libraries and models on the core design is also treated.

  19. ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN THE R AND P REACTOR VESSELS

    SciTech Connect

    Wiersma, B.

    2010-05-24

    operations in the R-reactor vessel is low for the Portland cement. Alternatively, if the grout fill rate is less than 0.5 inch/min and the grout is maintained at a temperature of 80 C, the risk is again low. Although these calculations are conservative, there are some measures that may be taken to further minimize the potential for hydrogen evolution. (1) Minimize the temperature of the grout as much as practical. Lower temperatures will mean lower hydrogen generation rates. For P-reactor, grout temperatures less than 100 C should provide an adequate safety margin for the pH 8 and pH 10.4 grout formulations. For R-reactor, grout temperatures less than 70 C or 80 C will provide an adequate safety margin for the Portland cement. The other grout formulations are also viable options for R-reactor. (2) Minimize the grout fill rate as much as practical. Lowering the fill rate takes advantage of passivation of the aluminum components and hence lower hydrogen generation rates. For P-reactor, fill rates that are less than 2 inches/min for the ceramicrete and the silica fume grouts will reduce the chance of significant hydrogen accumulation. For R-reactor, fill rates less than 1 inch/min will again minimize the risk of hydrogen accumulation. (3) Ventilate the building as much as practical (e.g., leave doors open) to further disperse hydrogen. The volumetric hydrogen generation rates in the P-reactor vessel, however, are low for the pH 8 and pH 10.4 grout, (i.e., less than 0.97 ft{sup 3}/min). If further walk-down inspections of the reactor vessels suggest an increase in the actual areal density of aluminum, the calculations should be re-visited.

  20. Generation IV Nuclear Energy Systems Construction Cost Reductions through the use of Virtual Environments: Task 1 Completion Report

    SciTech Connect

    Whisker, V.E.; Baratta, A.J.; Shaw, T.S.; Winters, J.W.; Trikouros, N.; Hess, C.

    2002-11-26

    OAK B204 The objective of this project is to demonstrate the feasibility and effectiveness of using full-scale virtual reality simulation in the design, construction, and maintenance of future nuclear power plants. Specifically, this project will test the suitability of Immersive Projection Display (IPD) technology to aid engineers in the design of the next generation nuclear power plant and to evaluate potential cost reductions that can be realized by optimization of installation and construction sequences. The intent is to see if this type of information technology can be used in capacities similar to those currently filled by full-scale physical mockups.

  1. Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression

    DOEpatents

    Lasche, George P.

    1988-01-01

    A high-power-density laser or charged-particle-beam fusion reactor system maximizes the directed kinetic energy imparted to a large mass of liquid lithium by a centrally located fusion target. A fusion target is embedded in a large mass of lithium, of sufficient radius to act as a tritium breeding blanket, and provided with ports for the access of beam energy to implode the target. The directed kinetic energy is converted directly to electricity with high efficiency by work done against a pulsed magnetic field applied exterior to the lithium. Because the system maximizes the blanket thickness per unit volume of lithium, neutron-induced radioactivities in the reaction chamber wall are several orders of magnitude less than is typical of other fusion reactor systems.

  2. Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression

    DOEpatents

    Lasche, G.P.

    1987-02-20

    A high-power-density-laser or charged-particle-beam fusion reactor system maximizes the directed kinetic energy imparted to a large mass of liquid lithium by a centrally located fusion target. A fusion target is embedded in a large mass of lithium, of sufficient radius to act as a tritium breeding blanket, and provided with ports for the access of beam energy to implode the target. The directed kinetic energy is converted directly to electricity with high efficiency by work done against a pulsed magnetic field applied exterior to the lithium. Because the system maximizes the blanket thickness per unit volume of lithium, neutron-induced radioactivities in the reaction chamber wall are several orders of magnitude less than is typical of other fusion reactor systems. 25 figs.

  3. Reactor Physics Parametric and Depletion Studies in Support of TRISO Particle Fuel Specification for the Next Generation Nuclear Plant

    SciTech Connect

    James W. Sterbentz; Bren Phillips; Robert L. Sant; Gray S. Chang; Paul D. Bayless

    2003-09-01

    Reactor physics calculations were initiated to answer several major questions related to the proposed TRISO-coated particle fuel that is to be used in the prismatic Very High Temperature Reactor (VHTR) or the Next Generation Nuclear Plant (NGNP). These preliminary design evaluation calculations help ensure that the upcoming fuel irradiation tests will test appropriate size and type of fuel particles for a future NGNP reactor design. Conclusions from these calculations are expected to confirm and suggest possible modifications to the current particle fuel parameters specified in the evolving Fuel Specification. Calculated results dispel the need for a binary fuel particle system, which is proposed in the General Atomics GT-MHR concept. The GT-MHR binary system is composed of both a fissile and fertile particle with 350- and 500- micron kernel diameters, respectively. For the NGNP reactor, a single fissile particle system (single UCO kernel size) can meet the reactivity and power cycle length requirements demanded of the NGNP. At the same time, it will provide substantial programmatic cost savings by eliminating the need for dual particle fabrication process lines and dual fuel particle irradiation tests required of a binary system. Use of a larger 425-micron kernel diameter single fissile particle (proposed here), as opposed to the 350-micron GT-MHR fissile particle size, helps alleviate current compact particle packing fractions fabrication limitations (<35%), improves fuel block loading for higher n-batch reload options, and tracks the historical correlation between particle size and enrichment (10 and 14 wt% U-235 particle enrichments are proposed for the NGNP). Overall, the use of the slightly larger kernel significantly broadens the NGNP reactor core design envelope and provides increased design margin to accommodate the (as yet) unknown final NGNP reactor design. Maximum power-peaking factors are calculated for both the initial and equilibrium NGNP cores

  4. ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN C-REACTOR DISASSEMBLY BASIN

    SciTech Connect

    Wiersma, B.

    2011-07-12

    C-reactor disassembly basin is being prepared for deactivation and decommissioning (D and D). D and D activities will consist primarily of immobilizing contaminated scrap components and structures in a grout-like formulation. The disassembly basin will be the first area of the C-reactor building that will be immobilized. The scrap components contain aluminum alloy materials. Any aluminum will corrode very rapidly when it comes in contact with the very alkaline grout (pH > 13), and as a result would produce hydrogen gas. To address this potential deflagration/explosion hazard, Savannah River National Laboratory (SRNL) reviewed and evaluated existing experimental and analytical studies of this issue to determine if any process constraints are necessary. The risk of accumulation of a flammable mixture of hydrogen above the surface of the water during the injection of grout into the C-reactor disassembly area is low if the assessment of the aluminum surface area is reliable. Conservative calculations estimate that there is insufficient aluminum present in the basin areas to result in significant hydrogen accumulation in this local region. The minimum safety margin (or factor) on a 60% LFL criterion for a local region of the basin (i.e., Horizontal Tube Storage) was greater than 3. Calculations also demonstrated that a flammable situation in the vapor space above the basin is unlikely. Although these calculations are conservative, there are some measures that may be taken to further minimize the risk of developing a flammable condition during grouting operations.

  5. Determination of the Sensitivity of the Antineutrino Probe for Reactor Core Monitoring

    SciTech Connect

    Cormon, S.; Fallot, M. Bui, V.-M.; Cucoanes, A.; Estienne, M.; Lenoir, M.; Onillon, A.; Shiba, T.; Yermia, F.; Zakari-Issoufou, A.-A.

    2014-06-15

    This paper presents a feasibility study of the use of the detection of reactor-antineutrinos (ν{sup ¯}{sub e}) for non proliferation purpose. To proceed, we have started to study different reactor designs with our simulation tools. We use a package called MCNP Utility for Reactor Evolution (MURE), initially developed by CNRS/IN2P3 labs to study Generation IV reactors. The MURE package has been coupled to fission product beta decay nuclear databases for studying reactor antineutrino emission. This method is the only one able to predict the antineutrino emission from future reactor cores, which don't use the thermal fission of {sup 235}U, {sup 239}Pu and {sup 241}Pu. It is also the only way to include off-equilibrium effects, due to neutron captures and time evolution of the fission product concentrations during a reactor cycle. We will present here the first predictions of antineutrino energy spectra from innovative reactor designs (Generation IV reactors). We will then discuss a summary of our results of non-proliferation scenarios involving the latter reactor designs, taking into account reactor physics constraints.

  6. Determination of the Sensitivity of the Antineutrino Probe for Reactor Core Monitoring

    NASA Astrophysics Data System (ADS)

    Cormon, S.; Fallot, M.; Bui, V.-M.; Cucoanes, A.; Estienne, M.; Lenoir, M.; Onillon, A.; Shiba, T.; Yermia, F.; Zakari-Issoufou, A.-A.

    2014-06-01

    This paper presents a feasibility study of the use of the detection of reactor-antineutrinos (νbare) for non proliferation purpose. To proceed, we have started to study different reactor designs with our simulation tools. We use a package called MCNP Utility for Reactor Evolution (MURE), initially developed by CNRS/IN2P3 labs to study Generation IV reactors. The MURE package has been coupled to fission product beta decay nuclear databases for studying reactor antineutrino emission. This method is the only one able to predict the antineutrino emission from future reactor cores, which don't use the thermal fission of 235U, 239Pu and 241Pu. It is also the only way to include off-equilibrium effects, due to neutron captures and time evolution of the fission product concentrations during a reactor cycle. We will present here the first predictions of antineutrino energy spectra from innovative reactor designs (Generation IV reactors). We will then discuss a summary of our results of non-proliferation scenarios involving the latter reactor designs, taking into account reactor physics constraints.

  7. Evaluation of anticipatory signal to steam generator pressure control program for 700 MWe Indian pressurized heavy water reactor

    SciTech Connect

    Pahari, S.; Hajela, S.; Rammohan, H. P.; Malhotra, P. K.; Ghadge, S. G.

    2012-07-01

    700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) is horizontal channel type reactor with partial boiling at channel outlet. Due to boiling, it has a large volume of vapor present in the primary loops. It has two primary loops connected with the help of pressurizer surge line. The pressurizer has a large capacity and is partly filled by liquid and partly by vapor. Large vapor volume improves compressibility of the system. During turbine trip or load rejection, pressure builds up in Steam Generator (SG). This leads to pressurization of Primary Heat Transport System (PHTS). To control pressurization of SG and PHTS, around 70% of the steam generated in SG is dumped into the condenser by opening Condenser Steam Dump Valves (CSDVs) and rest of the steam is released to the atmosphere by opening Atmospheric Steam Discharge Valves (ASDVs) immediately after sensing the event. This is accomplished by adding anticipatory signal to the output of SG pressure controller. Anticipatory signal is proportional to the thermal power of reactor and the proportionality constant is set so that SG pressure controller's output jacks up to ASDV opening range when operating at 100% FP. To simulate this behavior for 700 MWe IPHWR, Primary and secondary heat transport system is modeled. SG pressure control and other process control program have also been modeled to capture overall plant dynamics. Analysis has been carried out with 3-D neutron kinetics coupled thermal hydraulic computer code ATMIKA.T to evaluate the effect of the anticipatory signal on PHT pressure and over all plant dynamics during turbine trip in 700 MWe IPHWR. This paper brings out the results of the analysis with and without considering anticipatory signal in SG pressure control program during turbine trip. (authors)

  8. Current generation by helicons and lower hybrid waves in modern tokamaks and reactors ITER and DEMO. Scenarios, modeling and antennae

    SciTech Connect

    Vdovin, V. L.

    2013-02-15

    The innovative concept and 3D full-wave code modeling the off-axis current drive by radio-frequency (RF) waves in large-scale tokamaks, ITER and DEMO, for steady-state operation with high efficiency is proposed. The scheme uses the helicon radiation (fast magnetosonic waves at high (20-40) ion cyclotron frequency harmonics) at frequencies of 500-700 MHz propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by helicons, in conjunction with the bootstrap current, ensure the maintenance of a given value of the total current in the stability margin q(0) {>=} 2 and q(a) {>=} 4, and will help to have regimes with a negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure {beta}{sub N} > 3 (the so-called advanced scenarios) of interest for the commercial reactor. Modeling with full-wave three-dimensional codes PSTELION and STELEC showed flexible control of the current profile in the reactor plasmas of ITER and DEMO, using multiple frequencies, the positions of the antennae and toroidal wave slow down. Also presented are the results of simulations of current generation by helicons in the DIII-D, T-15MD, and JT-60AS tokamaks. Commercially available continuous-wave klystrons of the MW/tube range are promising for commercial stationary fusion reactors. The compact antennae of the waveguide type are proposed, and an example of a possible RF system for today's tokamaks is given. The advantages of the scheme (partially tested at lower frequencies in tokamaks) are a significant decline in the role of parametric instabilities in the plasma periphery, the use of electrically strong resonator-waveguide type antennae, and substantially greater antenna-plasma coupling.

  9. New generation of data acquisition and data storage systems of the IBR-2 reactor spectrometers complex

    NASA Astrophysics Data System (ADS)

    Kulikov, S. A.; Prikhodko, V. I.

    2016-07-01

    The paper presents an overview of works on the creation of data acquisition and data storage systems, which have been carried out in the Department of the IBR-2 spectrometers complex (DCS) of the Frank Laboratory of Neutron Physics (FLNP) over the past 15 years (before, during, and after the modernization of the IBR-2 reactor). These systems represent a unified set of identical (from the viewpoint of hardware) modules limited in type but functionally complete, wherein distinctions in parameters, functional capabilities, encoding, correction and preliminary data processing procedures specific to each spectrometer are realized on the level of microprograms, electronic tables, and integrated software control system.

  10. Welding IV.

    ERIC Educational Resources Information Center

    Allegheny County Community Coll., Pittsburgh, PA.

    Instructional objectives and performance requirements are outlined in this course guide for Welding IV, a competency-based course in advanced arc welding offered at the Community College of Allegheny County to provide students with proficiency in: (1) single vee groove welding using code specifications established by the American Welding Society…

  11. Spent caustic oxidation using electro-generated Fenton's reagent in a batch reactor.

    PubMed

    Rodriguez, Nicolas; Hansen, Henrik K; Nunez, Patricio; Guzman, Jaime

    2008-07-01

    This work shows the results of four Electro-Fenton laboratory tests to reduce the chemical oxygen demand (COD) in spent caustic solutions. The treatment consisted of (i) a pH reduction followed by (ii) an Electro-Fenton process, which was analyzed in this work. The Fenton's reagent was produced in a specially designed reactor, where the waste stream flowed through a labyrinth made by ferrous plates. These plates acted as sacrificial anodes-releasing Fe(2 +) cations to the solution, where H(2)O(2) was also added. The Electro-Fenton process was analyzed varying the ferrous ion concentration ([Fe(+ 2)]), the spent caustic's initial temperature and the initial pH. Close to 95% removal of COD (from 8800 mg L(- 1)) was achieved at a pH of 4, a temperature of 40 degrees C and 100 mg L(- 1) of Fe(+ 2) (applying 1 A). Two models were considered to simulate the behavior of the reactor considering (i) axial dispersion and (ii) kinetic rate, respectively. The model that was based on kinetics, proved to be the slightly closest fit to the experimental values. PMID:18569308

  12. Core thermal response and hydrogen generation of the N Reactor hydrogen mitigation design basis accident

    SciTech Connect

    White, M.D.; Lombardo, N.J.; Heard, F.J.; Ogden, D.M.; Quapp, W.J.

    1988-04-01

    Calculations were performed to determine core heatup, core damage, and subsequent hydrogen production of a hypothetical loss-of-cooling accident at the Department of Energy's N Reactor. The thermal transient response of the reactor core was solved using the TRUMP-BD computer program. Estimates of whole-core thermal damage and hydrogen production were made by weighting the results of multiple half-length pressure tube simulations at various power levels. The Baker-Just and Wilson parabolic rate equations for the metal-water chemical reactions modeled the key phenomena of chemical energy and hydrogen evolution. Unlimited steam was assumed available for continuous oxidation of exposed Zircaloy-2 surfaces and for uranium metal with fuel cladding beyond the failure temperature (1038 C). Intact fuel geometry was modeled. Maximum fuel temperatures (1181 C) in the cooled central regions of the core were predicted to occur one-half hour into the accident scenario. Maximum fuel temperatures of 1447 C occurred in the core GSCS-regions at the end of the 10-h transient. After 10-h 26% of the fuel inventory was predicted to have failed. Peak hydrogen evolution equaled 42 g/s, while 10-h integrated hydrogen evolution equaled 167 kg. 12 refs., 12 figs., 2 tabs.

  13. High Temperature Gas-Cooled Reactors Lessons Learned Applicable to the Next Generation Nuclear Plant

    SciTech Connect

    J. M. Beck; L. F. Pincock

    2011-04-01

    The purpose of this report is to identify possible issues highlighted by these lessons learned that could apply to the NGNP in reducing technical risks commensurate with the current phase of design. Some of the lessons learned have been applied to the NGNP and documented in the Preconceptual Design Report. These are addressed in the background section of this document and include, for example, the decision to use TRISO fuel rather than BISO fuel used in the Peach Bottom reactor; the use of a reactor pressure vessel rather than prestressed concrete found in Fort St. Vrain; and the use of helium as a primary coolant rather than CO2. Other lessons learned, 68 in total, are documented in Sections 2 through 6 and will be applied, as appropriate, in advancing phases of design. The lessons learned are derived from both negative and positive outcomes from prior HTGR experiences. Lessons learned are grouped according to the plant, areas, systems, subsystems, and components defined in the NGNP Preconceptual Design Report, and subsequent NGNP project documents.

  14. Designing a Gas Test Loop for the Advanced Test Reactor

    SciTech Connect

    James R. Parry

    2005-11-01

    The Generation IV Reactor Program and the Advanced Fuel Cycle Initiative are investigating some new reactor concepts which require extensive materials and fuels testing in a fast neutron spectrum. The capability to test materials and fuels in a fast neutron flux in the United States is very limited to non-existent. It has been proposed to install a gas test loop (GTL) in one of the lobes of the Advanced Test Reactor (ATR) at the Idaho National Laboratory and harden the spectrum to provide some fast neutron flux testing capabilities in the United States. This paper describes the neutronics investigation into the design of the GTL for the ATR.

  15. Neutron cross-sections for next generation reactors: new data from n_TOF.

    PubMed

    Colonna, N; Abbondanno, U; Aerts, G; Alvarez, H; Alvarez-Velarde, F; Andriamonje, S; Andrzejewski, J; Assimakopoulos, P; Audouin, L; Badurek, G; Baumann, P; Becvar, F; Berthoumieux, E; Calviani, M; Calviño, F; Cano-Ott, D; Capote, R; de Albornoz, A Carrillo; Cennini, P; Chepel, V; Chiaveri, E; Cortes, G; Couture, A; Cox, J; Dahlfors, M; David, S; Dillman, I; Dolfini, R; Domingo-Pardo, C; Dridi, W; Duran, I; Eleftheriadis, C; Ferrant, L; Ferrari, A; Ferreira-Marques, R; Frais-Koelbl, H; Fujii, K; Furman, W; Goncalves, I; González-Romero, E; Goverdovski, A; Gramegna, F; Griesmayer, E; Guerrero, C; Gunsing, F; Haas, B; Haight, R; Heil, M; Herrera-Martinez, A; Igashira, M; Isaev, S; Jericha, E; Käppeler, F; Kadi, Y; Karadimos, D; Karamanis, D; Kerveno, M; Ketlerov, V; Koehler, P; Konovalov, V; Kossionides, E; Krticka, M; Lampoudis, C; Leeb, H; Lindote, A; Lopes, I; Lozano, M; Lukic, S; Marganiec, J; Marques, L; Marrone, S; Martínez, T; Massimi, C; Mastinu, P; Mengoni, A; Milazzo, P M; Moreau, C; Mosconi, M; Neves, F; Oberhummer, H; O'Brien, S; Oshima, M; Pancin, J; Papachristodoulou, C; Papadopoulos, C; Paradela, C; Patronis, N; Pavlik, A; Pavlopoulos, P; Perrot, L; Pigni, M T; Plag, R; Plompen, A; Plukis, A; Poch, A; Pretel, C; Quesada, J; Rauscher, T; Reifarth, R; Rosetti, M; Rubbia, C; Rudolf, G; Rullhusen, P; Salgado, J; Sarchiapone, L; Savvidis, I; Stephan, C; Tagliente, G; Tain, J L; Tassan-Got, L; Tavora, L; Terlizzi, R; Vannini, G; Vaz, P; Ventura, A; Villamarin, D; Vicente, M C; Vlachoudis, V; Vlastou, R; Voss, F; Walter, S; Wendler, H; Wiescher, M; Wisshak, K

    2010-01-01

    In 2002, an innovative neutron time-of-flight facility started operation at CERN: n_TOF. The main characteristics that make the new facility unique are the high instantaneous neutron flux, high resolution and wide energy range. Combined with state-of-the-art detectors and data acquisition system, these features have allowed to collect high accuracy neutron cross-section data on a variety of isotopes, many of which radioactive, of interest for Nuclear Astrophysics and for applications to advanced reactor technologies. A review of the most important results on capture and fission reactions obtained so far at n_TOF is presented, together with plans for new measurements related to nuclear industry.

  16. Neutron cross-sections for next generation reactors: new data from n_TOF.

    PubMed

    Colonna, N; Abbondanno, U; Aerts, G; Alvarez, H; Alvarez-Velarde, F; Andriamonje, S; Andrzejewski, J; Assimakopoulos, P; Audouin, L; Badurek, G; Baumann, P; Becvar, F; Berthoumieux, E; Calviani, M; Calviño, F; Cano-Ott, D; Capote, R; de Albornoz, A Carrillo; Cennini, P; Chepel, V; Chiaveri, E; Cortes, G; Couture, A; Cox, J; Dahlfors, M; David, S; Dillman, I; Dolfini, R; Domingo-Pardo, C; Dridi, W; Duran, I; Eleftheriadis, C; Ferrant, L; Ferrari, A; Ferreira-Marques, R; Frais-Koelbl, H; Fujii, K; Furman, W; Goncalves, I; González-Romero, E; Goverdovski, A; Gramegna, F; Griesmayer, E; Guerrero, C; Gunsing, F; Haas, B; Haight, R; Heil, M; Herrera-Martinez, A; Igashira, M; Isaev, S; Jericha, E; Käppeler, F; Kadi, Y; Karadimos, D; Karamanis, D; Kerveno, M; Ketlerov, V; Koehler, P; Konovalov, V; Kossionides, E; Krticka, M; Lampoudis, C; Leeb, H; Lindote, A; Lopes, I; Lozano, M; Lukic, S; Marganiec, J; Marques, L; Marrone, S; Martínez, T; Massimi, C; Mastinu, P; Mengoni, A; Milazzo, P M; Moreau, C; Mosconi, M; Neves, F; Oberhummer, H; O'Brien, S; Oshima, M; Pancin, J; Papachristodoulou, C; Papadopoulos, C; Paradela, C; Patronis, N; Pavlik, A; Pavlopoulos, P; Perrot, L; Pigni, M T; Plag, R; Plompen, A; Plukis, A; Poch, A; Pretel, C; Quesada, J; Rauscher, T; Reifarth, R; Rosetti, M; Rubbia, C; Rudolf, G; Rullhusen, P; Salgado, J; Sarchiapone, L; Savvidis, I; Stephan, C; Tagliente, G; Tain, J L; Tassan-Got, L; Tavora, L; Terlizzi, R; Vannini, G; Vaz, P; Ventura, A; Villamarin, D; Vicente, M C; Vlachoudis, V; Vlastou, R; Voss, F; Walter, S; Wendler, H; Wiescher, M; Wisshak, K

    2010-01-01

    In 2002, an innovative neutron time-of-flight facility started operation at CERN: n_TOF. The main characteristics that make the new facility unique are the high instantaneous neutron flux, high resolution and wide energy range. Combined with state-of-the-art detectors and data acquisition system, these features have allowed to collect high accuracy neutron cross-section data on a variety of isotopes, many of which radioactive, of interest for Nuclear Astrophysics and for applications to advanced reactor technologies. A review of the most important results on capture and fission reactions obtained so far at n_TOF is presented, together with plans for new measurements related to nuclear industry. PMID:20096595

  17. Decommissioning of the secondary containment of the steam generating heavy water reactor at UKAEA-Winfrith

    SciTech Connect

    Miller, Keith; Cornell, Rowland; Parkinson, Steve; McIntyre, Kevin; Staples, Andy

    2007-07-01

    Available in abstract form only. Full text of publication follows: The Winfrith SGHWR was a prototype nuclear power plant operated for 23 years by the United Kingdom Atomic Energy Authority (UKAEA) until 1990 when it was shut down permanently. The current Stage 1 decommissioning contract is part of a multi-stage strategy. It involves the removal of all the ancillary plant and equipment in the secondary containment and non-containment areas ahead of a series of contracts for the decommissioning of the primary containment, the reactor core and demolition of the building and all remaining facilities. As an outcome of a competitive tending process, the Stage 1 decommissioning contract was awarded to NUKEM with operations commencing in April 2005. The decommissioning processes involved with these plant items will be described with some emphasis of the establishment of multiple work-fronts for the production, recovery, treatment and disposal of mainly tritium-contaminated waste arising from its contact with the direct cycle reactor coolant. The means of size reduction of a variety of large, heavy and complex items of plant made from a range of materials will also be described with some emphasis on the control of fumes during hot cutting operations and establishing effective containments within a larger secondary containment structure. Disposal of these wastes in a timely and cost-effective manner is a major challenge facing the decommissioning team and has required the development of a highly efficient means of packing the resultant materials into mainly one-third height ISO containers for disposal as LLW. Details of the quantities of LLW and exempt wastes handled during this process will be given with a commentary about the difficulty in segregating these two waste streams efficiently. (authors)

  18. IVS Organization

    NASA Technical Reports Server (NTRS)

    2004-01-01

    International VLBI Service (IVS) is an international collaboration of organizations which operate or support Very Long Baseline Interferometry (VLBI) components. The goals are: To provide a service to support geodetic, geophysical and astrometric research and operational activities. To promote research and development activities in all aspects of the geodetic and astrometric VLBI technique. To interact with the community of users of VLBI products and to integrate VLBI into a global Earth observing system.

  19. I-V and DLTS study of generation and annihilation of deep-level defects in an oxygen-ion irradiated bipolar junction transistor

    NASA Astrophysics Data System (ADS)

    Madhu, K. V.; Kulkarni, S. R.; Ravindra, M.; Damle, R.

    A commercial bipolar junction transistor (2N 2219A, npn) irradiated with 84 MeV O6+-ions with fluence of the order of 1013 ions cm-2 is studied for radiation-induced gain degradation and deep-level defects or recombination centers. I-V measurements are made to study the gain degradation as a function of ion fluence. Properties such as activation energy, trap concentration and capture cross section of deep levels are studied by deep-level transient spectroscopy. Minority carrier trap energy levels with energies ranging from EC -0.17 eV to EC -0.55 eV are observed in the base-collector junction of the transistor. Majority carrier defect levels are also observed with energies ranging from EV +0.26 eV to EV +0.44 eV. The irradiated device is subjected to isothermal and isochronal annealing. The defects are seen to anneal above 250 °C. The defects generated in the base region of the transistor by displacement damage appear to be responsible for an increase in base current through Shockley-Read-Hall or multi-phonon recombination and consequent transistor gain degradation.

  20. SPECIATION OF SELENIUM AND ARSENIC COMPOUNDS BY CAPILLARY ELECTROPHORESIS WITH HYDRODYNAMICALLY MODIFIED ELECTROOSMOTIC FLOW AND ON-LINE REDUCTION OF SELENIUM(VI) TO SELENIUM(IV) WITH HYDRIDE GENERATION INDUCTIVELY COUPLED PLASMA MASS SPECTROMETRIC DETECTION

    EPA Science Inventory

    Capillary electrophoresis (CE) with hydride generation inductively coupled plasma mass spectrometry was used to determine four arsenicals and two selenium species. Selenate (SeVI) was reduced on-line to selenite (SeIV') by mixing the CE effluent with concentrated HCl. A microporo...

  1. Model biogas steam reforming in a thin Pd-supported membrane reactor to generate clean hydrogen for fuel cells

    NASA Astrophysics Data System (ADS)

    Iulianelli, A.; Liguori, S.; Huang, Y.; Basile, A.

    2015-01-01

    Steam reforming of a model biogas mixture is studied for generating clean hydrogen by using an inorganic membrane reactor, in which a composite Pd/Al2O3 membrane separates part of the produced hydrogen through its selective permeation. The characteristics of H2 perm-selectivity of the fresh membrane is expressed in terms of H2/N2 ideal selectivity, in this case equal to 4300. Concerning biogas steam reforming reaction, at 380 °C, 2.0 bar H2O:CH4 = 3:1, GHSV = 9000 h-1 the permeate purity of the recovered hydrogen is around 96%, although the conversion (15%) and hydrogen recovery (>20%) are relatively low; on the contrary, at 450 °C, 3.5 bar H2O:CH4 = 4:1, GHSV = 11000 h-1 the conversion is increased up to more than 30% and the recovery of hydrogen to about 70%. This novel work constitutes a reference study for new developments on biogas steam reforming reaction in membrane reactors.

  2. Improvement of Analysis Technology for High Temperature Gas-Cooled Reactor by Using Data Obtained in High Temperature Engineering Test Reactor

    NASA Astrophysics Data System (ADS)

    Nakagawa, Shigeaki; Tochio, Daisuke; Takamatsu, Kuniyoshi; Goto, Minoru; Takeda, Tetsuaki

    The Very High Temperature Reactor (VHTR) system, which is one of generation IV reactors, is the high temperature gas-cooled reactor (HTGR) with capabilities of hydrogen production and high efficiency electricity generation. The High Temperature Engineering Test Reactor (HTTR) is the first HTGR in Japan. The HTTR achieved full power of 30MW at a reactor outlet coolant temperature of about 950°C in April, 2004 during the “rise-to-power tests” confirming the reactor performance. The safety demonstration tests by using the HTTR started from 2002 and are under going to demonstrate inherent safety features of HTGRs. The experimental data obtained in these tests are inevitable to design the VHTR with high cost performance. The analytical models validated through these tests in the HTTR are applicable to precise simulation of an HTGR performance and can contribute to the research and development of the VHTR.

  3. Decommissioning of the secondary containment of the steam generating heavy water reactor at UKAEA Winfrith

    SciTech Connect

    Miller, K.D.; Cornell, R.M.; Parkinson, S.J.; McIntyre, K.; Staples, A.

    2007-07-01

    The Winfrith SGHWR was a prototype nuclear power plant operated for 23 years by the United Kingdom Atomic Energy Authority (UKAEA) until 1990 when it was shut down permanently. The current Stage 1 decommissioning contract is part of a multi-stage strategy. It involves the removal of all the ancillary plant and equipment in the secondary containment and non-containment areas ahead of a series of contracts for the decommissioning of the primary containment, the reactor core and demolition of the building and ail remaining facilities. As an outcome of a competitive tending process, the Stage 1 decommissioning contract was awarded to NUKEM with operations commencing in April 2005. The decommissioning processes involved with these plant items will be described with some emphasis of the establishment of multiple work-fronts for the production, recovery, treatment and disposal of mainly tritium-contaminated waste arising from its contact with the direct cycle reactor coolant. The means of size reduction of a variety of large, heavy and complex items of plant made from a range of materials will also be described with some emphasis on the control of fumes during hot cutting operations and establishing effective containments within a larger secondary containment structure. Disposal of these wastes in a timely and cost-effective manner is a major challenge facing the decommissioning team and has required the development of a highly efficient means of packing the resultant materials into mainly one-third height IS0 containers for disposal as LLW. Details of the quantities of LLW and exempt wastes handled during this process will be given with a commentary about the difficulty in segregating these two waste streams efficiently. The paper sets out to demonstrate the considerable progress that has been made with these challenging decommissioning operations at the SGHWR plant and to highlight some of the techniques and processes that have contributed to the overall success of the

  4. Superiority of second over first generation chemotherapy in a randomized trial for stage III-IV intermediate and high-grade non-Hodgkin's lymphoma (NHL): the 1980-1985 EORTC trial. The EORTC Lymphoma Group.

    PubMed

    Carde, P; Meerwaldt, J H; van Glabbeke, M; Somers, R; Monconduit, M; Thomas, J; de Wolf-Peeters, C; de Pauw, B; Tanguy, A; Kluin-Nelemans, J C

    1991-06-01

    A first-generation CHOP-like cyclic combination chemotherapy (CT) regimen using cyclophosphamide 600 mg/m2 IV d1, hydroxorubicin (doxorubicin) 50 mg/m2 IV d1, VM26 60 mg/m2 IV d1, and prednisone 40 mg/m2 PO d1-5 (CHVmP) was compared to a second-generation combination wherein vincristine 1.4 mg/m2 IV and bleomycin 6 mg/m2 IM/IV were added at mid-interval (d15) to the former drugs (CHVmP + VB) in the treatment of intermediate- and high-grade malignant NHL. From April 1980 to January 1986, 141 eligible patients with stage III-IV unfavorable histologies (except T lymphoblastic NHL) entered this EORTC randomized trial. In both arms adjuvant radiotherapy (30 Gy) was given in instances of bulky or residual disease. In all patient subsets the outcome favored the second-generation regimen. The difference was even greater in patients with Diffuse Large Cell Lymphoma (DLCL). At 5 years, overall survival was 53% with CHVmP + VB versus 29% (p = 0.002). The advantage was due to a higher complete remission (CR) rate (80% versus 50%, p = 0.01). Indeed, once CR was achieved the relapse-free survival (RFS) was not significantly influenced (59% versus 49%). No significant additional toxicity could be attributed to vincristine and bleomycin. This study demonstrates a clear benefit for intermediate- and high-risk malignant NHL and particularly DLCL from intercalating non-myelotoxic drugs at mid-cycle intervals, without adverse effects. PMID:1722697

  5. Superiority of second over first generation chemotherapy in a randomized trial for stage III-IV intermediate and high-grade non-Hodgkin's lymphoma (NHL): the 1980-1985 EORTC trial. The EORTC Lymphoma Group.

    PubMed

    Carde, P; Meerwaldt, J H; van Glabbeke, M; Somers, R; Monconduit, M; Thomas, J; de Wolf-Peeters, C; de Pauw, B; Tanguy, A; Kluin-Nelemans, J C

    1991-06-01

    A first-generation CHOP-like cyclic combination chemotherapy (CT) regimen using cyclophosphamide 600 mg/m2 IV d1, hydroxorubicin (doxorubicin) 50 mg/m2 IV d1, VM26 60 mg/m2 IV d1, and prednisone 40 mg/m2 PO d1-5 (CHVmP) was compared to a second-generation combination wherein vincristine 1.4 mg/m2 IV and bleomycin 6 mg/m2 IM/IV were added at mid-interval (d15) to the former drugs (CHVmP + VB) in the treatment of intermediate- and high-grade malignant NHL. From April 1980 to January 1986, 141 eligible patients with stage III-IV unfavorable histologies (except T lymphoblastic NHL) entered this EORTC randomized trial. In both arms adjuvant radiotherapy (30 Gy) was given in instances of bulky or residual disease. In all patient subsets the outcome favored the second-generation regimen. The difference was even greater in patients with Diffuse Large Cell Lymphoma (DLCL). At 5 years, overall survival was 53% with CHVmP + VB versus 29% (p = 0.002). The advantage was due to a higher complete remission (CR) rate (80% versus 50%, p = 0.01). Indeed, once CR was achieved the relapse-free survival (RFS) was not significantly influenced (59% versus 49%). No significant additional toxicity could be attributed to vincristine and bleomycin. This study demonstrates a clear benefit for intermediate- and high-risk malignant NHL and particularly DLCL from intercalating non-myelotoxic drugs at mid-cycle intervals, without adverse effects.

  6. Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part II: Prismatic Reactor Cross Section Generation

    SciTech Connect

    Vincent Descotes

    2011-03-01

    The deep-burn prismatic high temperature reactor is made up of an annular core loaded with transuranic isotopes and surrounded in the center and in the periphery by reflector blocks in graphite. This disposition creates challenges for the neutronics compared to usual light water reactor calculation schemes. The longer mean free path of neutrons in graphite affects the neutron spectrum deep inside the blocks located next to the reflector. The neutron thermalisation in the graphite leads to two characteristic fission peaks at the inner and outer interfaces as a result of the increased thermal flux seen in those assemblies. Spectral changes are seen at least on half of the fuel blocks adjacent to the reflector. This spectral effect of the reflector may prevent us from successfully using the two step scheme -lattice then core calculation- typically used for light water reactors. We have been studying the core without control mechanisms to provide input for the development of a complete calculation scheme. To correct the spectrum at the lattice level, we have tried to generate cross-sections from supercell calculations at the lattice level, thus taking into account part of the graphite surrounding the blocks of interest for generating the homogenised cross-sections for the full-core calculation. This one has been done with 2 to 295 groups to assess if increasing the number of groups leads to more accurate results. A comparison with a classical single block model has been done. Both paths were compared to a reference calculation done with MCNP. It is concluded that the agreement with MCNP is better with supercells, but that the single block model remains quite close if enough groups are kept for the core calculation. 26 groups seems to be a good compromise between time and accu- racy. However, some trials with depletion have shown huge variations of the isotopic composition across a block next to the reflector. It may imply that at least an in- core depletion for the

  7. Transport of breeder reactor-fire-generated sodium oxide aerosols for building-wake-dominated meteorology

    SciTech Connect

    Fields, D.E.; Cooper, A.C.; Miller, C.W.

    1987-02-01

    This report describes the methodology used and results obtained in efforts to estimate the sodium aerosol concentrations at air intake ports of a liquid-metal cooled, fast-breeder nuclear reactor. An earlier version of this methodology has been previously discussed (Fields and Miller, 1985). A range of wind speeds from 2 to 10 m/s is assumed, and an effort is made to include building wake effects which, in many cases, dominate the dispersal of aerosols near buildings. For relatively small release rates, on the order of 1 to 10 kg/s, the plume rise is small and estimates of aerosol concentrations are derived using the methodology of Wilson and Britter (1982), which describes releases from surface vents. For release rates on the order of 100 kg/s much higher release velocities are expected, and plume rise is considered. An effective increase in release height is computed using the Split-H methodology with a parameterization suggested by Ramsdell (1983), and the release source strength is transformed to rooftop level. Evaluation of the acute release aerosol concentration is then based on the methodology for releases from a surface release of this transformed source strength. For a horizontal release, a methodology is developed to chart the plume path as a function of release and site meteorology parameters. Results described herein must be regarded as maximum aerosol concentrations, based on models derived from generic wind tunnel studies. More accurate and site-specific results may be obtained through wind tunnel simulations and through simulating emissions from release points other than those assumed here.

  8. The Role of Expansion and Fragmentation Phenomena on the Generation and Chemical Composition of Dust Particles in a Flash Converting Reactor

    NASA Astrophysics Data System (ADS)

    Duarte-Ruiz, Cirilo Andrés; Pérez-Tello, Manuel; Parra-Sánchez, Víctor Roberto; Sohn, Hong Yong

    2016-07-01

    A compositional fragmentation model was used to clarify the effect of expansion and fragmentation phenomena on the generation and chemical composition of dust particles in a flash converting reactor. A fragmentation index is introduced to represent the fraction of particles undergoing fragmentation, as opposed to expansion, within the particle population. Under typical operating conditions, the local dust content and the net amount of dust generated compared with the dust content in the feed first decreased and then increased along the reactor length, whereas the amount of particles undergoing fragmentation (fragmentation index) increased steadily. Dust generation was found to be the result of two competing phenomena, i.e., the expansion of dust particles in the feed and the production of dust from fragmentation of large particles. At short distance from the burner tip, the dust mostly consists of particles in the feed undergoing oxidation and expansion, whereas farther down the reactor it mostly consists of fragments of partially reacted particles. Based on the computer simulations under a variety of experimental conditions, a map of dust generation against fragmentation index was developed. For most practical purposes, dust generation may be approximated by the change in the mass fraction of dust in the population. At the reactor exit, the composition of the dust is approximately the same as the entire particle population.

  9. The Role of Expansion and Fragmentation Phenomena on the Generation and Chemical Composition of Dust Particles in a Flash Converting Reactor

    NASA Astrophysics Data System (ADS)

    Duarte-Ruiz, Cirilo Andrés; Pérez-Tello, Manuel; Parra-Sánchez, Víctor Roberto; Sohn, Hong Yong

    2016-10-01

    A compositional fragmentation model was used to clarify the effect of expansion and fragmentation phenomena on the generation and chemical composition of dust particles in a flash converting reactor. A fragmentation index is introduced to represent the fraction of particles undergoing fragmentation, as opposed to expansion, within the particle population. Under typical operating conditions, the local dust content and the net amount of dust generated compared with the dust content in the feed first decreased and then increased along the reactor length, whereas the amount of particles undergoing fragmentation (fragmentation index) increased steadily. Dust generation was found to be the result of two competing phenomena, i.e., the expansion of dust particles in the feed and the production of dust from fragmentation of large particles. At short distance from the burner tip, the dust mostly consists of particles in the feed undergoing oxidation and expansion, whereas farther down the reactor it mostly consists of fragments of partially reacted particles. Based on the computer simulations under a variety of experimental conditions, a map of dust generation against fragmentation index was developed. For most practical purposes, dust generation may be approximated by the change in the mass fraction of dust in the population. At the reactor exit, the composition of the dust is approximately the same as the entire particle population.

  10. A rule-based expert system for automatic control rod pattern generation for boiling water reactors

    SciTech Connect

    Lin, L.S.; Lin, C. )

    1991-07-01

    This paper reports on an expert system for generating control rod patterns that has been developed. The knowledge is transformed into IF-THEN rules. The inference engine uses the Rete pattern matching algorithm to match facts, and rule premises and conflict resolution strategies to make the system function intelligently. A forward-chaining mechanism is adopted in the inference engine. The system is implemented in the Common Lisp programming language. The three-dimensional core simulation model performs the core status and burnup calculations. The system is successfully demonstrated by generating control rod programming for the 2894-MW (thermal) Kuosheng nuclear power plant in Taiwan. The computing time is tremendously reduced compared to programs using mathematical methods.

  11. Segmented flow generation by chip reactors for highly parallelized cell cultivation.

    PubMed

    Grodrian, Andreas; Metze, Josef; Henkel, Thomas; Martin, Karin; Roth, Martin; Köhler, J Michael

    2004-06-15

    Micro system technology offers convenient tools for the production of handling devices for small liquid volumes which can be used in cell cultivation. Here, a modular system for the rapid generation of cell suspension aliquots is presented. The system is used to produce and analyze high numbers of well-separated culture volumes. Selected clones may be retrieved from the system. Therefore, the principle of segmented flow is applied. Portions of aqueous culture medium containing one cell or very small cell ensembles are separated from each other by a nonmiscible liquid like dodecane, tetradecane or mineral oil. In addition, the alkane separates the culture droplets from the innerside of the walls of chip channels and capillaries. This way, compatibility problems between cell wall surfaces and the chemical character of walls are excluded. The separated culture droplets are guided by micro flow transportation in different channel and chamber topologies. The whole system has the character of a serially operating cell processing system. The aliquot generation can be sped up to frequencies of about 30 Hz in each microchannel. That means, that about 10(5) individual cultural volumes can be produced per hour or about 2 million per day. The survival and the growth of microorganisms has been shown for model organisms as well as for organisms from a natural sample (soil). PMID:15093213

  12. Segmented flow generation by chip reactors for highly parallelized cell cultivation

    NASA Astrophysics Data System (ADS)

    Grodrian, A.; Metze, J.; Henkel, Thomas; Roth, M.; Kohler, Johann M.

    2002-11-01

    Micro system technology offers convenient tools for the production of handling devices for small liquid volumes which can be used in cell cultivation. Here, a modular system for the rapid generation of cell suspension aliquotes is presented. The system is used to produce and analyze high numbers of strongly separated cultural volumes. Selected clones may be retrieved from the system. Therefore, the principle of segmented flow is applies. Portions of aqueous culture medium containing one cell or very small cell ensembles are separated from each other by a nonmiscible liquid like dodecane or mineral oil. In addition, the oil separates the cultivation droplets from the innerside of the walls of chip channels and capillaries. This way, compatibility problems between cell wall surfaces and the chemical character of technical walls are excluded. The separated cultivation droplets are guided by micro flow transportation in different channel and chamber topologies. The whole system has the character of a serially working cell processing system. The aliquot generation can be speeded up to frequencies of about 30 Hz in each micro channel. That means, that about 105 individual cultural volumes can be produced per hour or about 2 million per day.

  13. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production

    SciTech Connect

    Philip MacDonald; Jacopo Buongiorno; James Sterbentz; Cliff Davis; Robert Witt; Gary Was; J. McKinley; S. Teysseyre; Luca Oriani; Vefa Kucukboyaci; Lawrence Conway; N. Jonsson: Bin Liu

    2005-02-13

    The supercritical water reactor (SCWR) has been the object of interest throughout the nuclear Generation IV community because of its high potential: a simple, direct cycle, compact configuration; elimination of many traditional LWR components, operation at coolant temperatures much higher than traditional LWRs and thus high thermal efficiency. It could be said that the SWR was viewed as the water counterpart to the high temperature gas reactor.

  14. Chemistry and technology of Molten Salt Reactors - history and perspectives

    NASA Astrophysics Data System (ADS)

    Uhlíř, Jan

    2007-01-01

    Molten Salt Reactors represent one of promising future nuclear reactor concept included also in the Generation IV reactors family. This reactor type is distinguished by an extraordinarily close connection between the reactor physics and chemical technology, which is given by the specific features of the chemical form of fuel, representing by molten fluoride salt and circulating through the reactor core and also by the requirements of continuous 'on-line' reprocessing of the spent fuel. The history of Molten Salt Reactors reaches the period of fifties and sixties, when the first experimental Molten Salt Reactors were constructed and tested in ORNL (US). Several molten salt techniques dedicated to fresh molten salt fuel processing and spent fuel reprocessing were studied and developed in those days. Today, after nearly thirty years of discontinuance, a renewed interest in the Molten Salt Reactor technology is observed. Current experimental R&D activities in the area of Molten Salt Reactor technology are realized by a relatively small number of research institutions mainly in the EU, Russia and USA. The main effort is directed primarily to the development of separation processes suitable for the molten salt fuel processing and reprocessing technology. The techniques under development are molten salt/liquid metal extraction processes, electrochemical separation processes from the molten salt media, fused salt volatilization techniques and gas extraction from the molten salt medium.

  15. The modular helium reactor for future energy needs

    SciTech Connect

    Richards, Matthew B.; Shenoy, Arkal S.; Venneri, Francesco; LaBar, Malcolm P.; Schultz, Kenneth R.; Brown, Lloyd C.

    2006-07-01

    The Modular Helium Reactor (MHR) is one of the advanced reactor concepts within the internationally-supported Generation IV program. Because of its design features and design maturity, the MHR was selected by the U.S. Department of Energy (DOE) as the U.S. Generation IV design concept for the Next Generation Nuclear Plant (NGNP). Other countries, including Russia, Japan, South Korea, China, South Africa, and France are also developing this technology, and large-scale deployment of MHR technology is a realistic element of future energy-growth scenarios. In this paper, we discuss MHR conceptual designs for electricity and hydrogen production and their role in a sustainable energy future with significant growth in nuclear energy. For hydrogen production, two conceptual designs are described; one coupling the MHR to thermochemical water splitting using the sulfur-iodine (SI) process and the other coupling the MHR to high-temperature electrolysis (HTE). (authors)

  16. Association of rat thoracic aorta dilatation by astragaloside IV with the generation of endothelium-derived hyperpolarizing factors and nitric oxide, and the blockade of Ca2+ channels

    PubMed Central

    HU, GUANYING; LI, XIXIONG; ZHANG, SANYIN; WANG, XIN

    2016-01-01

    The aim of the present study was to elucidate the roles of endothelium-derived hyperpolarizing factors (EDHFs) and nitric oxide (NO) in mediating the vasodilatation response to astragaloside IV and the effects of astragaloside IV on voltage-dependent Ca2+ channels and receptor-operated Ca2+ channels in rat thoracic aortic rings precontracted with potassium chloride (KCl; 60 mM) or phenylephrine (PHE; 1 µM). The results showed that astragaloside IV (1×10−4-3×10−1 g/l) concentration-dependently relaxed the contraction induced by KCl (10–90 mM) or PHE (1×10−9-3×10−5 µM) and inhibited concentration-contraction curves for the two vasoconstrictors in the aortic rings. Preincubation with Nω-nitro-L-arginine methyl ester (L-NAME, 100 µM) significantly attenuated astragaloside IV-induced relaxation in the endothelium-intact and -denuded arterial rings precontracted with PHE. Astragaloside IV, following preincubation with L-NAME (100 µM) plus indomethacin (10 µM), exerted vasodilatation, which was depressed by tetraethtylamine (1 mM) and propargylglycine (100 µM), but not by carbenoxolone (10 µM), catalase (500 U/ml) or proadifen hydrochloride (10 µM). The action mode of astragaloside IV was evident in comparison to nifedipine. Inhibition of PHE-induced contraction by astragaloside IV (100 mg/l) was more potent compared to inhibition of KCl-induced contraction, while inhibition of KCl-induced contraction by nifedipine (100 mg/l) was more potent compared to inhibition of PHE-induced contraction by nifedipine (100 mg/l). In addition, the combination of astragaloside IV and nifedipine exhibited synergistic and additive inhibitory effects on contraction evoked by KCl, which was similar to PHE. In conclusion, astragaloside IV, as a Ca2+ antagonist, relaxes the vessels through the blockade of superior receptor-operated Ca2+ and inferior voltage-dependent Ca2+ channels, which modulate NO from vascular endothelial cells and vascular smooth muscle cells, and

  17. Reactor Controllability of 3-Region-Core Molten Salt Reactor System - A Study on Load Following Capability

    SciTech Connect

    Takahisa Yamamoto; Koshi Mitachi; Masatoshi Nishio

    2006-07-01

    The Molten Salt Reactor (MSR) systems are liquid-fueled reactors that can be used for actinide burning, production of electricity, production of hydrogen, and production of fissile fuels (breeding). Thorium (Th) and uranium-233 ({sup 233}U) are fertile and fissile of the MSR systems, and dissolved in a high-temperature molten fluoride salt (fuel salt) with a very high boiling temperature (up to 1650 K), that is both the reactor nuclear fuel and the coolant. The MSR system is one of the six advanced reactor concepts identified by the Generation IV International Forum (GIF) as a candidate for cooperative development. In the MSR system, fuel salt flows through a fuel duct constructed around a reactor core and fuel channel of a graphite moderator accompanied by fission reaction and heat generation, and flows out to an external-loop system consisted of a heat exchanger and a circulation pump. Due to the motion of fuel salt, delayed neutron precursors that are one of the source of neutron production make to change their position between the fission reaction and neutron emission events and decay even occur in the external loop system. Hence the reactivity and effective delayed neutron precursor fraction of the MSR system are lower than those of solid fuel reactor systems such as Boiling Water Reactors (BWRs) and Pressurised Water Reactor (PWRs). Since all of the presently operating nuclear power reactors utilize solid fuel, little attention had been paid to the MSR analysis of the reactivity loss and reactor characteristics change caused by the fuel salt circulation. Sides et al. and Shimazu et al. developed MSR analytical models based on the point reactor kinetics model to consider the effect of fuel salt flow. Their models represented a reactor as having six zones for fuel salt and three zones for the graphite moderator. Since their models employed the point reactor kinetics model and the rough temperature approximation, their results were not sufficiently accurate to

  18. Operating experience feedback report -- turbine-generator overspeed protection systems: Commercial power reactors. Volume 11

    SciTech Connect

    Ornstein, H.L.

    1995-04-01

    This report presents the results of the US Nuclear Regulatory Commission`s Office for Analysis and Evaluation of Operational Data (AEOD) review of operating experience of main turbine-generator overspeed and overspeed protection systems. It includes an indepth examination of the turbine overspeed event which occurred on November 9, 1991, at the Salem Unit 2 Nuclear Power Plant. It also provides information concerning actions taken by other utilities and the turbine manufacturers as a result of the Salem overspeed event. AEOD`s study reviewed operating procedures and plant practices. It noted differences between turbine manufacturer designs and recommendations for operations, maintenance, and testing, and also identified significant variations in the manner that individual plants maintain and test their turbine overspeed protection systems. AEOD`s study provides insight into the shortcomings in the design, operation, maintenance, testing, and human factors associated with turbine overspeed protection systems. Operating experience indicates that the frequency of turbine overspeed events is higher than previously thought and that the bases for demonstrating compliance with NRC`s General Design Criterion (GDC) 4, Environmental and dynamic effects design bases, may be nonconservative with respect to the assumed frequency.

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

    SciTech Connect

    Chandler, David

    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

  20. Neutron behavior, reactor control, and reactor heat transfer. Volume four

    SciTech Connect

    Not Available

    1986-01-01

    Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant).

  1. Asteroids IV

    NASA Astrophysics Data System (ADS)

    Michel, Patrick; DeMeo, Francesca E.; Bottke, William F.

    . Asteroids, like planets, are driven by a great variety of both dynamical and physical mechanisms. In fact, images sent back by space missions show a collection of small worlds whose characteristics seem designed to overthrow our preconceived notions. Given their wide range of sizes and surface compositions, it is clear that many formed in very different places and at different times within the solar nebula. These characteristics make them an exciting challenge for researchers who crave complex problems. The return of samples from these bodies may ultimately be needed to provide us with solutions. In the book Asteroids IV, the editors and authors have taken major strides in the long journey toward a much deeper understanding of our fascinating planetary ancestors. This book reviews major advances in 43 chapters that have been written and reviewed by a team of more than 200 international authorities in asteroids. It is aimed to be as comprehensive as possible while also remaining accessible to students and researchers who are interested in learning about these small but nonetheless important worlds. We hope this volume will serve as a leading reference on the topic of asteroids for the decade to come. We are deeply indebted to the many authors and referees for their tremendous efforts in helping us create Asteroids IV. We also thank the members of the Asteroids IV scientific organizing committee for helping us shape the structure and content of the book. The conference associated with the book, "Asteroids Comets Meteors 2014" held June 30-July 4, 2014, in Helsinki, Finland, did an outstanding job of demonstrating how much progress we have made in the field over the last decade. We are extremely grateful to our host Karri Muinonnen and his team. The editors are also grateful to the Asteroids IV production staff, namely Renée Dotson and her colleagues at the Lunar and Planetary Institute, for their efforts, their invaluable assistance, and their enthusiasm; they made life as

  2. Asteroids IV

    NASA Astrophysics Data System (ADS)

    Michel, Patrick; DeMeo, Francesca E.; Bottke, William F.

    . Asteroids, like planets, are driven by a great variety of both dynamical and physical mechanisms. In fact, images sent back by space missions show a collection of small worlds whose characteristics seem designed to overthrow our preconceived notions. Given their wide range of sizes and surface compositions, it is clear that many formed in very different places and at different times within the solar nebula. These characteristics make them an exciting challenge for researchers who crave complex problems. The return of samples from these bodies may ultimately be needed to provide us with solutions. In the book Asteroids IV, the editors and authors have taken major strides in the long journey toward a much deeper understanding of our fascinating planetary ancestors. This book reviews major advances in 43 chapters that have been written and reviewed by a team of more than 200 international authorities in asteroids. It is aimed to be as comprehensive as possible while also remaining accessible to students and researchers who are interested in learning about these small but nonetheless important worlds. We hope this volume will serve as a leading reference on the topic of asteroids for the decade to come. We are deeply indebted to the many authors and referees for their tremendous efforts in helping us create Asteroids IV. We also thank the members of the Asteroids IV scientific organizing committee for helping us shape the structure and content of the book. The conference associated with the book, "Asteroids Comets Meteors 2014" held June 30-July 4, 2014, in Helsinki, Finland, did an outstanding job of demonstrating how much progress we have made in the field over the last decade. We are extremely grateful to our host Karri Muinonnen and his team. The editors are also grateful to the Asteroids IV production staff, namely Renée Dotson and her colleagues at the Lunar and Planetary Institute, for their efforts, their invaluable assistance, and their enthusiasm; they made life as

  3. Evaluation of cracking in feedwater piping adjacent to the steam generators in Nine Pressurized Water Reactor Plants

    SciTech Connect

    Goldberg, A.; Streit, R.D.; Scott, R.G.

    1980-06-25

    Cracking in ASTM A106-B and A106-C feedwater piping was detected near the inlet to the steam generators in a number of pressurized water reactor plants. We received sections with cracks from nine of the plants with the objective of identifying the cracking mechanism and assessing various factors that might contribute to this cracking. Variations were observed in piping surface irregularities, corrosion-product, pit, and crack morphology, surface elmental and crystal structure analyses, and steel microstructures and mechanical properties. However, with but two exceptions, namely, arrest bands and major surface irregularities, we were unable to relate the extent of cracking to any of these factors. Tensile and fracture toughness (J/sub Ic/ and tearing modulus) properties were measured over a range of temperatures and strain rates. No unusual properties or microstructures were observed that could be related to the cracking problem. All crack surfaces contained thick oxide deposits and showed evidence of cyclic events in the form of arrest bands. Transmission electron microscopy revealed fatigue striations on replicas of cleaned crack surfaces from one plant and possibly from three others. Calculations based on the observed striation spacings gave a value of ..delta..sigma = 150 MPa (22 ksi) for one of the major cracks. The direction of crack propagation was invariably related to the piping surface and not to the piping axis. These two factors are consistent with the proposed concept of thermally induced, cyclic, tensile surface stresses. Although surface irregularities and corrosion pits were sources for crack initiation and corrosion may have contributed to crack propagation, it is proposed that the overriding factor in the cracking problem is the presence of unforeseen cyclic loads.

  4. Status of materials handbooks for particle accelerator and nuclear reactor applications

    NASA Astrophysics Data System (ADS)

    Maloy, Stuart; Rogers, Berylene; Ren, Weiju; Rittenhouse, Philip

    2008-06-01

    In support of research and development for accelerator applications, a materials handbook was developed in August of 1998 funded by the Accelerator Production of Tritium Project. This handbook, presently called Advanced Fuel Cycle Initiative ( AFCI) Materials Handbook, Materials Data for Particle Accelerator Applications, has just issued Revision 5 and contains detailed information showing the effects of irradiation on many properties for a wide variety of materials. Development of a web-accessible materials database for Generation IV Reactor Programs has been ongoing for about three years. This handbook provides a single authoritative source for qualified materials data applicable to all Generation IV reactor concepts. A beta version of this Gen IV Materials Handbook has been completed and is presently under evaluation.

  5. The dependence of helium generation rate on nickel content of Fe-Cr-Ni alloys irradiated at high dpa levels in fast reactors

    SciTech Connect

    Garner, F.A.; Oliver, B.M.; Greenwood, L.R.

    1997-04-01

    With a few exceptions in the literature, it is generally accepted that it is nickel in Fe-Cr-Ni alloys that produces most of the transmutant helium and that the helium generation rate should scale linearly with the nickel content. Surprisingly, this assumption is based only on irradiations of pure nickel and has never been tested in an alloy series. There have also been no extensive tests of the predictions for helium production in alloys in various fast reactors spectra.

  6. Next Generation Nuclear Plant Project Technology Development Roadmaps: The Technical Path Forward for 750–800°C Reactor Outlet Temperature

    SciTech Connect

    John Collins

    2009-08-01

    This document presents the NGNP Critical PASSCs and defines their technical maturation path through Technology Development Roadmaps (TDRMs) and their associated Technology Readiness Levels (TRLs). As the critical PASSCs advance through increasing levels of technical maturity, project risk is reduced and the likelihood of within-budget and on-schedule completion is enhanced. The current supplier-generated TRLs and TDRMs for a 750–800°C reactor outlet temperature (ROT) specific to each supplier are collected in Appendix A.

  7. Neutronic/Thermalhydraulic Coupling Technigues for Sodium Cooled Fast Reactor Simulations

    SciTech Connect

    Jean Ragusa; Andrew Siegel; Jean-Michel Ruggieri

    2010-09-28

    The objective of this project was to test new coupling algorithms and enable efficient and scalable multi-physics simulations of advanced nuclear reactors, with considerations regarding the implementation of such algorithms in massively parallel environments. Numerical tests were carried out to verify the proposed approach and the examples included some reactor transients. The project was directly related to the Sodium Fast Reactor program element of the Generation IV Nuclear Energy Systems Initiative and the Advanced Fuel cycle Initiative, and, supported the requirement of high-fidelity simulation as a mean of achieving the goals of the presidential Global Nuclear Energy Partnership (GNEP) vision.

  8. Impact of commercial precooking of common bean (Phaseolus vulgaris) on the generation of peptides, after pepsin-pancreatin hydrolysis, capable to inhibit dipeptidyl peptidase-IV.

    PubMed

    Mojica, Luis; Chen, Karen; de Mejía, Elvira González

    2015-01-01

    The objective of this research was to determine the bioactive properties of the released peptides from commercially available precook common beans (Phaseolus vulgaris). Bioactive properties and peptide profiles were evaluated in protein hydrolysates of raw and commercially precooked common beans. Five varieties (Black, Pinto, Red, Navy, and Great Northern) were selected for protein extraction, protein and peptide molecular mass profiles, and peptide sequences. Potential bioactivities of hydrolysates, including antioxidant capacity and inhibition of α-amylase, α-glucosidase, dipeptidyl peptidase-IV (DPP-IV), and angiotensin converting enzyme I (ACE) were analyzed after digestion with pepsin/pancreatin. Hydrolysates from Navy beans were the most potent inhibitors of DPP-IV with no statistical differences between precooked and raw (IC50 = 0.093 and 0.095 mg protein/mL, respectively). α-Amylase inhibition was higher for raw Red, Navy and Great Northern beans (36%, 31%, 27% relative to acarbose (rel ac)/mg protein, respectively). α-Glucosidase inhibition among all bean hydrolysates did not show significant differences; however, inhibition values were above 40% rel ac/mg protein. IC50 values for ACE were not significantly different among all bean hydrolysates (range 0.20 to 0.34 mg protein/mL), except for Red bean that presented higher IC50 values. Peptide molecular mass profile ranged from 500 to 3000 Da. A total of 11 and 17 biologically active peptide sequences were identified in raw and precooked beans, respectively. Peptide sequences YAGGS and YAAGS from raw Great Northern and precooked Pinto showed similar amino acid sequences and same potential ACE inhibition activity. Processing did not affect the bioactive properties of released peptides from precooked beans. Commercially precooked beans could contribute to the intake of bioactive peptides and promote health.

  9. Enhanced electricity generation by triclosan and iron anodes in the three-chambered membrane bio-chemical reactor (TC-MBCR).

    PubMed

    Song, Jing; Liu, Lifen; Yang, Fenglin; Ren, Nanqi; Crittenden, John

    2013-11-01

    A three-chambered membrane bio-chemical reactor (TC-MBCR) was developed. The stainless steel membrane modules were used as cathodes and iron plates in the middle chamber served as the anode. The TC-MBCR was able to reduce fouling, remove triclosan (TCS) from a synthetic wastewater treatment and enhance electricity generation by ~60% compared with the cell voltage before TCS addition. The TC-MBCR system generated a relatively stable power output (cell voltage ~0.2V) and the corrosion of iron plates contributed to electricity generation together with microbes on iron anode. The permeation flow from anode to cathode chamber was considered important in electricity generation. In addition, the negatively charged cathode membrane and Fe(2+)/Fe(3+) released by iron plates mitigated membrane fouling by approximately 30%, as compared with the control. The removal of COD and total phosphorus was approximately 99% and 90%. The highest triclosan removal rate reached 97.9%.

  10. Modular High Temperature Gas-Cooled Reactor Safety Basis and Approach

    SciTech Connect

    David Petti; Jim Kinsey; Dave Alberstein

    2014-01-01

    Various international efforts are underway to assess the safety of advanced nuclear reactor designs. For example, the International Atomic Energy Agency has recently held its first Consultancy Meeting on a new cooperative research program on high temperature gas-cooled reactor (HTGR) safety. Furthermore, the Generation IV International Forum Reactor Safety Working Group has recently developed a methodology, called the Integrated Safety Assessment Methodology, for use in Generation IV advanced reactor technology development, design, and design review. A risk and safety assessment white paper is under development with respect to the Very High Temperature Reactor to pilot the Integrated Safety Assessment Methodology and to demonstrate its validity and feasibility. To support such efforts, this information paper on the modular HTGR safety basis and approach has been prepared. The paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach. The paper gives those involved in the assessment of advanced reactor designs an opportunity to assess an advanced design that has already received extensive review by regulatory authorities and to judge the utility of recently proposed new methods for advanced reactor safety assessment such as the Integrated Safety Assessment Methodology.

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

    NASA Astrophysics Data System (ADS)

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

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

  12. A computational modeling approach of the jet-like acoustic streaming and heat generation induced by low frequency high power ultrasonic horn reactors.

    PubMed

    Trujillo, Francisco Javier; Knoerzer, Kai

    2011-11-01

    High power ultrasound reactors have gained a lot of interest in the food industry given the effects that can arise from ultrasonic-induced cavitation in liquid foods. However, most of the new food processing developments have been based on empirical approaches. Thus, there is a need for mathematical models which help to understand, optimize, and scale up ultrasonic reactors. In this work, a computational fluid dynamics (CFD) model was developed to predict the acoustic streaming and induced heat generated by an ultrasonic horn reactor. In the model it is assumed that the horn tip is a fluid inlet, where a turbulent jet flow is injected into the vessel. The hydrodynamic momentum rate of the incoming jet is assumed to be equal to the total acoustic momentum rate emitted by the acoustic power source. CFD velocity predictions show excellent agreement with the experimental data for power densities higher than W(0)/V ≥ 25kWm(-3). This model successfully describes hydrodynamic fields (streaming) generated by low-frequency-high-power ultrasound.

  13. The PLATO IV Student Terminal.

    ERIC Educational Resources Information Center

    Stifle, Jack

    This report describes the remote computer terminal designed for student use in the PLATO IV computer-assisted instruction system. The terminal features a plasma display panel, self-contained character and line generators, and the ability to communicate over voice grade telephone circuits. Operating modes and control characters are described in…

  14. Current generation by helicons and LH waves in modern tokamaks and reactors FNSF-AT, ITER and DEMO. Scenarios, modeling and antennae

    NASA Astrophysics Data System (ADS)

    Vdovin, V.

    2014-02-01

    The Innovative concept and 3D full wave code modeling Off-axis current drive by RF waves in large scale tokamaks, reactors FNSF-AT, ITER and DEMO for steady state operation with high efficiency was proposed [1] to overcome problems well known for LH method [2]. The scheme uses the helicons radiation (fast magnetosonic waves at high (20-40) IC frequency harmonics) at frequencies of 500-1000 MHz, propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by Helicons will help to have regimes with negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure βN > 3 (the so-called Advanced scenarios) of interest for FNSF and the commercial reactor. Modeling with full wave three-dimensional codes PSTELION and STELEC2 showed flexible control of the current profile in the reactor plasmas of ITER, FNSF-AT and DEMO [2,3], using multiple frequencies, the positions of the antennae and toroidal waves slow down. Also presented are the results of simulations of current generation by helicons in tokamaks DIII-D, T-15MD and JT-60SA [3]. In DEMO and Power Plant antenna is strongly simplified, being some analoge of mirrors based ECRF launcher, as will be shown. For spherical tokamaks the Helicons excitation scheme does not provide efficient Off-axis CD profile flexibility due to strong coupling of helicons with O-mode, also through the boundary conditions in low aspect machines, and intrinsic large amount of trapped electrons, as is shown by STELION modeling for the NSTX tokamak. Brief history of Helicons experimental and modeling exploration in straight plasmas, tokamaks and tokamak based fusion Reactors projects is given, including planned joint DIII-D - Kurchatov Institute experiment on helicons CD [1].

  15. Current generation by helicons and LH waves in modern tokamaks and reactors FNSF-AT, ITER and DEMO. Scenarios, modeling and antennae

    SciTech Connect

    Vdovin, V.

    2014-02-12

    The Innovative concept and 3D full wave code modeling Off-axis current drive by RF waves in large scale tokamaks, reactors FNSF-AT, ITER and DEMO for steady state operation with high efficiency was proposed [1] to overcome problems well known for LH method [2]. The scheme uses the helicons radiation (fast magnetosonic waves at high (20–40) IC frequency harmonics) at frequencies of 500–1000 MHz, propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by Helicons will help to have regimes with negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure β{sub N} > 3 (the so-called Advanced scenarios) of interest for FNSF and the commercial reactor. Modeling with full wave three-dimensional codes PSTELION and STELEC2 showed flexible control of the current profile in the reactor plasmas of ITER, FNSF-AT and DEMO [2,3], using multiple frequencies, the positions of the antennae and toroidal waves slow down. Also presented are the results of simulations of current generation by helicons in tokamaks DIII-D, T-15MD and JT-60SA [3]. In DEMO and Power Plant antenna is strongly simplified, being some analoge of mirrors based ECRF launcher, as will be shown. For spherical tokamaks the Helicons excitation scheme does not provide efficient Off-axis CD profile flexibility due to strong coupling of helicons with O-mode, also through the boundary conditions in low aspect machines, and intrinsic large amount of trapped electrons, as is shown by STELION modeling for the NSTX tokamak. Brief history of Helicons experimental and modeling exploration in straight plasmas, tokamaks and tokamak based fusion Reactors projects is given, including planned joint DIII-D – Kurchatov Institute experiment on helicons CD [1].

  16. The Dynomak: An advanced spheromak reactor system with imposed-dynamo current drive and next-generation nuclear power technologies

    NASA Astrophysics Data System (ADS)

    Sutherland, D. A.; Jarboe, T. R.; Marklin, G.; Morgan, K. D.; Nelson, B. A.

    2013-10-01

    A high-beta spheromak reactor system has been designed with an overnight capital cost that is competitive with conventional power sources. This reactor system utilizes recently discovered imposed-dynamo current drive (IDCD) and a molten salt blanket system for first wall cooling, neutron moderation and tritium breeding. Currently available materials and ITER developed cryogenic pumping systems were implemented in this design on the basis of technological feasibility. A tritium breeding ratio of greater than 1.1 has been calculated using a Monte Carlo N-Particle (MCNP5) neutron transport simulation. High-temperature superconducting tapes (YBCO) were used for the equilibrium coil set, substantially reducing the recirculating power fraction when compared to previous spheromak reactor studies. Using zirconium hydride for neutron shielding, a limiting equilibrium coil lifetime of at least thirty full-power years has been achieved. The primary FLiBe loop was coupled to a supercritical carbon dioxide Brayton cycle due to attractive economics and high thermal efficiencies. With these advancements, an electrical output of 1000 MW from a thermal output of 2486 MW was achieved, yielding an overall plant efficiency of approximately 40%. A paper concerning the Dynomak reactor design is currently being reviewed for publication.

  17. Investigation of materials for fusion power reactors

    NASA Astrophysics Data System (ADS)

    Bouhaddane, A.; Slugeň, V.; Sojak, S.; Veterníková, J.; Petriska, M.; Bartošová, I.

    2014-06-01

    The possibility of application of nuclear-physical methods to observe radiation damage to structural materials of nuclear facilities is nowadays a very actual topic. The radiation damage to materials of advanced nuclear facilities, caused by extreme radiation stress, is a process, which significantly limits their operational life as well as their safety. In the centre of our interest is the study of the radiation degradation and activation of the metals and alloys for the new nuclear facilities (Generation IV fission reactors, fusion reactors ITER and DEMO). The observation of the microstructure changes in the reactor steels is based on experimental investigation using the method of positron annihilation spectroscopy (PAS). The experimental part of the work contains measurements focused on model reactor alloys and ODS steels. There were 12 model reactor steels and 3 ODS steels. We were investigating the influence of chemical composition on the production of defects in crystal lattice. With application of the LT 9 program, the spectra of specimen have been evaluated and the most convenient samples have been determined.

  18. Catalytic reactor

    SciTech Connect

    Aaron, Timothy Mark; Shah, Minish Mahendra; Jibb, Richard John

    2009-03-10

    A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

  19. Generation IV Nuclear Energy Systems Construction Cost Reductions through the Use of Virtual Environments - Task 4 Report: Virtual Mockup Maintenance Task Evaluation

    SciTech Connect

    Timothy Shaw; Anthony Baratta; Vaughn Whisker

    2005-02-28

    Task 4 report of 3 year DOE NERI-sponsored effort evaluating immersive virtual reality (CAVE) technology for design review, construction planning, and maintenance planning and training for next generation nuclear power plants. Program covers development of full-scale virtual mockups generated from 3D CAD data presented in a CAVE visualization facility. This report focuses on using Full-scale virtual mockups for nuclear power plant training applications.

  20. Development of a sixth-generation down-flow hanging sponge (DHS) reactor using rigid sponge media for post-treatment of UASB treating municipal sewage.

    PubMed

    Onodera, Takashi; Tandukar, Madan; Sugiyana, Doni; Uemura, Shigeki; Ohashi, Akiyoshi; Harada, Hideki

    2014-01-01

    A sixth-generation down-flow hanging sponge reactor (DHS-G6), using rigid sponge media, was developed as a novel aerobic post-treatment unit for upflow anaerobic sludge blanket (UASB) treating municipal sewage. The rigid sponge media were manufactured by copolymerizing polyurethane with epoxy resin. The UASB and DHS system had a hydraulic retention time (HRT) of 10.6 h (8.6 h for UASB and 2 h for DHS) when operated at 10-28 °C. The system gave reasonable organic and nitrogen removal efficiencies. The final effluent had a total biochemical oxygen demand of only 12 mg/L and a total Kjeldahl nitrogen content of 6 mg/L. The DHS reactor gave particularly good nitrification performance, which was attributed to the new rigid sponge media. The sponge media helped to provide a sufficient HRT, and retained a high biomass concentration, extending the solids retention time. The DHS reactor maintained a high dissolved oxygen concentration under natural ventilation.

  1. Intelligent Virtual Station (IVS)

    NASA Technical Reports Server (NTRS)

    2002-01-01

    The Intelligent Virtual Station (IVS) is enabling the integration of design, training, and operations capabilities into an intelligent virtual station for the International Space Station (ISS). A viewgraph of the IVS Remote Server is presented.

  2. Fuel Development For Gas-Cooled Fast Reactors

    SciTech Connect

    M. K. Meyer

    2006-06-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High Temperature Reactor (VHTR), as well as actinide burning concepts [ ]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is a dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the U.S. and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic ‘honeycomb’ structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  3. ADVANCED SEISMIC BASE ISOLATION METHODS FOR MODULAR REACTORS

    SciTech Connect

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

    2010-09-20

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

  4. Ovarian Cancer Stage IV

    MedlinePlus

    ... hyphen, e.g. -historical Searches are case-insensitive Ovarian Cancer Stage IV Add to My Pictures View /Download : ... 1200x1335 View Download Large: 2400x2670 View Download Title: Ovarian Cancer Stage IV Description: Drawing of stage IV shows ...

  5. A Fast-Spectrum Test Reactor Concept

    SciTech Connect

    Shatilla, Youssef A.; Loewen, Eric P.

    2005-09-15

    The need for a new steady-state fast-neutron reactor has been the subject of numerous national meetings and discussions. This type of reactor will be able to open new frontiers of research for Generation IV reactors, the Space Propulsion Program, and the Advanced Fuel Cycle Initiative. With the confluence of these three programs' fast-spectrum testing needs, we set out to conceptualize a new system by looking at previous successful reactor concepts. This paper presents a new concept for a fast-spectrum test reactor that is horizontal in orientation, with individual pressure tubes running the entire length of the scattering-medium tank filled with a liquid heavy metal. This approach for a test reactor will provide more flexibility in refueling, sample removal, and ability to completely reconfigure the core to meet different users' requirements. Full core neutronic analysis of more than 14 combinations showed that a large hexagonal steam-cooled U-10Zr fuel, with a core power of 267 MW(thermal), produced a fast flux (>0.1 MeV) of 1.3 x 10{sup 15} n/cm{sup 2}.s averaged over the whole length of the irradiation channel. A depletion run with an initial enrichment of 20 wt% {sup 235}U had a flat reactivity curve for the first 180 days of cycle due to in-core breeding. Although considerable neutronic optimization and a thermal-hydraulic analysis remain to be performed, it appears that a reactor core with this innovative geometry could meet future fast flux testing needs.

  6. A Fast-Spectrum Test Reactor Concept

    SciTech Connect

    Youssef A Shatilla; Eric Loewen

    2005-09-01

    The need for a new steady-state fast-neutron reactor has been the subject of numerous national meetings and discussions. This type of reactor will be able to open new frontiers of research for Generation IV reactors, the Space Propulsion Program, and the Advanced Fuel Cycle Initiative. With the confluence of these three programs' fast-spectrum testing needs, we set out to conceptualize a new system by looking at previous successful reactor concepts. This paper presents a new concept for a fast-spectrum test reactor that is horizontal in orientation, with individual pressure tubes running the entire length of the scattering-medium tank filled with a liquid heavy metal. This approach for a test reactor will provide more flexibility in refueling, sample removal, and ability to completely reconfigure the core to meet different users' requirements. Full core neutronic analysis of more than 14 combinations showed that a large hexagonal steam-cooled U-10Zr fuel, with a core power of 267 MW(thermal), produced a fast flux (>0.1 MeV) of 1.3 × 1015 n/cm2s averaged over the whole length of the irradiation channel. A depletion run with an initial enrichment of 20 wt% 235U had a flat reactivity curve for the first 180 days of cycle due to in-core breeding. Although considerable neutronic optimization and a thermal-hydraulic analysis remain to be performed, it appears that a reactor core with this innovative geometry could meet future fast flux testing needs.

  7. Speciation of arsenic(III)/arsenic(V) and selenium(IV)/ selenium(VI) using coupled ion chromatography - hydride generation atomic absorption spectrometry

    Technology Transfer Automated Retrieval System (TEKTRAN)

    Simple analytical methods have been developed to speciate inorganic arsenic and selenium in the ppb range using coupled ion chromatography-hydride generation atomic absorption spectrometry. Because of the differences in toxicity and adsorption behavior, determinations of the redox states arsenite A...

  8. GUIDCOUN: A Comprehensive FORTRAN IV Computer Program for Generating Item and Test Analyses as Well as a Complete Standard Scores Distribution

    ERIC Educational Resources Information Center

    Noble, Gilbert H.

    1977-01-01

    A computer program providing comprehensive test and item analysis is presented. Completing its performance on one run, the program, written in Fortran and emphasizing ease of use, integrates various statistical techniques for analyzing individual items and the overall test, in addition to generating a variety of standard scores. (Author/JKS)

  9. Design of pilot-scale solar photocatalytic reactor for the generation of hydrogen from alkaline sulfide wastewater of sewage treatment plant.

    PubMed

    Priya, R; Kanmani, S

    2013-01-01

    Experiments were conducted for photocatalytic generation of renewable fuel hydrogen from sulphide wastewater from the sewage treatment plant. In this study, pilot-scale solar photocatalytic reactor was designed for treating 1 m3 of sulphide wastewater and also for the simultaneous generation of hydrogen. Bench-scale studies were conducted both in the batch recycle and continuous modes under solar irradiation at similar experimental conditions. The maximum of 89.7% conversion was achieved in the continuous mode. The length of the pilot-scale solar photocatalytic reactor was arrived using the design parameters such as volumetric flow rate (Q) (11 x 10(-2) m3/s), inlet concentration of sulphide ion (C(in)) (28 mol/m3), conversion (89.7%) and average mass flow destruction rate (3.488 x 10(-6) mol/m2 s). The treatment cost of the process was estimated to be 6 US$/m3. This process would be suitable for India like sub-tropical country where sunlight is abundantly available throughout the year.

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

    PubMed

    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.

  11. New reactor technology: safety improvements in nuclear power systems.

    PubMed

    Corradini, M L

    2007-11-01

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

  12. New reactor technology: safety improvements in nuclear power systems.

    PubMed

    Corradini, M L

    2007-11-01

    Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems. PMID:18049233

  13. Gas-Cooled Fast Reactor (GFR) FY04 Annual Report

    SciTech Connect

    K. D. Weaver; T. C. Totemeier; D. E. Clark; E. E. Feldman; E. A. Hoffman; R. B. Vilim; T. Y. C. Wei; J. Gan; M. K. Meyer; W. F. Gale; M. J. Driscoll; M. Golay; G. Apostolakis; K. Czerwinski

    2004-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection.

  14. The long-term future for civilian nuclear power generation in France: The case for breeder reactors. Breeder reactors: The physical and physical chemistry parameters, associate material thermodynamics and mechanical engineering: Novelties and issues

    NASA Astrophysics Data System (ADS)

    Dautray, Robert

    2011-06-01

    , etc., radioprotection measures and treatment for the "transuranic" elements. For a long period of time, France was in the forefront of nuclear breeder power generation science, technological research and also in the knowledge base related to breeder reactors. It is in the country's interest to pursue these efforts and this could per se constitute one of the national priorities. Nous sommes naturellement bien conscients de l'énorme problème qui se pose au Japon actuellement comme suite au tremblement de terre et au tsunami de mars 2011 et leurs conséquences, notamment sur des installations électronucléaires. Le texte que nous présentons concerne des conditions totalement générales, indépendantes des problèmes spécifiques de sûreté qu'il faudra, de toute façon, traiter dans le cadre d'un développement éventuel de l'énergie nucléaire.We are aware, of course, of the huge problem that Japan has to deal with the aftermath of the quake and tsunami of March 2011 and their consequences on electronuclear power plants. The text that we present here concerns general physical topics independent of the specific safety problems, general physical topics which will have to be solved in the case of a contingent development of electronuclear power plants.

  15. Health Monitoring to Support Advanced Small Modular Reactors

    SciTech Connect

    Coble, Jamie B.; Meyer, Ryan M.; Ramuhalli, Pradeep

    2013-08-01

    Advanced small modular reactors (aSMRs) are based on advanced reactor concepts, some of which were promoted by the Generation IV International Forum, and are being considered for diverse missions including desalination of water, production of hydrogen, etc. While the existing fleet of commercial nuclear reactors provides baseload electricity, it is conceivable that aSMRs could be implemented for both baseload and load following applications. The effect of diverse operating missions and unit modularity on plant operations and maintenance (O&M) is not fully understood and limiting these costs will be essential to successful deployment of aSMRs. Integrated health monitoring concepts are proposed to support the safe and affordable operation of aSMRs over their lifetime by enabling management of significant in-vessel and in-containment active and passive components.

  16. Sonochemical Reactors.

    PubMed

    Gogate, Parag R; Patil, Pankaj N

    2016-10-01

    Sonochemical reactors are based on the generation of cavitational events using ultrasound and offer immense potential for the intensification of physical and chemical processing applications. The present work presents a critical analysis of the underlying mechanisms for intensification, available reactor configurations and overview of the different applications exploited successfully, though mostly at laboratory scales. Guidelines have also been presented for optimum selection of the important operating parameters (frequency and intensity of irradiation, temperature and liquid physicochemical properties) as well as the geometric parameters (type of reactor configuration and the number/position of the transducers) so as to maximize the process intensification benefits. The key areas for future work so as to transform the successful technique at laboratory/pilot scale into commercial technology have also been discussed. Overall, it has been established that there is immense potential for sonochemical reactors for process intensification leading to greener processing and economic benefits. Combined efforts from a wide range of disciplines such as material science, physics, chemistry and chemical engineers are required to harness the benefits at commercial scale operation.

  17. Sonochemical Reactors.

    PubMed

    Gogate, Parag R; Patil, Pankaj N

    2016-10-01

    Sonochemical reactors are based on the generation of cavitational events using ultrasound and offer immense potential for the intensification of physical and chemical processing applications. The present work presents a critical analysis of the underlying mechanisms for intensification, available reactor configurations and overview of the different applications exploited successfully, though mostly at laboratory scales. Guidelines have also been presented for optimum selection of the important operating parameters (frequency and intensity of irradiation, temperature and liquid physicochemical properties) as well as the geometric parameters (type of reactor configuration and the number/position of the transducers) so as to maximize the process intensification benefits. The key areas for future work so as to transform the successful technique at laboratory/pilot scale into commercial technology have also been discussed. Overall, it has been established that there is immense potential for sonochemical reactors for process intensification leading to greener processing and economic benefits. Combined efforts from a wide range of disciplines such as material science, physics, chemistry and chemical engineers are required to harness the benefits at commercial scale operation. PMID:27573503

  18. Analysis and selection of high pressure heaters design for a new generation of NPP with BN-1200 reactor plant

    NASA Astrophysics Data System (ADS)

    Yurchenko, A. Yu.; Sukhorukov, Yu. G.; Trifonov, N. N.; Grigor'eva, E. B.; Esin, S. B.; Svyatkin, F. A.; Nikolaenkova, E. K.; Prikhod'ko, P. Yu.; Nazarov, V. V.

    2016-09-01

    In the development of advanced high-power steam-turbine plants (STP), special attention is placed on the design of reliable and economical high-pressure heater (HPH) capable to maintain the specified thermal hydraulic performance during the entire service life. Comparative analysis of the known designs of HPH, such as the spiral-collector HPH, the collector-coiled HPH, the collector-platen HPH, modular HPH, and the chamber HPH, was carried out. The advantages and disadvantages of each design were pointed. For better comparison, the heaters are separated into two groups—horizontal and vertical ones. The weight and dimension characteristics, the materials and features of the basic elements, and operating features of variety HPH are presented. At operating the spiral-collector HPH used in the majority of regenerative schemes of high-pressure STP of thermal and nuclear power plants, the disadvantages reducing the economy and reliability of their operation were revealed. The recommendations directed to the reliability growth of HPH, the decrease of subcooling the feed water, the increase of compactness are stated. Some of these were developed by the specialists of OAO NPO TsKTI and are successfully implemented on the thermal power plants and nuclear power plants. Technical solutions to reduce the cost of regeneration system and the weight of chamber HPH, reduce the thickness of the tube plate of HPH, and reliability assurance of the cooler of steam and condensate built in the HPH casing under all operating conditions were proposed. Three types of feed water chambers for vertical and horizontal chamber HPH are considered in detail, the constructive solutions that have been implemented in HPH of the regeneration system of turbines of 1000 and 1200 MW capacity with water-moderated water-cooled power reactor (WMWCPR) are described. The optimal design of HPH for the regeneration system of high-pressure turbine plant with BN-1200 reactor was selected.

  19. The role of minerals in the thermal alteration of organic matter. IV - Generation of n-alkanes, acyclic isoprenoids, and alkenes in laboratory experiments

    NASA Technical Reports Server (NTRS)

    Huizinga, Bradley J.; Tannenbaum, Eli; Kaplan, Isaac R.

    1987-01-01

    The effect of common sedimentary minerals (illite, Na-montmorillonite, or calcite) under different water concentrations on the generation and release of n-alkanes, acyclic isoprenoids, and select alkenes from oil-prone kerogens was investigated. Matrices containing Green River Formation kerogen or Monterey Formation kerogen, alone or in the presence of minerals, were heated at 200 or 300 C for periods of up to 1000 hours, and the pyrolysis products were analyzed. The influence of the first two clay minerals was found to be critically dependent on the water content. Under the dry pyrolysis conditions, both minerals significantly reduced alkene formation; the C12+ n-alkanes and acyclic isoprenoids were mostly destroyed by montmorillonite, but underwent only minor alteration with illite. Under hydrous conditions (mineral/water of 2/1), the effects of both minerals were substantially reduced. Calcite had no significant effect on the thermal evolution of the hydrocarbons.

  20. Feasibility study on ultralong-cycle operation and material performance for compact liquid metal-cooled fast reactors: a review work

    SciTech Connect

    Tak, Taewoo; Choe, Jiwon; Jeong, Yongjin; Lee, Deokjung; Kim, T. K.; Hong, Ser Gi

    2015-11-01

    This paper reviews the feasibility of ultralong-cycle operation on a compact liquid metal-cooled fast reactor (LMR) firstly by assessing the operation of a long-life fast reactor core and secondly by evaluating material performance in respect to both long-cycle operation and compact-size fast reactor. Many kinds of reactor concepts have been proposed, and LMR and small modular reactor (SMR) are the issued leading technologies for generation four (Gen-IV) reactor system development. The breed-and-burn strategy was proposed as a core burning strategy to operate a long cycle, and it has been evaluated in this paper with two reactor concepts: constant axial shape of neutron flux, nuclide densities, and power shape during life of energy and ultralong cycle fast reactor. In addition, Super-Safe, Small, and Simple and small modular fast reactor, compact LMR concepts, have been simulated to evaluate their long-life operation strategies. For the other practical issues, the materials for fuel, coolant, and structure have been identified and some of them are selected to have their performance optimized specifically for compact LMR with a long-cycle operation. It is believed that this comprehensive review will propose a proper direction for future reactor development and will be followed by the next step research for a complete reactor model with the other reactor components.

  1. Phase IV of Drug Development.

    PubMed

    Suvarna, Viraj

    2010-04-01

    Not all Phase IV studies are post-marketing surveillance (PMS) studies but every PMS study is a phase IV study. Phase IV is also an important phase of drug development. In particular, the real world effectiveness of a drug as evaluated in an observational, non-interventional trial in a naturalistic setting which complements the efficacy data that emanates from a pre-marketing randomized controlled trial (RCT). No matter how many patients are studied pre-marketing in a controlled environment, the true safety profile of a drug is characterized only by continuing safety surveillance through a spontaneous adverse event monitoring system and a post-marketing surveillance/non-interventional study. Prevalent practice patterns can generate leads that could result in further evaluation of a new indication via the RCT route or even a signal that may necessitate regulatory action (change in labeling, risk management/minimization action plan). Disease registries are another option as are the large simple hybrid trials. Surveillance of spontaneously reported adverse events continues as long as a product is marketed. And so Phase IV in that sense never ends.

  2. Energy levels and lifetimes of Nd IV, Pm IV, Sm IV, and Eu IV

    SciTech Connect

    Dzuba, V. A.; Safronova, U. I.; Johnson, W. R.

    2003-09-01

    To address the shortage of experimental data for electron spectra of triply ionized rare-earth elements we have calculated energy levels and lifetimes of 4f{sup n+1} and 4f{sup n}5d configurations of Nd IV (n=2), Pm IV (n=3), Sm IV (n=4), and Eu IV (n=5) using Hartree-Fock and configuration-interaction methods. To control the accuracy of our calculations we also performed similar calculations for Pr III, Nd III, and Sm III, for which experimental data are available. The results are important, in particular, for physics of magnetic garnets.

  3. Post-scram Liquid Metal cooled Fast Breeder Reactor (LMFBR) heat transport system dynamics and steam generator control: Figures

    NASA Astrophysics Data System (ADS)

    Brukx, J. F. L. M.

    1982-06-01

    Dynamic modeling of LMFBR heat transport system is discussed. Uncontrolled transient behavior of individual components and of the integrated heat transport system are considered. For each component, results showing specific dynamic features of the component and/or model capability were generated. Controlled dynamic behavior for alternative steam generator control systems during forced and natural sodium coolant circulation was analyzed. Combined free and forced convection of laminar and turbulent vertical pipe flow of liquid metals was investigated.

  4. Co3O4-based honeycombs as compact redox reactors/heat exchangers for thermochemical storage in the next generation CSP plants

    NASA Astrophysics Data System (ADS)

    Pagkoura, Chrysoula; Karagiannakis, George; Halevas, Eleftherios; Konstandopoulos, Athanasios G.

    2016-05-01

    Over the last years, several research groups have focused on developing efficient thermochemical heat storage (THS) systems, in-principle capable of being coupled with next generation high temperature Concentrated Solar Power plants. Among systems studied, the Co3O4/CoO redox system is a promising candidate. Currently, research efforts extend beyond basic level identification of promising materials to more application-oriented approaches aiming at validation of THS performance at pilot scale reactors. The present work focuses on the investigation of cobalt oxide based honeycomb structures as candidate reactors/heat exchangers to be employed for such purposes. In the evaluation conducted and presented here, cobalt oxide-based structures with different composition and geometrical characteristics were subjected to redox cycles in the temperature window between 800 and 1000°C under air flow. Basic aspects related to redox performance of each system are briefly discussed but the main focus lies on the evaluation of the segments structural stability after multi-cyclic operation. The latter is based on macroscopic visual observation and also supplemented by pre- (i.e. fresh samples) and post-characterization (i.e. after long term exposure) of extruded honeycombs via combined mercury porosimetry and SEM analysis.

  5. Equilibrium and kinetic studies of in situ generation of ammonia from urea in a batch reactor for flue gas conditioning of thermal power plants

    SciTech Connect

    Sahu, J.N.; Patwardhan, A.V.; Meikap, B.C.

    2009-03-15

    Ammonia has long been known to be useful in the treatment of flue/tail/stack gases from industrial furnaces, incinerators, and electric power generation industries. In this study, urea hydrolysis for production of ammonia, in different application areas that require safe use of ammonia at in situ condition, was investigated in a batch reactor. The equilibrium and kinetic study of urea hydrolysis was done in a batch reactor at reaction pressure to investigate the effect of reaction temperature, initial feed concentration, and time on ammonia production. This study reveals that conversion increases exponentially with an increase in temperature but with increases in initial feed concentration of urea the conversion decreases marginally. Further, the effect of time on conversion has also been studied; it was found that conversion increases with increase in time. Using collision theory, the temperature dependency of forward rate constant developed from which activation energy of the reaction and the frequency factor has been calculated. The activation energy and frequency factor of urea hydrolysis reaction at atmospheric pressure was found to be 73.6 kJ/mol and 2.89 x 10{sup 7} min{sup -1}, respectively.

  6. A spectroscopic study of ethylene destruction and by-product generation using a three-stage atmospheric packed-bed plasma reactor

    SciTech Connect

    Huebner, M.; Roepcke, J.; Guaitella, O.; Rousseau, A.

    2013-07-21

    Using a three-stage dielectric packed-bed plasma reactor at atmospheric pressure, the destruction of ethylene, a typical volatile organic compound, and the generation of major by-products have been studied by means of Fourier Transform Infrared Spectroscopy. A test gas mixture air at a gas flow of 1 slm containing 0.12% humidity with 0.1% ethylene has been used. In addition to the fragmentation of the precursor gas, the evolution of the concentration of ten stable reaction products, CO, CO{sub 2}, O{sub 3}, NO{sub 2}, N{sub 2}O, HCN, H{sub 2}O, HNO{sub 3}, CH{sub 2}O, and CH{sub 2}O{sub 2} has been monitored. The concentrations of the by-products range between 5 ppm, in the case of NO{sub 2}, and 1200 ppm, for H{sub 2}O. By the application of three sequentially working discharge cells at a frequency of f = 4 kHz and voltage values between 9 and 12 kV, a nearly complete decomposition of C{sub 2}H{sub 4} could be achieved. Furthermore, the influence of the specific energy deposition (SED) on the destruction process has been studied and the maximum value of SED was about 900 J l{sup -1}. The value of the characteristic energy {beta}, characterizing the energy efficiency of the ethylene destruction in the reactor, was found to be 330 J l{sup -1}. It was proven that the application of three reactor stages suppresses essentially the production of harmful by-products as formaldehyde, formic acid, and NO{sub 2} compared to the use of only one or two stages. Based on the multi-component detection, the carbon balance of the plasma chemical conversion of ethylene has been analyzed. The dependence of the fragmentation efficiencies of ethylene (R{sub F}(C{sub 2}H{sub 4}) = 5.5 Multiplication-Sign 10{sup 19} molecules J{sup -1}) and conversion efficiencies to the produced molecular species (R{sub C} = (0.1-3) Multiplication-Sign 10{sup 16} molecules J{sup -1}) on the discharge conditions could be estimated in the multistage plasma reactor.

  7. Interaction of radiation-generated radicals with myoglobin in aqueous solution—IV. Mechanism of interaction of hydroxyl radicals with oxymyoglobin

    NASA Astrophysics Data System (ADS)

    Whitburn, Kevin D.; Hoffman, Morton Z.

    The interaction of radiation-generated ·OH/H· with oxymyoglobin (MbO 2) has been studied in the presence of catalase at pH 7.3 over the range of 5-510microM O 2. The conversion of MbO 2 to heme-modified products has been examined under conditions where depletion of O 2 in irradiated solutions both can and cannot be compensated by O 2-transfer across the solution phase boundary. In the theoretical limit of [O 2] → 0 in bulk solution, MbO 2 is converted stoichiometrically to ferri- and ferromyoglobin with G( MbO 2) ⋍ 6.0, G(ferroMb) ⋍ 3.0, and G(ferriMb) ⋍ 3.0. An increase in [O 2] in bulk solution beyond the zero-limit progressively suppresses the conversion of MbO 2 to the heme-modified derivatives. At [O 2] ⩾ 300 microM, an O 2-independent path of ferriMb formation with G ⋍ 0.6 is evident. Two sources of ferriMb induced by ·OH/H· are proposed: an O 2-independent path involving direct oxidative attack of ·OH at the oxyferroheme, and O 2-dependent paths of production of ferriMb and ferroMb involving the mediation of O 2-scavengable secondary hemeprotein radicals. It is suggested that the modifications of the heme group in the absence of O 2 are accompanied by redox modifications on the globin moiety. With increasing [O 2], similar redox modifications on the globin can occur without a mediating involvement of the prosthetic group. At high [O 2], involvement of the heme in modification of the globin is eliminated.

  8. Mechanism of assembly of the dimanganese-tyrosyl radical cofactor of class Ib ribonucleotide reductase: enzymatic generation of superoxide is required for tyrosine oxidation via a Mn(III)Mn(IV) intermediate.

    PubMed

    Cotruvo, Joseph A; Stich, Troy A; Britt, R David; Stubbe, JoAnne

    2013-03-13

    Ribonucleotide reductases (RNRs) utilize radical chemistry to reduce nucleotides to deoxynucleotides in all organisms. In the class Ia and Ib RNRs, this reaction requires a stable tyrosyl radical (Y(•)) generated by oxidation of a reduced dinuclear metal cluster. The Fe(III)2-Y(•) cofactor in the NrdB subunit of the class Ia RNRs can be generated by self-assembly from Fe(II)2-NrdB, O2, and a reducing equivalent. By contrast, the structurally homologous class Ib enzymes require a Mn(III)2-Y(•) cofactor in their NrdF subunit. Mn(II)2-NrdF does not react with O2, but it binds the reduced form of a conserved flavodoxin-like protein, NrdIhq, which, in the presence of O2, reacts to form the Mn(III)2-Y(•) cofactor. Here we investigate the mechanism of assembly of the Mn(III)2-Y(•) cofactor in Bacillus subtilis NrdF. Cluster assembly from Mn(II)2-NrdF, NrdI(hq), and O2 has been studied by stopped flow absorption and rapid freeze quench EPR spectroscopies. The results support a mechanism in which NrdI(hq) reduces O2 to O2(•-) (40-48 s(-1), 0.6 mM O2), the O2(•-) channels to and reacts with Mn(II)2-NrdF to form a Mn(III)Mn(IV) intermediate (2.2 ± 0.4 s(-1)), and the Mn(III)Mn(IV) species oxidizes tyrosine to Y(•) (0.08-0.15 s(-1)). Controlled production of O2(•-) by NrdIhq during class Ib RNR cofactor assembly both circumvents the unreactivity of the Mn(II)2 cluster with O2 and satisfies the requirement for an "extra" reducing equivalent in Y(•) generation.

  9. The implementation of a 3D characteristics solver for the generation of incremental cross sections for reactivity devices in a CANDU reactor

    SciTech Connect

    Le Tellier, R.; Hebert, A.; Marleau, G.

    2006-07-01

    We are presenting issues related to the generation of consistent incremental cross sections for the reactivity devices in a CANDU reactor. Such calculations involve the solution of the neutron transport equation over complex 3D geometries representing a single vertical reactivity device inserted mid-way between two horizontal fuel channels. The DRAGON lattice code has recently been upgraded and can handle the exact geometry of such configurations for trajectory-based transport solvers. Within this framework, the detailed representation of the reactivity devices implies an increase in the number of regions when the strongly absorbing regions and fuel clusters are described without cylinderization. In this paper, a solution based on the characteristics method is compared with the standard procedure, based on the collision probabilities method. The coherence of both solvers is highlighted and a comparison of their computational costs is presented. (authors)

  10. Development of a multiphysics analysis system for sodium-water reaction phenomena in steam generators of sodium-cooled fast reactors

    SciTech Connect

    Uchibori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki

    2015-12-31

    A multiphysics analysis system for sodium-water reaction phenomena in a steam generator of sodium-cooled fast reactors was newly developed. The analysis system consists of the mechanistic numerical analysis codes, SERAPHIM, TACT, and RELAP5. The SERAPHIM code calculates the multicomponent multiphase flow and sodium-water chemical reaction caused by discharging of pressurized water vapor. Applicability of the SERAPHIM code was confirmed through the analyses of the experiment on water vapor discharging in liquid sodium. The TACT code was developed to calculate heat transfer from the reacting jet to the adjacent tube and to predict the tube failure occurrence. The numerical models integrated into the TACT code were verified through some related experiments. The RELAP5 code evaluates thermal hydraulic behavior of water inside the tube. The original heat transfer correlations were corrected for the tube rapidly heated by the reacting jet. The developed system enables evaluation of the wastage environment and the possibility of the failure propagation.

  11. Development of a multiphysics analysis system for sodium-water reaction phenomena in steam generators of sodium-cooled fast reactors

    NASA Astrophysics Data System (ADS)

    Uchibori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki

    2015-12-01

    A multiphysics analysis system for sodium-water reaction phenomena in a steam generator of sodium-cooled fast reactors was newly developed. The analysis system consists of the mechanistic numerical analysis codes, SERAPHIM, TACT, and RELAP5. The SERAPHIM code calculates the multicomponent multiphase flow and sodium-water chemical reaction caused by discharging of pressurized water vapor. Applicability of the SERAPHIM code was confirmed through the analyses of the experiment on water vapor discharging in liquid sodium. The TACT code was developed to calculate heat transfer from the reacting jet to the adjacent tube and to predict the tube failure occurrence. The numerical models integrated into the TACT code were verified through some related experiments. The RELAP5 code evaluates thermal hydraulic behavior of water inside the tube. The original heat transfer correlations were corrected for the tube rapidly heated by the reacting jet. The developed system enables evaluation of the wastage environment and the possibility of the failure propagation.

  12. Procedure of calculation of the spatial distribution of temperatures and heat fluxes in the steam generator of a nuclear power installation with an RBEC fast-neutron reactor

    NASA Astrophysics Data System (ADS)

    Frolov, A. A.; Sedov, A. A.

    2016-08-01

    A method for combined 3D/1D-modeling of thermohydraulics of a once-through steam generator (SG) based on the joint analysis of three-dimensional thermo- and hydrodynamics of a single-phase heating coolant in the intertube space and one-dimensional thermohydraulics of steam-generating channels (tubes) with the use of well-known friction and heat-transfer correlations under various boiling conditions is discussed. This method allows one to determine the spatial distribution of temperatures and heat fluxes of heat-exchange surfaces of SGs with a single-phase heating coolant in the intertube space and with steam generation within tubes. The method was applied in the analytical investigation of typical operation of a once-through SG of a nuclear power installation with an RBEC fast-neutron heavy-metal reactor that is being designed by Kurchatov Institute in collaboration with OKB GIDROPRESS and Leipunsky Institute of Physics and Power Engineering. Flow pattern and temperature fields were obtained for the heavy-metal heating coolant in the intertube space. Nonuniformities of heating of the steam-water coolant in different heat-exchange tubes and nonuniformities in the distribution of heat fluxes at SG heat-exchange surfaces were revealed.

  13. RELAP5/MOD3 Analysis of Transient Steam-Generator Behavior During Turbine Trip Test of a Prototype Fast Breeder Reactor MONJU

    SciTech Connect

    Yoshihisa Shindo; Hiroshi Endo; Tomoko Ishizu; Kazuo Haga

    2006-07-01

    In order to develop a thermal-hydraulic model of the steam-generator (SG) to simulate transient phenomena in the sodium cooled fast breeder reactor (FBR) MONJU, Japan Nuclear Energy Safety Organization (JNES) verified the SG model using the RELAP5/MOD3 code against the results of the turbine trip test at a 40% power load of MONJU. The modeling by using RELAP5 was considered to explain the significant observed behaviors of the pressure and the temperature of the EV steam outlet, and the temperature of water supply distributing piping till 600 seconds after the turbine trip. The analysis results of these behaviors showed good agreement with the test results based on results of parameter study as the blow efficiency (release coef.) and heat transferred from the helical coil region to the down-comer (temperature heating down-comer tubes). It was found that the RELAP5/MOD3 code with a two-fluids model can predict well the physical situation: the gas-phase of steam generated by the decompression boiling moves upward in the down-comer tubes accompanied by the enthalpy increase of the water supply chambers; and that the pressure change of a 'shoulder' like shape is induced by the mass balance between the steam mass generated in the down-comer tubes and the steam mass blown from the SG. The applicability of RELAP5/MOD3 to SG modeling was confirmed by simulating the actual FBR system. (authors)

  14. Power Conversion Study for High Temperature Gas-Cooled Reactors

    SciTech Connect

    Chang Oh; Richard Moore; Robert Barner

    2005-05-01

    The Idaho National Laboratory (INL) is investigating a Brayton cycle efficiency improvement on a high temperature gas-cooled reactor (HTGR) as part of Generation-IV nuclear engineering research initiative. There are some technical issues to be resolved before the selection of the final design of the high temperature gascooled reactor, called as a Next Generation Nuclear Plant (NGNP), which is supposed to be built at the INEEL by year 2017. The technical issues are the selection of the working fluid, direct vs. indirect cycle, power cycle type, the optimized design in terms of a number of intercoolers, and others. In this paper, we investigated a number of working fluids for the power conversion loop, direct versus indirect cycle, the effect of intercoolers, and other thermal hydraulics issues. However, in this paper, we present part of the results we have obtained. HYSYS computer code was used along with a computer model developed using Visual Basic computer language.

  15. Using PLATO IV.

    ERIC Educational Resources Information Center

    Meller, David V.

    This beginning reference manual describes PLATO IV hardware for prospective users and provides an introduction to PLATO for new authors. The PLATO terminal is described in detail in Chapter 1. Chapter 2 provides a block diagram of the PLATO IV system. Procedures for getting on line are described in Chapter 3, and Chapter 4 provides references to…

  16. SSTAR: The U.S. Lead-Cooled Fast Reactor (LFR)

    SciTech Connect

    Smith, C F; Halsey, W G; Brown, N W; Sienicki, J J; Moisseytsev, A; Wade, D C

    2007-09-25

    It is widely recognized that the developing world is the next area for major energy demand growth, including demand for new and advanced nuclear energy systems. With limited existing industrial and grid infrastructures, there will be an important need for future nuclear energy systems that can provide small or moderate increments of electric power (10-700 MWe) on small or immature grids in developing nations. Most recently, the Global Nuclear Energy Partnership (GNEP) has identified, as one of its key objectives, the development and demonstration of concepts for small and medium sized reactors (SMRs) that can be globally deployed while assuring a high level of proliferation resistance. Lead-cooled systems offer several key advantages in meeting these goals. The small lead-cooled fast reactor concept known as the Small Secure Transportable Autonomous Reactor (SSTAR) reactor has been under ongoing development under the U.S. Generation IV Nuclear Energy Systems Initiative. It a system designed to provide energy security to developing nations while incorporating features to achieve nonproliferation aims, anticipating GNEP objectives. This paper presents the motivation for development of internationally deployable nuclear energy systems as well as a summary of one such system, SSTAR, which is the U.S. Generation IV Lead-cooled Fast Reactor system.

  17. Generations.

    PubMed

    Chambers, David W

    2005-01-01

    Groups naturally promote their strengths and prefer values and rules that give them an identity and an advantage. This shows up as generational tensions across cohorts who share common experiences, including common elders. Dramatic cultural events in America since 1925 can help create an understanding of the differing value structures of the Silents, the Boomers, Gen Xers, and the Millennials. Differences in how these generations see motivation and values, fundamental reality, relations with others, and work are presented, as are some applications of these differences to the dental profession. PMID:16623137

  18. Pressure Vessel and Internals of the International Reactor Innovative and Secure

    SciTech Connect

    Lombardi, C.V.; Padovani, E.; Cammi, A.; Collado, J.M.; Santoro, R.T.; Barnes, J.M.

    2002-07-01

    IRIS (International Reactor Innovative and Secure) is a modular, integral light water cooled, low-to-medium power reactor, which addresses the requirements defined by the US DOE for Generation IV reactors. Its integrated layout features a pressure vessel containing all the main primary circuit components: the internals and the biological shield, here described together with the pressure vessel, plus the steam generators, the pressurizer, and the main coolant pumps described in companion papers. For this reason the pressure vessel is a crucial component of the plant, which deserves the most demanding design effort. The wide inner annulus around the core is exploited to insert steel plates, in order to improve the inner shielding capability up to the elimination of the external biological shielding and to simplify decommissioning activities by having all the irradiated components inside the vessel. (authors)

  19. Simulation of hydration/dehydration of CaO/Ca(OH){sub 2} chemical heat pump reactor for cold/hot heat generation

    SciTech Connect

    Ogura, Hironao; Shimojyo, Rui; Kage, Hiroyuki; Matsuno, Yoshizo; Mujumdar, A.S.

    1999-09-01

    A chemical heat pump (CHP) utilizes reversible reactions involving significant endothermic and exothermic heats of reaction in order to develop a heat pump effect by storing and releasing energy while transforming it from chemical to thermal energy and vice versa. In this paper, the authors present a mathematical model and its numerical solution for the heat and mass transport phenomena occurring in the reactant particle bed of the CHP for heat storage and cold/hot heat generation based on the CaO/Ca(OH){sub 2} reversible hydration/dehydration reaction. Transient conservation equations of mass and energy transport including chemical kinetics are solved numerically subject to appropriate boundary and initial conditions to examine the influence of the mass transfer resistance on the overall performance of this CHP configuration. These results are presented and discussed with the aim of enhancing the CHP performance in the next generation reactor designs. The CHP can store thermal energy in industrial waste heat, solar heat, terrestrial heat, etc. in the form of chemical energy, and release it at various temperature levels during the heat-demand period.

  20. The generation of gravitational waves. IV - Bremsstrahlung

    NASA Technical Reports Server (NTRS)

    Kovacs, S. J., Jr.; Thorne, K. S.

    1978-01-01

    Previously derived waveforms for gravitational bremsstrahlung are discussed along with their spectra and their limiting structure at high and low relative velocities. Waveforms and spectra are presented for a low-velocity bremsstrahlung encounter, and waveforms are given for encounters of arbitrary relative velocity. Limiting forms for the gravitational-wave amplitudes in the 'forward', 'intermediate', and 'backward' regions are derived in the high-velocity limit. The energy spectra seen by observers in the three regions are computed for arbitrary and high velocities. Simpler methods for analyzing special cases of the bremsstrahlung problem are examined, and the results of those methods are compared with the present results. Those methods include the quadrupole-moment and post-Newtonian formalisms, linear perturbations of the Schwarzschild metric, the method of colliding plane waves, the method of virtual quanta, and the zero-frequency limit. Classical gravitational bremsstrahlung is then compared with classical electromagnetic bremsstrahlung.

  1. IV treatment at home

    MedlinePlus

    ... home; PICC line - home; Infusion therapy - home; Home health care - IV treatment ... Often, home health care nurses will come to your home to give you the medicine. Sometimes, a family member, a friend, or ...

  2. LFR "Lead-Cooled Fast Reactor"

    SciTech Connect

    Cinotti, L; Fazio, C; Knebel, J; Monti, S; Abderrahim, H A; Smith, C; Suh, K

    2006-05-11

    The main purpose of this paper is to present the current status of development of the Lead-cooled Fast Reactor (LFR) in Generation IV (GEN IV), including the European contribution, to identify needed R&D and to present the corresponding GEN IV International Forum (GIF) R&D plan [1] to support the future development and deployment of lead-cooled fast reactors. The approach of the GIF plan is to consider the research priorities of each member country in proposing an integrated, coordinated R&D program to achieve common objectives, while avoiding duplication of effort. The integrated plan recognizes two principal technology tracks: (1) a small, transportable system of 10-100 MWe size that features a very long refuelling interval, and (2) a larger-sized system rated at about 600 MWe, intended for central station power generation. This paper provides some details of the important European contributions to the development of the LFR. Sixteen European organizations have, in fact, taken the initiative to present to the European Commission the proposal for a Specific Targeted Research and Training Project (STREP) devoted to the development of a European Lead-cooled System, known as the ELSY project; two additional organizations from the US and Korea have joined the project. Consequently, ELSY will constitute the reference system for the large lead-cooled reactor of GEN IV. The ELSY project aims to demonstrate the feasibility of designing a competitive and safe fast power reactor based on simple technical engineered features that achieves all of the GEN IV goals and gives assurance of investment protection. As far as new technology development is concerned, only a limited amount of R&D will be conducted in the initial phase of the ELSY project since the first priority is to define the design guidelines before launching a larger and expensive specific R&D program. In addition, the ELSY project is expected to benefit greatly from ongoing lead and lead-alloy technology

  3. Nuclear Reactor Physics

    NASA Astrophysics Data System (ADS)

    Stacey, Weston M.

    2001-02-01

    An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

  4. GCF Mark IV development

    NASA Technical Reports Server (NTRS)

    Mortensen, L. O.

    1982-01-01

    The Mark IV ground communication facility (GCF) as it is implemented to support the network consolidation program is reviewed. Changes in the GCF are made in the area of increased capacity. Common carrier circuits are the medium for data transfer. The message multiplexing in the Mark IV era differs from the Mark III era, in that all multiplexing is done in a GCF computer under GCF software control, which is similar to the multiplexing currently done in the high speed data subsystem.

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

    SciTech Connect

    C. A. Wemple

    2005-09-01

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

  6. Constraints on silicates formation in the Si-Al-Fe system: Application to hard deposits in steam generators of PWR nuclear reactors

    NASA Astrophysics Data System (ADS)

    Berger, Gilles; Million-Picallion, Lisa; Lefevre, Grégory; Delaunay, Sophie

    2015-04-01

    Introduction: The hydrothermal crystallization of silicates phases in the Si-Al-Fe system may lead to industrial constraints that can be encountered in the nuclear industry in at least two contexts: the geological repository for nuclear wastes and the formation of hard sludges in the steam generator of the PWR nuclear plants. In the first situation, the chemical reactions between the Fe-canister and the surrounding clays have been extensively studied in laboratory [1-7] and pilot experiments [8]. These studies demonstrated that the high reactivity of metallic iron leads to the formation of Fe-silicates, berthierine like, in a wide range of temperature. By contrast, the formation of deposits in the steam generators of PWR plants, called hard sludges, is a newer and less studied issue which can affect the reactor performance. Experiments: We present here a preliminary set of experiments reproducing the formation of hard sludges under conditions representative of the steam generator of PWR power plant: 275°C, diluted solutions maintained at low potential by hydrazine addition and at alkaline pH by low concentrations of amines and ammoniac. Magnetite, a corrosion by-product of the secondary circuit, is the source of iron while aqueous Si and Al, the major impurities in this system, are supplied either as trace elements in the circulating solution or by addition of amorphous silica and alumina when considering confined zones. The fluid chemistry is monitored by sampling aliquots of the solution. Eh and pH are continuously measured by hydrothermal Cormet© electrodes implanted in a titanium hydrothermal reactor. The transformation, or not, of the solid fraction was examined post-mortem. These experiments evidenced the role of Al colloids as precursor of cements composed of kaolinite and boehmite, and the passivation of amorphous silica (becoming unreactive) likely by sorption of aqueous iron. But no Fe-bearing was formed by contrast to many published studies on the Fe

  7. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.; Zinn, W.H.; Anderson, H.L.

    1958-09-16

    Means are presenied for increasing the reproduction ratio of a gaphite- moderated neutronic reactor by diminishing the neutron loss due to absorption or capture by gaseous impurities within the reactor. This means comprised of a fluid-tight casing or envelope completely enclosing the reactor and provided with a valve through which the casing, and thereby the reactor, may be evacuated of atmospheric air.

  8. Future Scenarios for Fission Based Reactors

    NASA Astrophysics Data System (ADS)

    David, S.

    2005-04-01

    The coming century will see the exhaustion of standard fossil fuels, coal, gas and oil, which today represent 75% of the world energy production. Moreover, their use will have caused large-scale emission of greenhouse gases (GEG), and induced global climate change. This problem is exacerbated by a growing world energy demand. In this context, nuclear power is the only GEG-free energy source available today capable of responding significantly to this demand. Some scenarios consider a nuclear energy production of around 5 Gtoe in 2050, wich would represent a 20% share of the world energy supply. Present reactors generate energy from the fission of U-235 and require around 200 tons of natural Uranium to produce 1GWe.y of energy, equivalent to the fission of one ton of fissile material. In a scenario of a significant increase in nuclear energy generation, these standard reactors will consume the whole of the world's estimated Uranium reserves in a few decades. However, natural Uranium or Thorium ore, wich are not themselves fissile, can produce a fissile material after a neutron capture ( 239Pu and 233U respectively). In a breeder reactor, the mass of fissile material remains constant, and the fertile ore is the only material to be consumed. In this case, only 1 ton of natural ore is needed to produce 1GWe.y. Thus, the breeding concept allows optimal use of fertile ore and development of sustainable nuclear energy production for several thousand years into the future. Different sustainable nuclear reactor concepts are studied in the international forum "generation IV". Different types of coolant (Na, Pb and He) are studied for fast breeder reactors based on the Uranium cycle. The thermal Thorium cycle requires the use of a liquid fuel, which can be reprocessed online in order to extract the neutron poisons. This paper presents these different sustainable reactors, based on the Uranium or Thorium fuel cycles and will compare the different options in term of fissile

  9. Interplanetary Type IV Bursts

    NASA Astrophysics Data System (ADS)

    Hillaris, A.; Bouratzis, C.; Nindos, A.

    2016-08-01

    We study the characteristics of moving type IV radio bursts that extend to hectometric wavelengths (interplanetary type IV or type {IV}_{{IP}} bursts) and their relationship with energetic phenomena on the Sun. Our dataset comprises 48 interplanetary type IV bursts observed with the Radio and Plasma Wave Investigation (WAVES) instrument onboard Wind in the 13.825 MHz - 20 kHz frequency range. The dynamic spectra of the Radio Solar Telescope Network (RSTN), the Nançay Decametric Array (DAM), the Appareil de Routine pour le Traitement et l' Enregistrement Magnetique de l' Information Spectral (ARTEMIS-IV), the Culgoora, Hiraso, and the Institute of Terrestrial Magnetism, Ionosphere and Radio Wave Propagation (IZMIRAN) Radio Spectrographs were used to track the evolution of the events in the low corona. These were supplemented with soft X-ray (SXR) flux-measurements from the Geostationary Operational Environmental Satellite (GOES) and coronal mass ejections (CME) data from the Large Angle and Spectroscopic Coronagraph (LASCO) onboard the Solar and Heliospheric Observatory (SOHO). Positional information of the coronal bursts was obtained by the Nançay Radioheliograph (NRH). We examined the relationship of the type IV events with coronal radio bursts, CMEs, and SXR flares. The majority of the events (45) were characterized as compact, their duration was on average 106 minutes. This type of events was, mostly, associated with M- and X-class flares (40 out of 45) and fast CMEs, 32 of these events had CMEs faster than 1000 km s^{-1}. Furthermore, in 43 compact events the CME was possibly subjected to reduced aerodynamic drag as it was propagating in the wake of a previous CME. A minority (three) of long-lived type {IV}_{{IP}} bursts was detected, with durations from 960 minutes to 115 hours. These events are referred to as extended or long duration and appear to replenish their energetic electron content, possibly from electrons escaping from the corresponding coronal

  10. Roles of individual radicals generated by a submerged dielectric barrier discharge plasma reactor during Escherichia coli O157:H7 inactivation

    NASA Astrophysics Data System (ADS)

    Khan, Muhammad Saiful Islam; Lee, Eun-Jung; Kim, Yun-Ji

    2015-10-01

    A submerged dielectric barrier discharge plasma reactor (underwater DBD) has been used on Escherichia coli O157:H7 (ATCC 35150). Plasma treatment was carried out using clean dry air gas to investigate the individual effects of the radicals produced by underwater DBD on an E. coli O157:H7 suspension (8.0 log CFU/ml). E. coli O157:H7 was reduced by 6.0 log CFU/ml for 2 min of underwater DBD plasma treatment. Optical Emission Spectra (OES) shows that OH and NO (α, β) radicals, generated by underwater DBD along with ozone gas. E. coli O157:H7 were reduced by 2.3 log CFU/ml for 10 min of underwater DBD plasma treatment with the terephthalic acid (TA) OH radical scavenger solution, which is significantly lower (3.7 log CFU/ml) than the result obtained without using the OH radical scavenger. A maximum of 1.5 ppm of ozone gas was produced during the discharge of underwater DBD, and the obtained reduction difference in E.coli O157:H7 in presence and in absence of ozone gas was 1.68 log CFU/ml. The remainder of the 0.62 log CFU/ml reduction might be due to the effect of the NO (α, β) radicals or due to the combined effect of all the radicals produced by underwater DBD. A small amount of hydrogen peroxide was also generated but does not play any role in E. coli O157:H7 inactivation.

  11. Roles of individual radicals generated by a submerged dielectric barrier discharge plasma reactor during Escherichia coli O157:H7 inactivation

    SciTech Connect

    Khan, Muhammad Saiful Islam; Lee, Eun-Jung; Kim, Yun-Ji

    2015-10-15

    A submerged dielectric barrier discharge plasma reactor (underwater DBD) has been used on Escherichia coli O157:H7 (ATCC 35150). Plasma treatment was carried out using clean dry air gas to investigate the individual effects of the radicals produced by underwater DBD on an E. coli O157:H7 suspension (8.0 log CFU/ml). E. coli O157:H7 was reduced by 6.0 log CFU/ml for 2 min of underwater DBD plasma treatment. Optical Emission Spectra (OES) shows that OH and NO (α, β) radicals, generated by underwater DBD along with ozone gas. E. coli O157:H7 were reduced by 2.3 log CFU/ml for 10 min of underwater DBD plasma treatment with the terephthalic acid (TA) OH radical scavenger solution, which is significantly lower (3.7 log CFU/ml) than the result obtained without using the OH radical scavenger. A maximum of 1.5 ppm of ozone gas was produced during the discharge of underwater DBD, and the obtained reduction difference in E.coli O157:H7 in presence and in absence of ozone gas was 1.68 log CFU/ml. The remainder of the 0.62 log CFU/ml reduction might be due to the effect of the NO (α, β) radicals or due to the combined effect of all the radicals produced by underwater DBD. A small amount of hydrogen peroxide was also generated but does not play any role in E. coli O157:H7 inactivation.

  12. Gas dynamics and heat transfer in a packed pebble-bed reactor for the 4th generation nuclear energy

    NASA Astrophysics Data System (ADS)

    Abdulmohsin, Rahman

    -AlOx-Al junctions, we show that, despite excellent temperature stability, temperature fluctuations induce observable critical current fluctuations. Particularly, becuase 1/ f critical current noise has decreased with improved fabrication techniques in recent years, it is important to understand and eliminate this additional noise source. Next, we introduce a numerical method of calculating the mean square flux noise F2 from independently fluctuating spins on the surface of thin-film loops of arbitrary geometry. By reciprocity, F2 is proportional to Br2 , where B(r) is the magnetic field generated by a circulating current around the loop and r varies over the loop surface. By discretizing the loop nonuniformly, we efficiently and accurately compute the current distribution and resulting magnetic field, which may vary rapidly across the loop. We use this method to compute F2 in a number of scenarios in which we systematically vary physical parameters of the loop. We compare our simulations to an earlier analytic result predicting that F2 ∝ R/W in the limit where the loop radius R is much greater than the linewidth W. We further show that the previously neglected contribution of edge spins to F2 is significant---even dominant---in narrow-linewidth loops. To calculate theoretical dephasing rates in qubits, we consider flux noise with a spectral density Sphi( f) = A2/ (f/1 Hz) alpha, where A is of the order of 1 muphi 0 Hz--1/2 and 0.6 ≤ alpha ≤ 1.2; applied flux, our calculations of the dependence of the pure dephasing time tau φ Ramsey and echo pulse sequences on alpha for fixed A show that tauφ decreases rapidly as alpha is reduced. We find that tauφ is relatively insensitive to the noise bandwidth, f1 ≤ f ≤ f2 for all alpha provided the ultraviolet cutoff frequency f2 > 1/tauφ. We calculate the ratio tauφ,E/tau φ, R of the echo (E) and Ramsey (R) sequences, and the dependence of the decay function on alpha and f2. We investigate the case in which S phi(f0) is fixed

  13. Zirconium carbide coating for corium experiments related to water-cooled and sodium-cooled reactors

    NASA Astrophysics Data System (ADS)

    Plevacova, K.; Journeau, C.; Piluso, P.; Zhdanov, V.; Baklanov, V.; Poirier, J.

    2011-07-01

    Since the TMI and Chernobyl accidents the risk of nuclear severe accident is intensively studied for existing and future reactors. In case of a core melt-down accident in a nuclear reactor, a complex melt, called corium, forms. To be able to perform experiments with prototypic corium materials at high temperature, a coating which resists to different corium melts related to Generation I and II Water Reactors and Generation IV sodium fast reactor was researched in our experimental platforms both in IAE NNC in Kazakhstan and in CEA in France. Zirconium carbide was selected as protective coating for graphite crucibles used in our induction furnaces: VCG-135 and VITI. The method of coating application, called reactive wetting, was developed. Zirconium carbide revealed to resist well to the (U x, Zr y)O 2-z water reactor corium. It has also the advantage not to bring new elements to this chemical system. The coating was then tested with sodium fast reactor corium melts containing steel or absorbers. Undesirable interactions were observed between the coating and these materials, leading to the carburization of the corium ingots. Concerning the resistance of the coating to oxide melts without ZrO 2, the zirconium carbide coating keeps its role of protective barrier with UO 2-Al 2O 3 below 2000 °C but does not resist to a UO 2-Eu 2O 3 mixture.

  14. Preparation of cyclotron-produced 186Re and comparison with reactor-produced 186Re and generator-produced 188Re for the labeling of bombesin.

    PubMed

    Moustapha, Moustapha E; Ehrhardt, Gary J; Smith, Charles J; Szajek, Lawrence P; Eckelman, William C; Jurisson, Silvia S

    2006-01-01

    The radioisotopes (186)Re and (188)Re have been extensively investigated for various forms of radiotherapy due to their useful and high-abundance beta particle emissions, low-abundance and imageable gamma-rays, and chemical resemblance to technetium. In addition, (188)Re is available in no-carrier-added (NCA) form from long lived W-188 generators, whereas (186)Re can be produced in large quantities from reactors, although not in NCA form. However, NCA (186)Re can be produced on a cyclotron by a (p,n) reaction on (186)W. The purpose of this study was to compare labeling of the peptide bombesin with these three forms of rhenium radioisotopes. Cyclotron-produced NCA (186)Re was separated radiochemically from enriched (186)W (96.9%) targets using high-purity methyl ethyl ketone (MEK). The resulting (186)Re-MEK was then loaded onto a small alumina column to separate the resulting NCA (186)Re from any remaining (186)W. The experimental levels of impurities associated with (186)Re at the end of the separation process were found to be 5.7 x 10(-6) Ci of (182)Re (0.57%, t(1/2) = 12.7 h) and 1.283 x 10(-5) Ci of (182m)Re (1.28%, t(1/2) = 2.67 days). The radionuclidic purity of the separated (186)Re was found to be 99.6%, whereas the chemical identity was determined by reversed phase high-performance liquid chromatography (RP-HPLC) to be perrhenate ((186)ReO(4)(-)). Generator-produced (188)ReO(4)(-) from a (188)W/(188)Re generator (Oak Ridge National Laboratory) and CA (186)ReO(4)(-) produced from a (185)Re(n,gamma)(186)Re reaction at the University of Missouri Research Reactor (MURR) were used for comparison with the NCA (186)Re in subsequent studies. N(3)S-5-Ava-BBN(7-14)NH(2) conjugates provide flexibility for designing (186,188)Re-labeled conjugates that retain high in vitro and in vivo specificity targeting of GRP receptor-expressing cells. This study showed that the N(3)S-5-Ava-BBN(7-14)NH(2) could be labeled with (186,188)Re following the preconjugation

  15. NEUTRONIC REACTOR

    DOEpatents

    Daniels, F.

    1962-12-18

    A power plant is described comprising a turbine and employing round cylindrical fuel rods formed of BeO and UO/sub 2/ and stacks of hexagonal moderator blocks of BeO provided with passages that loosely receive the fuel rods so that coolant may flow through the passages over the fuels to remove heat. The coolant may be helium or steam and fiows through at least one more heat exchanger for producing vapor from a body of fluid separate from the coolant, which fluid is to drive the turbine for generating electricity. By this arrangement the turbine and directly associated parts are free of particles and radiations emanating from the reactor. (AEC)

  16. Treatment of industrial effluents by electrochemical generation of H2O2 using an RVC cathode in a parallel plate reactor.

    PubMed

    Bustos, Yaneth A; Rangel-Peraza, Jesús Gabriel; Rojas-Valencia, Ma Neftalí; Bandala, Erick R; Álvarez-Gallegos, Alberto; Vargas-Estrada, Laura

    2016-01-01

    Electrochemical techniques have been used for the discolouration of synthetic textile industrial wastewater by Fenton's process using a parallel plate reactor with a reticulated vitreous carbon (RVC) cathode. It has been shown that RVC is capable of electro-generating and activating H2O2 in the presence of Fe(2+) added as catalyst and using a stainless steel mesh as anode material. A catholyte comprising 0.05 M Na2SO4, 0.001 M FeSO4.7H2O, 0.01 M H2SO4 and fed with oxygen was used to activate H2O2.The anolyte contained only 0.8 M H2SO4. The operating experimental conditions were 170 mA (2.0 V < ΔECell < 3.0 V) to generate 5.3 mM H2O2. Synthetic effluents containing various concentrations (millimolar - mM) of three different dyes, Blue Basic 9 (BB9), Reactive Black 5 (RB5) and Acid Orange 7 (AO7), were evaluated for discolouration using the electro-assisted Fenton reaction. Water discolouration was measured by UV-VIS absorbance reduction. Dye removal by electrolysis was a function of time: 90% discolouration of 0.08, 0.04 and 0.02 mM BB9 was obtained at 14, 10 and 6 min, respectively. In the same way, 90% discolouration of 0.063, 0.031 and 0.016 mM RB5 was achieved at 90, 60 and 30 min, respectively. Finally, 90% discolouration of 0.14, 0.07 and 0.035 mM AO7 was achieved at 70, 40 and 20 min, respectively. The experimental results confirmed the effectiveness of electro-assisted Fenton reaction as a strong oxidizing process in water discolouration and the ability of RVC cathode to electro-generate and activate H2O2 in situ.

  17. Treatment of industrial effluents by electrochemical generation of H2O2 using an RVC cathode in a parallel plate reactor.

    PubMed

    Bustos, Yaneth A; Rangel-Peraza, Jesús Gabriel; Rojas-Valencia, Ma Neftalí; Bandala, Erick R; Álvarez-Gallegos, Alberto; Vargas-Estrada, Laura

    2016-01-01

    Electrochemical techniques have been used for the discolouration of synthetic textile industrial wastewater by Fenton's process using a parallel plate reactor with a reticulated vitreous carbon (RVC) cathode. It has been shown that RVC is capable of electro-generating and activating H2O2 in the presence of Fe(2+) added as catalyst and using a stainless steel mesh as anode material. A catholyte comprising 0.05 M Na2SO4, 0.001 M FeSO4.7H2O, 0.01 M H2SO4 and fed with oxygen was used to activate H2O2.The anolyte contained only 0.8 M H2SO4. The operating experimental conditions were 170 mA (2.0 V < ΔECell < 3.0 V) to generate 5.3 mM H2O2. Synthetic effluents containing various concentrations (millimolar - mM) of three different dyes, Blue Basic 9 (BB9), Reactive Black 5 (RB5) and Acid Orange 7 (AO7), were evaluated for discolouration using the electro-assisted Fenton reaction. Water discolouration was measured by UV-VIS absorbance reduction. Dye removal by electrolysis was a function of time: 90% discolouration of 0.08, 0.04 and 0.02 mM BB9 was obtained at 14, 10 and 6 min, respectively. In the same way, 90% discolouration of 0.063, 0.031 and 0.016 mM RB5 was achieved at 90, 60 and 30 min, respectively. Finally, 90% discolouration of 0.14, 0.07 and 0.035 mM AO7 was achieved at 70, 40 and 20 min, respectively. The experimental results confirmed the effectiveness of electro-assisted Fenton reaction as a strong oxidizing process in water discolouration and the ability of RVC cathode to electro-generate and activate H2O2 in situ. PMID:26419746

  18. PLATO IV Accountancy Index.

    ERIC Educational Resources Information Center

    Pondy, Dorothy, Comp.

    The catalog was compiled to assist instructors in planning community college and university curricula using the 48 computer-assisted accountancy lessons available on PLATO IV (Programmed Logic for Automatic Teaching Operation) for first semester accounting courses. It contains information on lesson access, lists of acceptable abbreviations for…

  19. The PLATO IV Architecture.

    ERIC Educational Resources Information Center

    Stifle, Jack

    The PLATO IV computer-based instructional system consists of a large scale centrally located CDC 6400 computer and a large number of remote student terminals. This is a brief and general description of the proposed input/output hardware necessary to interface the student terminals with the computer's central processing unit (CPU) using available…

  20. IVS Technology Coordinator Report

    NASA Technical Reports Server (NTRS)

    Whitney, Alan

    2013-01-01

    This report of the Technology Coordinator includes the following: 1) continued work to implement the new VLBI2010 system, 2) the 1st International VLBI Technology Workshop, 3) a VLBI Digital- Backend Intercomparison Workshop, 4) DiFX software correlator development for geodetic VLBI, 5) a review of progress towards global VLBI standards, and 6) a welcome to new IVS Technology Coordinator Bill Petrachenko.

  1. Future reactor experiments

    SciTech Connect

    Wen, Liangjian

    2015-07-15

    The non-zero neutrino mixing angle θ{sub 13} has been discovered and precisely measured by the current generation short-baseline reactor neutrino experiments. It opens the gate of measuring the leptonic CP-violating phase and enables the neutrino mass ordering. The JUNO and RENO-50 proposals aim at resolving the neutrino mass ordering using reactors. The experiment design, physics sensitivity, technical challenges as well as the progresses of those two proposed experiments are reviewed in this paper.

  2. GIF sodium fast reactor project R and D on safety and operation

    SciTech Connect

    Vasile, A.; Sofu, T.; Jeong, H. Y.; Sakai, T.

    2012-07-01

    The 'Safety and Operation' project is started in 2009 within the framework of Generation-IV International Forum (GIF) Sodium Fast Reactor (SFR) research and development program. In the safety area, the project involves R and D activities on phenomenological model development and experimental programs, conceptual studies in support of the design of safety provisions, preliminary assessment of safety systems, framework and methods for analysis of safety architecture. In the operation area, the project involves R and D activities on fast reactors safety tests and analysis of reactor operations, feedback from decommissioning, in-service inspection technique development, under-sodium viewing and sodium chemistry. This paper presents a summary of such activities and the main achievements. (authors)

  3. R and D program for core instrumentation improvements devoted for French sodium fast reactors

    SciTech Connect

    Jeannot, J. P.; Rodriguez, G.; Jammes, C.; Bernardin, B.; Portier, J. L.; Jadot, F.; Maire, S.; Verrier, D.; Loisy, F.; Prele, G.

    2011-07-01

    Under the framework of French R and D studies for Generation IV reactors and more specifically for sodium-cooled fast reactors (SFR); the CEA, EDF and AREVA have launched a joint coordinated research programme. This paper deals with the R and D sets out to achieve better inspection, maintenance, availability and decommissioning. In particular the instrumentation requirements for core monitoring and detection in the case of accidental events. Requirements mainly involve diversifying the means of protection and improving instrumentation performance in terms of responsiveness and sensitivity. Operation feedback from the Phenix and Superphenix prototype reactors and studies, carried out within the scope of the EFR projects, has been used to define the needs for instrumentation enhancement. (authors)

  4. Reducing Nitrogen Oxide Emissions: 1996 Compliance with Title IV Limits

    EIA Publications

    1998-01-01

    The purpose of this article is to summarize the existing federal nitrogen oxide (Nox) regulations and the 1996 performance of the 239 Title IV generating units. It also reviews the basics of low-Nox burner technology and presents cost and performance data for retrofits at Title IV units.

  5. NUCLEAR REACTOR

    DOEpatents

    Treshow, M.

    1961-09-01

    A boiling-water nuclear reactor is described wherein control is effected by varying the moderator-to-fuel ratio in the reactor core. This is accomplished by providing control tubes containing a liquid control moderator in the reactor core and providing means for varying the amount of control moderatcr within the control tubes.

  6. NEUTRONIC REACTOR

    DOEpatents

    Daniels, F.

    1959-10-27

    A reactor in which at least a portion of the moderator is in the form of movable refractory balls is described. In addition to their moderating capacity, these balls may serve as carriers for fissionable material or fertile material, or may serve in a coolant capacity to remove heat from the reactor. A pneumatic system is used to circulate the balls through the reactor.

  7. Research reactors

    SciTech Connect

    Tonneson, L.C.; Fox, G.J.

    1996-04-01

    There are currently 284 research reactors in operation, and 12 under construction around the world. Of the operating reactors, nearly two-thirds are used exclusively for research, and the rest for a variety of purposes, including training, testing, and critical assembly. For more than 50 years, research reactor programs have contributed greatly to the scientific and educational communities. Today, six of the world`s research reactors are being shut down, three of which are in the USA. With government budget constraints and the growing proliferation concerns surrounding the use of highly enriched uranium in some of these reactors, the future of nuclear research could be impacted.

  8. CONVECTION REACTOR

    DOEpatents

    Hammond, R.P.; King, L.D.P.

    1960-03-22

    An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

  9. Neutron Cross Section Covariances: Recent Workshop and Advanced Reactor Systems

    NASA Astrophysics Data System (ADS)

    Oblozinsky, Pavel

    2008-10-01

    The recent Workshop on Neutron Cross Section Covariances, organized by BNL and attended by more than 50 scientists, responded to demands of many user groups, including advanced reactor systems, for uncertainty and correlation information. These demands can be explained by considerable progress in advanced neutronics simulation that probe covariances and their impact on design and operational margins of nuclear systems. The Workshop addressed evaluation methodology, recent evaluations as well as user's perspective, marking era of revival of covariance development that started some two years ago. We illustrate urgent demand for covariances in the case of advanced reactor systems, including fast actinide burner under GNEP, new generation of power reactors, Gen-IV, and reactors under AFCI. A common feature of many of these systems is presence of large amount of minor actinides and fission products that require improved nuclear data. Advanced simulation codes rely on quality input, to be obtained by adjusting the data library, such as the new ENDF/B-VII.0, by considering integral experiments as currently pursued by GNEP. To this end the nuclear data community is developing covariances for formidable amount of 112 materials (isotopes).

  10. Development of Improved Burnable Poisons for Commercial Nuclear Power Reactors

    SciTech Connect

    M. L. Grossbeck J-P.A. Renier Tim Bigelow

    2003-09-30

    Burnable poisons are used in nuclear reactors to produce a more level distribution of power in the reactor core and to reduce to necessity for a large control system. An ideal burnable poison would burn at the same rate as the fuel. In this study, separation of neutron-absorbing isotopes was investigated in order to eliminate isotopes that remain as absorbers at the end of fuel life, thus reducing useful fuel life. The isotopes Gd-157, Dy-164, and Er-167 were found to have desirable properties. These isotopes were separated from naturally occurring elements by means of plasma separation to evaluate feasibility and cost. It was found that pure Gd-157 could save approximately $6 million at the end of four years. However, the cost of separation, using the existing facility, made separation cost- ineffective. Using a magnet with three times the field strength is expected to reduce the cost by a factor of ten, making isotopically separated burnable poisons a favorable method of increasing fuel life in commercial reactors, in particular Generation-IV reactors. The project also investigated various burnable poison configurations, and studied incorporation of metallic burnable poisons into fuel cladding.

  11. Small modular reactor (SMR) development plan in Korea

    SciTech Connect

    Shin, Yong-Hoon Park, Sangrok; Kim, Byong Sup; Choi, Swongho; Hwang, Il Soon

    2015-04-29

    Since the first nuclear power was engaged in Korean electricity grid in 1978, intensive research and development has been focused on localization and standardization of large pressurized water reactors (PWRs) aiming at providing Korean peninsula and beyond with economical and safe power source. With increased priority placed on the safety since Chernobyl accident, Korean nuclear power R and D activity has been diversified into advanced PWR, small modular PWR and generation IV reactors. After the outbreak of Fukushima accident, inherently safe small modular reactor (SMR) receives growing interest in Korea and Europe. In this paper, we will describe recent status of evolving designs of SMR, their advantages and challenges. In particular, the conceptual design of lead-bismuth cooled SMR in Korea, URANUS with 40∼70 MWe is examined in detail. This paper will cover a framework of the program and a strategy for the successful deployment of small modular reactor how the goals would entail and the approach to collaboration with other entities.

  12. Small modular reactor (SMR) development plan in Korea

    NASA Astrophysics Data System (ADS)

    Shin, Yong-Hoon; Park, Sangrok; Kim, Byong Sup; Choi, Swongho; Hwang, Il Soon

    2015-04-01

    Since the first nuclear power was engaged in Korean electricity grid in 1978, intensive research and development has been focused on localization and standardization of large pressurized water reactors (PWRs) aiming at providing Korean peninsula and beyond with economical and safe power source. With increased priority placed on the safety since Chernobyl accident, Korean nuclear power R&D activity has been diversified into advanced PWR, small modular PWR and generation IV reactors. After the outbreak of Fukushima accident, inherently safe small modular reactor (SMR) receives growing interest in Korea and Europe. In this paper, we will describe recent status of evolving designs of SMR, their advantages and challenges. In particular, the conceptual design of lead-bismuth cooled SMR in Korea, URANUS with 40˜70 MWe is examined in detail. This paper will cover a framework of the program and a strategy for the successful deployment of small modular reactor how the goals would entail and the approach to collaboration with other entities.

  13. Enhanced Design Alternative IV

    SciTech Connect

    N. E. Kramer

    1999-05-18

    This report evaluates Enhanced Design Alternative (EDA) IV as part of the second phase of the License Application Design Selection (LADS) effort. The EDA IV concept was compared to the VA reference design using criteria from the ''Design Input Request for LADS Phase II EDA Evaluations'' (CRWMS M&O 1999b) and (CRWMS M&O 1999f). Briefly, the EDA IV concept arranges the waste packages close together in an emplacement configuration known as ''line load''. Continuous pre-closure ventilation keeps the waste packages from exceeding the 350 C cladding and 200 C (4.3.13) drift wall temperature limits. This EDA concept keeps relatively high, uniform emplacement drift temperatures (post-closure) to drive water away from the repository and thus dry out the pillars between emplacement drifts. The waste package is shielded to permit human access to emplacement drifts and includes an integral filler inside the package to reduce the amount of water that can contact the waste form. Closure of the repository is desired 50 years after first waste is emplaced. Both backfill and a drip shields will be emplaced at closure to improve post-closure performance.

  14. Reactor Safety Research Programs

    SciTech Connect

    Edler, S. K.

    1981-07-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  15. A sputnik IV saga

    NASA Astrophysics Data System (ADS)

    Lundquist, Charles A.

    2009-12-01

    The Sputnik IV launch occurred on May 15, 1960. On May 19, an attempt to deorbit a 'space cabin' failed and the cabin went into a higher orbit. The orbit of the cabin was monitored and Moonwatch volunteer satellite tracking teams were alerted to watch for the vehicle demise. On September 5, 1962, several team members from Milwaukee, Wisconsin made observations starting at 4:49 a.m. of a fireball following the predicted orbit of Sputnik IV. Requests went out to report any objects found under the fireball path. An early morning police patrol in Manitowoc had noticed a metal object on a street and had moved it to the curb. Later the officers recovered the object and had it dropped off at the Milwaukee Journal. The Moonwarch team got the object and reported the situation to Moonwatch Headquarters at the Smithsonian Astrophysical Observatory. A team member flew to Cambridge with the object. It was a solid, 9.49 kg piece of steel with a slag-like layer attached to it. Subsequent analyses showed that it contained radioactive nuclei produced by cosmic ray exposure in space. The scientists at the Observatory quickly recognized that measurements of its induced radioactivity could serve as a calibration for similar measurements of recently fallen nickel-iron meteorites. Concurrently, the Observatory directorate informed government agencies that a fragment from Sputnik IV had been recovered. Coincidently, a debate in the UN Committee on Peaceful Uses of Outer Space involved the issue of liability for damage caused by falling satellite fragments. On September 12, the Observatory delivered the bulk of the fragment to the US Delegation to the UN. Two days later, the fragment was used by US Ambassador Francis Plimpton as an exhibit that the time had come to agree on liability for damage from satellite debris. He offered the Sputnik IV fragment to USSR Ambassador P.D. Morozov, who refused the offer. On October 23, Drs. Alla Massevitch and E.K. Federov of the USSR visited the

  16. Development of a new non-aeration-based sewage treatment technology: Performance evaluation of a full-scale down-flow hanging sponge reactor employing third-generation sponge carriers.

    PubMed

    Okubo, Tsutomu; Kubota, Kengo; Yamaguchi, Takashi; Uemura, Shigeki; Harada, Hideki

    2016-10-01

    A practical-scale down-flow hanging sponge (DHS) reactor using third-generation (G3) sponge carriers was applied for treatment of the effluent from an up-flow anaerobic sludge blanket (UASB) reactor treating municipal sewage. The process performance of the DHS reactor filled with G3 sponge carriers (DHS-G3) was evaluated by conducting an on-site experiment in India over one year. The performance of the DHS-G3 for removal of organic matter and ammonium-nitrogen at a relatively short hydraulic retention time (HRT) of only 0.66 h satisfied the Indian effluent quality standards except for fecal coliform. The removal rate constants for total biochemical oxygen demand (BOD) and fecal coliform determined based on the water quality profiles along the DHS-G3 almost reached equilibrium approximately four months after the start of operation, i.e., 2.45 h(-1) for BOD and 2.30 h(-1) for fecal coliform, respectively. The oxygen utilization activity of retained sludge was determined to assess the distribution of heterotrophic and autotrophic bacteria along the DHS-G3. Nitrification was promoted in the lower portion of the DHS-G3 reactor in the duration with low organic load, while it decreased when the organic load was increased, probably due to proliferation of heterotrophic bacteria. PMID:27340815

  17. R and D of On-line Reprocessing Technology for Molten-Salt Reactor Systems

    SciTech Connect

    Uhlir, Jan; Tulackova, Radka; Chuchvalcova Bimova, Karolina

    2006-07-01

    The Molten Salt Reactor (MSR) represents one of promising future nuclear reactor concept included in the Generation IV reactors family. The reactor can be operated as the thorium breeder or as the actinide transmuter. However, the future deployment of Molten-Salt Reactors will be significantly dependent on the successful mastering of advanced reprocessing technologies dedicated to their fuel cycle. Here the on-line reprocessing technology connected with the fuel circuit of MSR is of special importance because the reactor cannot be operated for a long run without the fuel salt clean-up. Generally, main MSR reprocessing technologies are pyrochemical, majority of them are fluoride technologies. The proposed flow-sheets of MSR on-line reprocessing are based on a combination of molten-salt / liquid metal extraction and electro-separation processes, which can be added to the gas extraction process already verified during the MSRE project in ORNL. The crucial separation method proposed for partitioning of actinides from fission products is based on successive Anodic dissolution and Cathodic deposition processes in molten fluoride media. (authors)

  18. Insights from the WGRISK workshop on the PSA of advanced and new reactors

    SciTech Connect

    Georgescu, G.; Ahn, K. I.; Amri, A.

    2012-07-01

    Probabilistic Safety Assessment /Probabilistic Risk Assessment for new and advanced reactors is recognized as an essential complement of the deterministic approaches to achieve improved safety and performances of new nuclear power plants, comparing to the operating plants. However, the development of PSA to these reactors is encountered to concurrent challenges, mainly due to the limited available design information, as well as due to potentially new initiating events, accident sequences and phenomena. The use of PSA in the decision making process is also challenging since the resulting PSA may not sufficiently reflect the future as-built, as-operated plant information. In order to address these aspects, the OECD/NEA/WGRISK initiated two coordinated tasks on 'PSA for Advanced Reactors' and 'PSA in the frame of Design and Commissioning of New NPPs'. In this context, a joint workshop was organized by OECD, during which related subjects were presented and discussed, including PSA for generation IV reactors, PSA for evolutionary reactors, PSA for small modular reactors, severe accidents and Level 2 PSA, Level 3 PSA and consequences analysis, digital I and C modeling, passive systems reliability, safety-security interface, as well as the results of the surveys performed in the frame of theses WGRISK tasks. (authors)

  19. Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors

    SciTech Connect

    Shropshire, D.E.; Herring, J.S.

    2004-10-03

    The objective of this paper is to facilitate a better understanding of the fuel-cycle and nuclear material disposition issues associated with high-temperature gas reactors (HTGRs). This paper reviews the nuclear fuel cycles supporting early and present day gas reactors, and identifies challenges for the advanced fuel cycles and waste management systems supporting the next generation of HTGRs, including the Very High Temperature Reactor, which is under development in the Generation IV Program. The earliest gas-cooled reactors were the carbon dioxide (CO2)-cooled reactors. Historical experience is available from over 1,000 reactor-years of operation from 52 electricity-generating, CO2-cooled reactor plants that were placed in operation worldwide. Following the CO2 reactor development, seven HTGR plants were built and operated. The HTGR came about from the combination of helium coolant and graphite moderator. Helium was used instead of air or CO2 as the coolant. The helium gas has a significant technical base due to the experience gained in the United States from the 40-MWe Peach Bottom and 330-MWe Fort St. Vrain reactors designed by General Atomics. Germany also built and operated the 15-MWe Arbeitsgemeinschaft Versuchsreaktor (AVR) and the 300-MWe Thorium High-Temperature Reactor (THTR) power plants. The AVR, THTR, Peach Bottom and Fort St. Vrain all used fuel containing thorium in various forms (i.e., carbides, oxides, thorium particles) and mixtures with highly enriched uranium. The operational experience gained from these early gas reactors can be applied to the next generation of nuclear power systems. HTGR systems are being developed in South Africa, China, Japan, the United States, and Russia. Elements of the HTGR system evaluated included fuel demands on uranium ore mining and milling, conversion, enrichment services, and fuel fabrication; fuel management in-core; spent fuel characteristics affecting fuel recycling and refabrication, fuel handling, interim

  20. Testing and economical evaluation of U(IV) in Purex

    SciTech Connect

    Hoisington, J.E.; Hsu, T.C.

    1983-01-01

    The use of uranous nitrate, U(IV), as a plutonium reductant in the Purex solvent extraction process could significantly reduce the waste generation at the Savannah River Plant. The current reductant is a ferrous sulfamate (FS)/hydroxylamine nitrate (HAN) mixture. The iron and sulfate in the FS are major contributors to waste generation. The U(IV) reductant oxidizes to U(VI) producing no waste. The Savannah River Laboratory has developed an efficient electrochemical cell for U(IV) production and has demonstrated the effectiveness of U(IV) as a plutonium reductant. Plant tests and economic analyses are currently being conducted to determine the cost effectiveness of U(IV) implementation. The results of recent studies are presented.

  1. Thermal-hydraulics of internally heated molten salts and application to the Molten Salt Fast Reactor

    NASA Astrophysics Data System (ADS)

    Fiorina, Carlo; Cammi, Antonio; Luzzi, Lelio; Mikityuk, Konstantin; Ninokata, Hisashi; Ricotti, Marco E.

    2014-04-01

    The Molten Salt Reactors (MSR) are an innovative kind of nuclear reactors and are presently considered in the framework of the Generation IV International Forum (GIF-IV) for their promising performances in terms of low resource utilization, waste minimization and enhanced safety. A unique feature of MSRs is that molten fluoride salts play the distinctive role of both fuel (heat source) and coolant. The presence of an internal heat generation perturbs the temperature field and consequences are to be expected on the heat transfer characteristics of the molten salts. In this paper, the problem of heat transfer for internally heated fluids in a straight circular channel is first faced on a theoretical ground. The effect of internal heat generation is demonstrated to be described by a corrective factor applied to traditional correlations for the Nusselt number. It is shown that the corrective factor can be fully characterized by making explicit the dependency on Reynolds and Prandtl numbers. On this basis, a preliminary correlation is proposed for the case of molten fluoride salts by interpolating the results provided by an analytic approach previously developed at the Politecnico di Milano. The experimental facility and the related measuring procedure for testing the proposed correlation are then presented. Finally, the developed correlation is used to carry out a parametric investigation on the effect of internal heat generation on the main out-of-core components of the Molten Salt Fast Reactor (MSFR), the reference circulating-fuel MSR design in the GIF-IV. The volumetric power determines higher temperatures at the channel wall, but the effect is significant only in case of large diameters and/or low velocities.

  2. RAPHAEL: The European Union's (Very) High Temperature Reactor Technology Project

    SciTech Connect

    Fuetterer, Michael A.; Besson, D.; Bogusch, E.; Carluec, B.; Hittner, D.; Verrier, D.; Billot, Ph.; Phelip, M.; Buckthorpe, D.; Casalta, S.; Chauvet, V.; Van Heek, A.; Von Lensa, W.; Pirson, J.; Scheuermann, W.

    2006-07-01

    Since the late 1990, the European Union (EU) was conducting work on High Temperature Reactors (HTR) confirming their high potential in terms of safety (inherent safety features), environmental impact (robust fuel with no significant radioactive release), sustainability (high efficiency, potential suitability for various fuel cycles), and economics (simplifications arising from safety features). In April 2005, the EU Commission has started a new 4-year Integrated Project on Very High Temperature Reactors (RAPHAEL: Reactor for Process Heat And Electricity) as part of its 6{sup th} Framework Programme. The European Commission and the 33 partners from industry, R and D organizations and academia finance the project together. After the successful performance of earlier HTR-related EU projects which included the recovery of some earlier German experience and the re-establishment of strategically important R and D capabilities in Europe, RAPHAEL focuses now on key technologies required for an industrial VHTR deployment, both specific to very high temperature and generic to all types of modular HTR with emphasis on combined process heat and electricity generation. Advanced technologies are explored in order to meet the performance challenges required for a VHTR (900-1000 deg C, up to 200 GWd/tHM). To facilitate the planned sharing of significant parts of RAPHAEL results with the signatories of the Generation IV International Forum (GIF) VHTR projects, RAPHAEL is structured in a similar way as the corresponding GIF VHTR projects. (authors)

  3. REACTOR GROUT THERMAL PROPERTIES

    SciTech Connect

    Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.

    2011-01-28

    Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.

  4. PMD IVS Analysis Center

    NASA Technical Reports Server (NTRS)

    Tornatore, Vincenza

    2013-01-01

    The main activities carried out at the PMD (Politecnico di Milano DIIAR) IVS Analysis Center during 2012 are briefly higlighted, and future plans for 2013 are sketched out. We principally continued to process European VLBI sessions using different approaches to evaluate possible differences due to various processing choices. Then VLBI solutions were also compared to the GPS ones as well as the ones calculated at co-located sites. Concerning the observational aspect, several tests were performed to identify the most suitable method to achieve the highest possible accuracy in the determination of GNSS (GLOBAL NAVIGATION SATELLITE SYSTEM) satellite positions using the VLBI technique.

  5. Plant maintenance and advanced reactors issue, 2008

    SciTech Connect

    Agnihotri, Newal

    2009-09-15

    The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Technologies of national importance, by Tsutomu Ohkubo, Japan Atomic Energy Agency, Japan; Modeling and simulation advances brighten future nuclear power, by Hussein Khalil, Argonne National Laboratory, Energy and desalination projects, by Ratan Kumar Sinha, Bhabha Atomic Research Centre, India; A plant with simplified design, by John Higgins, GE Hitachi Nuclear Energy; A forward thinking design, by Ray Ganthner, AREVA; A passively safe design, by Ed Cummins, Westinghouse Electric Company; A market-ready design, by Ken Petrunik, Atomic Energy of Canada Limited, Canada; Generation IV Advanced Nuclear Energy Systems, by Jacques Bouchard, French Commissariat a l'Energie Atomique, France, and Ralph Bennett, Idaho National Laboratory; Innovative reactor designs, a report by IAEA, Vienna, Austria; Guidance for new vendors, by John Nakoski, U.S. Nuclear Regulatory Commission; Road map for future energy, by John Cleveland, International Atomic Energy Agency, Vienna, Austria; and, Vermont's largest source of electricity, by Tyler Lamberts, Entergy Nuclear Operations, Inc. The Industry Innovation article is titled Intelligent monitoring technology, by Chris Demars, Exelon Nuclear.

  6. NEUTRONIC REACTOR

    DOEpatents

    Fraas, A.P.; Mills, C.B.

    1961-11-21

    A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)

  7. REACTOR COOLING

    DOEpatents

    Quackenbush, C.F.

    1959-09-29

    A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

  8. Surgical research IV.

    PubMed

    Toledo-Pereyra, Luis H

    2010-08-01

    Harvey W. Cushing (1869-1939) is the only surgeon represented in Surgical Research IV and one of the most accomplished American contributors to surgical research in general and to neurological and endocrine surgery research in particular. Other surgical research leaders of the 19th and 20th centuries who preceded Harvey Cushing have been introduced before. First, we highlighted the "importance of medical and surgical research" as the basic elements in the advancement of medicine and surgery could be considered as Surgical Research I. Second, in Surgical Research II, we presented William Beaumont, Samuel Gross, and William Halsted as the most important participants of the first wave of American surgical researchers. Next, in Surgical Research III, we considered surgeon researchers who moved ahead in the field of surgery with their research initiatives at the time, including John B. Murphy, the Mayo Brothers William J. and Charles H. Mayo, and George W. Crile. With Harvey Cushing, we enter an era of surgical research associated with neurosurgery and endocrine surgery as part of Surgical Research IV. PMID:20690841

  9. Division Iv: Stars

    NASA Astrophysics Data System (ADS)

    Corbally, Christopher; D'Antona, Francesca; Spite, Monique; Asplund, Martin; Charbonnel, Corinne; Docobo, Jose Angel; Gray, Richard O.; Piskunov, Nikolai E.

    2012-04-01

    This Division IV was started on a trial basis at the General Assembly in The Hague 1994 and was formally accepted at the Kyoto General Assembly in 1997. Its broad coverage of ``Stars'' is reflected in its relatively large number of Commissions and so of members (1266 in late 2011). Its kindred Division V, ``Variable Stars'', has the same history of its beginning. The thinking at the time was to achieve some kind of balance between the number of members in each of the 12 Divisions. Amid the current discussion of reorganizing the number of Divisions into a more compact form it seems advisable to make this numerical balance less of an issue than the rationalization of the scientific coverage of each Division, so providing more effective interaction within a particular field of astronomy. After all, every star is variable to a certain degree and such variability is becoming an ever more powerful tool to understand the characteristics of every kind of normal and peculiar star. So we may expect, after hearing the reactions of members, that in the restructuring a single Division will result from the current Divisions IV and V.

  10. IVS contribution to ITRF2014

    NASA Astrophysics Data System (ADS)

    Bachmann, Sabine; Thaller, Daniela; Roggenbuck, Ole; Lösler, Michael; Messerschmitt, Linda

    2016-07-01

    Every few years the International Terrestrial Reference System (ITRS) Center of the International Earth Rotation and Reference Systems Service (IERS) decides to generate a new version of the International Terrestrial Reference Frame (ITRF). For the upcoming ITRF2014 the official contribution of the International VLBI Service for Geodesy and Astrometry (IVS) comprises 5796 combined sessions in SINEX file format from 1979.6 to 2015.0 containing 158 stations, overall. Nine AC contributions were included in the combination process, using five different software packages. Station coordinate time series of the combined solution show an overall repeatability of 3.3 mm for the north, 4.3 mm for the east and 7.5 mm for the height component over all stations. The minimum repeatabilities are 1.5 mm for north, 2.1 mm for east and 2.9 mm for height. One of the important differences between the IVS contribution to the ITRF2014 and the routine IVS combination is the omission of the correction for non-tidal atmospheric pressure loading (NTAL). Comparisons between the amplitudes of the annual signals derived by the VLBI observations and the annual signals from an NTAL model show that for some stations, NTAL has a high impact on station height variation. For other stations, the effect of NTAL is low. Occasionally other loading effects have a higher influence (e.g. continental water storage loading). External comparisons of the scale parameter between the VTRF2014 (a TRF based on combined VLBI solutions), DTRF2008 (DGFI-TUM realization of ITRS) and ITRF2008 revealed a significant difference in the scale. A scale difference of 0.11 ppb (i.e. 0.7 mm on the Earth's surface) has been detected between the VTRF2014 and the DTRF2008, and a scale difference of 0.44 ppb (i.e. 2.8 mm on the Earth's surface) between the VTRF2014 and ITRF2008. Internal comparisons between the EOP of the combined solution and the individual solutions from the AC contributions show a WRMS in X- and Y-Pole between

  11. Nuclear Data Needs for the Assessment of Gen IV Systems

    NASA Astrophysics Data System (ADS)

    Rimpault, Gerald

    2006-04-01

    Four of the six nuclear systems identified by the Gen-IV international forum are relying in fast reactors. The high performance required from these future FR's calls for very innovative core characteristics compared with conventional fast reactor designs, which in turns give rise to new challenges for the available neutronics methods and data. The ERANOS “formulaire” was developed for reliable, precise and efficient calculations of sodium-cooled fast neutron reactor cores. Such a “formulaire” enables the prediction of all the neutronic quantities of interest for reactor design, operation and safety studies, along with their corresponding uncertainties. Given its achievement in terms of accuracy for existing sodium cooled reactors, the way the ERANOS “formulaire” has been designed should be looked as an example for developing similar tools for GENIV fast cores. The methodology for defining nuclear data needs is briefly covered. Target accuracies for GEN-IV neutronic characteristics is an important point to start with. The covariance data is of significant importance for quantifying the needs and evaluators should provide them with their nuclear data. Hence, nuclear data requests should be associated to uncertainty values. Integral experiments have a complementary role to differential measurements for meeting some nuclear data needs (for instance Pu239 fission). High Priority Level Requests should include those nuclides accessible with current differential measurement technology and should include less important nuclides A list of potential requests is existing, quantifying their uncertainties remain a significant effort particularly when looking at fuel cycle quantities.

  12. The integral fast reactor and its role in a new generation of nuclear power plants, Tokai, Japan, November 19-21, 1986

    SciTech Connect

    Smith, R.R.

    1986-01-01

    This report presents information on the Integral Fast Reactor and its role in the future. Information is presented in the areas of: inherent safety; other virtues of sodium-cooled breeder; and solving LWR fuel cycle problems with IFR technologies. (JDB)

  13. 78 FR 2390 - CSOLAR IV South, LLC, Wistaria Ranch Solar, LLC, CSOLAR IV West, LLC, CSOLAR IV North, LLC v...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-01-11

    ... Energy Regulatory Commission CSOLAR IV South, LLC, Wistaria Ranch Solar, LLC, CSOLAR IV West, LLC, CSOLAR IV North, LLC v. California Independent System Operator Corporation; Notice of Complaint Take notice... IV South, LLC, Wistaria Ranch Solar, LLC, CSOLAR IV West, LLC and CSOLAR IV North, LLC...

  14. Liquid fuel molten salt reactors for thorium utilization

    DOE PAGES

    Gehin, Jess C.; Powers, Jeffrey J.

    2016-04-08

    Molten salt reactors (MSRs) represent a class of reactors that use liquid salt, usually fluoride- or chloride-based, as either a coolant with a solid fuel (such as fluoride salt-cooled high temperature reactors) or as a combined coolant and fuel with fuel dissolved in a carrier salt. For liquid-fuelled MSRs, the salt can be processed online or in a batch mode to allow for removal of fission products as well as introduction of fissile fuel and fertile materials during reactor operation. The MSR is most commonly associated with the 233U/thorium fuel cycle, as the nuclear properties of 233U combined with themore » online removal of parasitic absorbers allow for the ability to design a thermal-spectrum breeder reactor; however, MSR concepts have been developed using all neutron energy spectra (thermal, intermediate, fast, and mixed-spectrum zoned concepts) and with a variety of fuels including uranium, thorium, plutonium, and minor actinides. Early MSR work was supported by a significant research and development (R&D) program that resulted in two experimental systems operating at ORNL in the 1960s, the Aircraft Reactor Experiment and the Molten Salt Reactor Experiment. Subsequent design studies in the 1970s focusing on thermal-spectrum thorium-fueled systems established reference concepts for two major design variants: (1) a molten salt breeder reactor (MSBR), with multiple configurations that could breed additional fissile material or maintain self-sustaining operation; and (2) a denatured molten salt reactor (DMSR) with enhanced proliferation-resistance. T MSRs has been selected as one of six most promising Generation IV systems and development activities have been seen in fast-spectrum MSRs, waste-burning MSRs, MSRs fueled with low-enriched uranium (LEU), as well as more traditional thorium fuel cycle-based MSRs. This study provides an historical background of MSR R&D efforts, surveys and summarizes many of the recent development, and provides analysis comparing

  15. NEUTRONIC REACTOR CONTROL

    DOEpatents

    Untermyer, S.; Hutter, E.

    1959-08-01

    This patent relates to "shadow" control of a nuclear reactor. The control means comprises a plurality ot elongated rods disposed adjacent and parallel to each other, The morphology and effects of gases generated within sections of neutron absorbing materials and equal length sections of neutron permeable materials together with means for longitudinally pcsitioning the rcds relative to each other.

  16. Reactor Safety Research Programs

    SciTech Connect

    Dotson, CW

    1980-08-01

    This document summarizes the work performed by Pacific Northwest laboratory from October 1 through December 31, 1979, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, lspra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  17. STAR: The Secure Transportable Autonomous Reactor System - Encapsulated Fission Heat Source

    SciTech Connect

    Ehud Greenspan

    2003-10-31

    OAK-B135 The Encapsulated Nuclear Heat Source (ENHS) is a novel 125 MWth fast spectrum reactor concept that was selected by the 1999 DOE NERI program as a candidate ''Generation-IV'' reactor. It uses Pb-Bi or other liquid-metal coolant and is intended to be factory manufactured in large numbers to be economically competitive. It is anticipated to be most useful to developing countries. The US team studying the feasibility of the ENHS reactor concept consisted of the University of California, Berkeley, Argonne National Laboratory (ANL), Lawrence Livermore National Laboratory (LLNL) and Westinghouse. Collaborating with the US team were three Korean organizations: Korean Atomic Energy Research Institute (KAERI), Korean Advanced Institute for Science and Technology (KAIST) and the University of Seoul, as well as the Central Research Institute of the Electrical Power Industry (CRIEPI) of Japan. Unique features of the ENHS include at least 20 years of operation without refueling; no fuel handling in the host country; no pumps and valves; excess reactivity does not exceed 1$; fully passive removal of the decay heat; very small probability of core damaging accidents; autonomous operation and capability of load-following over a wide range; very long plant life. In addition it offers a close match between demand and supply, large tolerance to human errors, is likely to get public acceptance via demonstration of superb safety, lack of need for offsite response, and very good proliferation resistance. The ENHS reactor is designed to meet the requirements of Generation IV reactors including sustainable energy supply, low waste, high level of proliferation resistance, high level of safety and reliability, acceptable risk to capital and, hopefully, also competitive busbar cost of electricity.

  18. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1958-04-22

    A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

  19. NEUTRONIC REACTOR

    DOEpatents

    Metcalf, H.E.; Johnson, H.W.

    1961-04-01

    BS>A nuclear reactor incorporating fuel rods passing through a moderator and including tubes of a material of higher Thermal conductivity than the fuel in contact with the fuel is described. The tubes extend beyond the active portion of the reactor into contant with a fiuld coolant.

  20. Reactor building

    SciTech Connect

    Hista, J. C.

    1984-09-18

    Reactor building comprising a vessel shaft anchored in a slab which is peripherally locked. This reactor building comprises a confinement enclosure within which are positioned internal structures constituted by an internal structure floor, a vessel shaft, a slab being positioned between the general floor and the internal structure floor, the vesse

  1. dBASE IV basics

    SciTech Connect

    O`Connor, P.

    1994-09-01

    This is a user`s manual for dBASE IV. dBASE IV is a popular software application that can be used on your personal computer to help organize and maintain your database files. It is actually a set of tools with which you can create, organize, select and manipulate data in a simple yet effective manner. dBASE IV offers three methods of working with the product: (1) control center: (2) command line; and (3) programming.

  2. On-Line NDE for Advanced Reactor Designs

    NASA Astrophysics Data System (ADS)

    Nakagawa, N.; Inanc, F.; Thompson, R. B.; Junker, W. R.; Ruddy, F. H.; Beatty, J. M.; Arlia, N. G.

    2003-03-01

    This expository paper introduces the concept of on-line sensor methodologies for monitoring the integrity of components in next generation power systems, and explains general benefits of the approach, while describing early conceptual developments of suitable NDE methodologies. The paper first explains the philosophy behind this approach (i.e. the design-for-inspectability concept). Specifically, we describe where and how decades of accumulated knowledge and experience in nuclear power system maintenance are utilized in Generation IV power system designs, as the designs are being actively developed, in order to advance their safety and economy. Second, we explain that Generation IV reactor design features call for the replacement of the current outage-based maintenance by on-line inspection and monitoring. Third, the model-based approach toward design and performance optimization of on-line sensor systems, using electromagnetic, ultrasonic, and radiation detectors, will be explained. Fourth, general types of NDE inspections that are considered amenable to on-line health monitoring will be listed. Fifth, we will describe specific modeling developments to be used for radiography, EMAT UT, and EC detector design studies.

  3. Painlevé IV coherent states

    SciTech Connect

    Bermudez, David; Contreras-Astorga, Alonso; Fernández C, David J.

    2014-11-15

    A simple way to find solutions of the Painlevé IV equation is by identifying Hamiltonian systems with third-order differential ladder operators. Some of these systems can be obtained by applying supersymmetric quantum mechanics (SUSY QM) to the harmonic oscillator. In this work, we will construct families of coherent states for such subset of SUSY partner Hamiltonians which are connected with the Painlevé IV equation. First, these coherent states are built up as eigenstates of the annihilation operator, then as displaced versions of the extremal states, both involving the related third-order ladder operators, and finally as extremal states which are also displaced but now using the so called linearized ladder operators. To each SUSY partner Hamiltonian corresponds two families of coherent states: one inside the infinite subspace associated with the isospectral part of the spectrum and another one in the finite subspace generated by the states created through the SUSY technique. - Highlights: • We use SUSY QM to obtain Hamiltonians with third-order differential ladder operators. • We show that these systems are related with the Painlevé IV equation. • We apply different definitions of coherent states to these Hamiltonians using the third-order ladder operators and some linearized ones. • We construct families of coherent states for such systems, which we called Painlevé IV coherent states.

  4. Industrial Waste Landfill IV upgrade package

    SciTech Connect

    Not Available

    1994-03-29

    The Y-12 Plant, K-25 Site, and ORNL are managed by DOE`s Operating Contractor (OC), Martin Marietta Energy Systems, Inc. (Energy Systems) for DOE. Operation associated with the facilities by the Operating Contractor and subcontractors, DOE contractors and the DOE Federal Building result in the generation of industrial solid wastes as well as construction/demolition wastes. Due to the waste streams mentioned, the Y-12 Industrial Waste Landfill IV (IWLF-IV) was developed for the disposal of solid industrial waste in accordance to Rule 1200-1-7, Regulations Governing Solid Waste Processing and Disposal in Tennessee. This revised operating document is a part of a request for modification to the existing Y-12 IWLF-IV to comply with revised regulation (Rule Chapters 1200-1-7-.01 through 1200-1-7-.08) in order to provide future disposal space for the ORR, Subcontractors, and the DOE Federal Building. This revised operating manual also reflects approved modifications that have been made over the years since the original landfill permit approval. The drawings referred to in this manual are included in Drawings section of the package. IWLF-IV is a Tennessee Department of Environmental and Conservation/Division of Solid Waste Management (TDEC/DSWM) Class 11 disposal unit.

  5. Gas-Cooled Fast Reactor (GFR) FY05 Annual Report

    SciTech Connect

    K. D. Weaver; T. Marshall; T. Totemeier; J. Gan; E.E. Feldman; E.A Hoffman; R.F. Kulak; I.U. Therios; C. P. Tzanos; T.Y.C. Wei; L-Y. Cheng; H. Ludewig; J. Jo; R. Nanstad; W. Corwin; V. G. Krishnardula; W. F. Gale; J. W. Fergus; P. Sabharwall; T. Allen

    2005-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection. Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in

  6. Fusion reactor pumped laser

    DOEpatents

    Jassby, Daniel L.

    1988-01-01

    A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam.

  7. NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Daniels, F.

    1957-10-15

    Gas-cooled solid-moderator type reactors wherein the fissionable fuel and moderator materials are each in the form of solid pebbles, or discrete particles, and are substantially homogeneously mixed in the proper proportion and placed within the core of the reactor are described. The shape of these discrete particles must be such that voids are present between them when mixed together. Helium enters the bottom of the core and passes through the voids between the fuel and moderator particles to absorb the heat generated by the chain reaction. The hot helium gas is drawn off the top of the core and may be passed through a heat exchanger to produce steam.

  8. Confirmatory Factor Analysis of the WAIS-IV/WMS-IV

    ERIC Educational Resources Information Center

    Holdnack, James A.; Zhou, Xiaobin; Larrabee, Glenn J.; Millis, Scott R.; Salthouse, Timothy A.

    2011-01-01

    The Wechsler Adult Intelligence Scale-fourth edition (WAIS-IV) and the Wechsler Memory Scale-fourth edition (WMS-IV) were co-developed to be used individually or as a combined battery of tests. The independent factor structure of each of the tests has been identified; however, the combined factor structure has yet to be determined. Confirmatory…

  9. Improving IV-A/IV-D Interface. Trainer Guide.

    ERIC Educational Resources Information Center

    National Inst. for Child Support Enforcement, Chevy Chase, MD.

    Effective interface between the Aid to Families with Dependent Children (IV-A) and the Child Support Enforcement (IV-D) programs is a key factor in assisting families in becoming self-sufficient, reducing welfare expenditures, and enforcing parental responsibility to support their children. Consequently, overcoming the procedural, technological,…

  10. Improving IV-A/IV-D Interface. Handbook.

    ERIC Educational Resources Information Center

    National Inst. for Child Support Enforcement, Chevy Chase, MD.

    Effective interface between the Aid to Families with Dependent Children (IV-A) and the Child Support Enforcement (IV-D) programs is a key factor in assisting families in becoming self-sufficient, reducing welfare expenditures, and enforcing parental responsibility to support their children. Consequently, overcoming the procedural, technological,…

  11. Compact Reactor

    SciTech Connect

    Williams, Pharis E.

    2007-01-30

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date.

  12. NUCLEAR REACTOR

    DOEpatents

    Moore, R.V.; Bowen, J.H.; Dent, K.H.

    1958-12-01

    A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

  13. DOE/NNSA perspective safeguard by design: GEN III/III+ light water reactors and beyond

    SciTech Connect

    Pan, Paul Y

    2010-12-10

    An overview of key issues relevant to safeguards by design (SBD) for GEN III/IV nuclear reactors is provided. Lessons learned from construction of typical GEN III+ water reactors with respect to SBD are highlighted. Details of SBD for safeguards guidance development for GEN III/III+ light water reactors are developed and reported. This paper also identifies technical challenges to extend SBD including proliferation resistance methodologies to other GEN III/III+ reactors (except HWRs) and GEN IV reactors because of their immaturity in designs.

  14. A NEUTRONIC REACTOR

    DOEpatents

    Luebke, E.A.; Vandenberg, L.B.

    1959-09-01

    A nuclear reactor for producing thermoelectric power is described. The reactor core comprises a series of thermoelectric assemblies, each assembly including fissionable fuel as an active element to form a hot junction and a thermocouple. The assemblies are disposed parallel to each other to form spaces and means are included for Introducing an electrically conductive coolant between the assemblies to form cold junctions of the thermocouples. An electromotive force is developed across the entire series of the thermoelectric assemblies due to fission heat generated in the fuel causing a current to flow perpendicular to the flow of coolant and is distributed to a load outside of the reactor by means of bus bars electrically connected to the outermost thermoelectric assembly.

  15. Merchant Marine Ship Reactor

    DOEpatents

    Sankovich, M. F.; Mumm, J. F.; North, Jr, D. C.; Rock, H. R.; Gestson, D. K.

    1961-05-01

    A nuclear reactor for use in a merchant marine ship is described. The reactor is of pressurized, light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements that are confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass. (AEC)

  16. MERCHANT MARINE SHIP REACTOR

    DOEpatents

    Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.

    1961-05-01

    A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.

  17. The soft, fluctuating UVB at z ˜ 6 as traced by C IV, Si IV, and C II

    NASA Astrophysics Data System (ADS)

    Finlator, Kristian; Oppenheimer, B. D.; Davé, Romeel; Zackrisson, E.; Thompson, Robert; Huang, Shuiyao

    2016-07-01

    The sources that drove cosmological reionization left clues regarding their identity in the slope and inhomogeneity of the ultraviolet ionizing background (UVB): bright quasars (QSOs) generate a hard UVB with predominantly large-scale fluctuations while Population II stars generate a softer one with smaller scale fluctuations. Metal absorbers probe the UVB's slope because different ions are sensitive to different energies. Likewise, they probe spatial fluctuations because they originate in regions where a galaxy-driven UVB is harder and more intense. We take a first step towards studying the reionization-epoch UVB's slope and inhomogeneity by comparing observations of 12 metal absorbers at z ˜ 6 versus predictions from a cosmological hydrodynamic simulation using three different UVBs: a soft, spatially inhomogeneous `galaxies+QSOs' UVB; a homogeneous `galaxies+QSOs' UVB, and a `QSOs-only' model. All UVBs reproduce the observed column density distributions of C II, Si IV, and C IV reasonably well although high-column, high-ionization absorbers are underproduced, reflecting numerical limitations. With upper limits treated as detections, only a soft, fluctuating UVB reproduces both the observed Si IV/C IV and C II/C IV distributions. The QSOs-only UVB overpredicts both C IV/C II and C IV/Si IV, indicating that it is too hard. The Haardt & Madau (2012) UVB underpredicts C IV/Si IV, suggesting that it lacks amplifications near galaxies. Hence current observations prefer a soft, fluctuating UVB as expected from a predominantly Population II background although they cannot rule out a harder one. Future observations probing a factor of 2 deeper in metal column density will distinguish between the soft, fluctuating and QSOs-only UVBs.

  18. NATIONAL COASTAL CONDITION REPORT IV

    EPA Science Inventory

    The National Coastal Condition Report IV (NCCR IV) is the fourth in a series of environmental assessments of U.S. coastal waters and the Great Lakes. The report includes assessments of all the nation’s estuaries in the contiguous 48 states and Puerto Rico, south-eastern Alaska, ...

  19. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.

    1960-04-01

    A nuclear reactor is described consisting of blocks of graphite arranged in layers, natural uranium bodies disposed in holes in alternate layers of graphite blocks, and coolant tubes disposed in the layers of graphite blocks which do not contain uranium.

  20. NEUTRONIC REACTOR

    DOEpatents

    Anderson, H.L.

    1960-09-20

    A nuclear reactor is described comprising fissionable material dispersed in graphite blocks, helium filling the voids of the blocks and the spaces therebetween, and means other than the helium in thermal conductive contact with the graphite for removing heat.

  1. Chemical Reactors.

    ERIC Educational Resources Information Center

    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)

  2. NEUTRONIC REACTOR

    DOEpatents

    Hurwitz, H. Jr.; Brooks, H.; Mannal, C.; Payne, J.H.; Luebke, E.A.

    1959-03-24

    A reactor of the heterogeneous, liquid cooled type is described. This reactor is comprised of a central region of a plurality of vertically disposed elongated tubes surrounded by a region of moderator material. The central region is comprised of a central core surrounded by a reflector region which is surrounded by a fast neutron absorber region, which in turn is surrounded by a slow neutron absorber region. Liquid sodium is used as the primary coolant and circulates through the core which contains the fuel elements. Control of the reactor is accomplished by varying the ability of the reflector region to reflect neutrons back into the core of the reactor. For this purpose the reflector is comprised of moderator and control elements having varying effects on reactivity, the control elements being arranged and actuated by groups to give regulation, shim, and safety control.

  3. NUCLEAR REACTOR

    DOEpatents

    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.

  4. REACTOR SHIELD

    DOEpatents

    Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

    1959-02-17

    Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

  5. NUCLEAR REACTOR

    DOEpatents

    Sherman, J.; Sharbaugh, J.E.; Fauth, W.L. Jr.; Palladino, N.J.; DeHuff, P.G.

    1962-10-23

    A nuclear reactor incorporating seed and blanket assemblies is designed. Means are provided for obtaining samples of the coolant from the blanket assemblies and for varying the flow of coolant through the blanket assemblies. (AEC)

  6. NEUTRONIC REACTORS

    DOEpatents

    Vernon, H.C.

    1959-01-13

    A neutronic reactor of the heterogeneous, fluid cooled tvpe is described. The reactor is comprised of a pressure vessel containing the moderator and a plurality of vertically disposed channels extending in spaced relationship through the moderator. Fissionable fuel material is placed within the channels in spaced relationship thereto to permit circulation of the coolant fluid. Separate means are provided for cooling the moderator and for circulating a fluid coolant thru the channel elements to cool the fuel material.

  7. NUCLEAR REACTOR

    DOEpatents

    Anderson, C.R.

    1962-07-24

    A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)

  8. NUCLEAR REACTOR

    DOEpatents

    Breden, C.R.; Dietrich, J.R.

    1961-06-20

    A water-soluble non-volatile poison may be introduced into a reactor to nullify excess reactivity. The poison is removed by passing a side stream of the water containing the soluble poison to an evaporation chamber. The vapor phase is returned to the reactor to decrease the concentration of soluble poison and the liquid phase is returned to increase the concentration of soluble poison.

  9. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1959-07-14

    High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

  10. SCC and corrosion evaluations of the F/M steels for a supercritical water reactor

    NASA Astrophysics Data System (ADS)

    Hwang, Seong Sik; Lee, Byung Hak; Kim, Jung Gu; Jang, Jinsung

    2008-01-01

    As one of the Generation IV nuclear reactors, a supercritical water cooled reactor (SCWR) is being considered as a candidate reactor due to its high thermal efficiency and simple reactor design without steam generators and steam separators. For the application of a structural material to a core's internals and a fuel cladding, the material should be evaluated in terms of its corrosion and stress corrosion cracking susceptibility. Stress corrosion cracking and general corrosion tests of ferritic-martensitic (F/M) steels, high Ni alloys and an oxide dispersion strengthened (ODS) alloy were performed. Stress corrosion cracking (SCC) was not observed on the fractured surface of the T 91 steel in the supercritical water at 500, 550 and 600 °C. As the test temperature increased, the ultimate tensile strength (UTS) and yield strength (YS) of T 91 decreased, and a high dissolved oxygen level induced corrosion and low ductility. The F/M steels showed a high corrosion rate whereas the Ni base alloys showed a little corrosion at 500 and 550 °C. Corrosion rate of the F/M steels at 600 °C test was up to three times larger than that at 500 °C. A thin layer composed of Mo and Ni seems to retard the Cr diffusion into the out layer of the corrosion product of T 92 and T 122.

  11. Ferritic steels for sodium-cooled fast reactors: Design principles and challenges

    NASA Astrophysics Data System (ADS)

    Raj, Baldev; Vijayalakshmi, M.

    2010-09-01

    An overview of the current status of development of ferritic steels for emerging fast reactor technologies is presented in this paper. The creep-resistant 9-12Cr ferritic/martensitic steels are classically known for steam generator applications. The excellent void swelling resistance of ferritic steels enabled the identification of their potential for core component applications of fast reactors. Since then, an extensive knowledge base has been generated by identifying the empirical correlations between chemistry of the steels, heat treatment, structure, and properties, in addition to their in-reactor behavior. A few concerns have also been identified which pertain to high-temperature irradiation creep, embrittlement, Type IV cracking in creep-loaded weldments, and hard zone formation in dissimilar joints. The origin of these problems and the methodologies to overcome the limitations are highlighted. Finally, the suitability of the ferritic steels is re-evaluated in the emerging scenario of the fast reactor technology, with a target of achieving better breeding ratio and improved thermal efficiency.

  12. In situ generation of Co(II) by use of a solid-phase reactor in an FIA assembly for the spectrophotometric determination of penicillamine.

    PubMed

    Corominas, B Gómez-Taylor; Pferzschner, Julia; Icardo, M Catalá; Zamora, L Lahuerta; Martínez Calatayud, J

    2005-09-01

    A flow injection analysis (FIA) manifold for the determination of penicillamine in pharmaceutical preparations is proposed. The manifold includes a solid-phase reactor for the in situ production of the derivatizing reagent, Co(II) ion, which forms a coloured complex with penicillamine in an alkaline medium. The reactor is prepared by natural immobilization of cobalt carbonate on a polymer matrix, which endows it with a high mechanical and microbiological stability. The cobalt released by passage of a 5 x 10(-4) mol l(-1) sulphuric acid stream at a flow-rate of 2.3 ml min(-1) is merged with a volume of 314 microl of sample containing penicillamine in ammonium-ammonia buffer at pH 9.5 to measure the absorbance at 360 nm. Beer's law is obeyed over the penicillamine concentration range 5-60 mg l(-1). The limit of detection (LOD) of the method is 1 mg l(-1) and its throughput 70 samples h(-1).

  13. Catalytic hydrolysis of urea with fly ash for generation of ammonia in a batch reactor for flue gas conditioning and NOx reduction

    SciTech Connect

    Sahu, J.N.; Gangadharan, P.; Patwardhan, A.V.; Meikap, B.C.

    2009-01-15

    Ammonia is a highly volatile noxious material with adverse physiological effects, which become intolerable even at very low concentrations and present substantial environmental and operating hazards and risk. Yet ammonia has long been known to be used for feedstock of flue gas conditioning and NOx reduction. Urea as the source of ammonia for the production of ammonia has the obvious advantages that no ammonia shipping, handling, and storage is required. The process of this invention minimizes the risks and hazards associated with the transport, storage, and use of anhydrous and aqueous ammonia. Yet no such rapid urea conversion process is available as per requirement of high conversion in shorter time, so here we study the catalytic hydrolysis of urea for fast conversion in a batch reactor. The catalyst used in this study is fly ash, a waste material originating in great amounts in combustion processes. A number of experiments were carried out in a batch reactor at different catalytic doses, temperatures, times, and at a constant concentration of urea solution 10% by weight, and equilibrium and kinetic studies have been made.

  14. Aqueous complexation of thorium(IV), uranium(IV), neptunium(IV), plutonium(III/IV), and cerium(III/IV) with DTPA.

    PubMed

    Brown, M Alex; Paulenova, Alena; Gelis, Artem V

    2012-07-16

    Aqueous complexation of Th(IV), U(IV), Np(IV), Pu(III/IV), and Ce(III/IV) with DTPA was studied by potentiometry, absorption spectrophotometry, and cyclic voltammetry at 1 M ionic strength and 25 °C. The stability constants for the 1:1 complex of each trivalent and tetravalent metal were calculated. From the potentiometric data, we report stability constant values for Ce(III)DTPA, Ce(III)HDTPA, and Th(IV)DTPA of log β(101) = 20.01 ± 0.02, log β(111) = 22.0 ± 0.2, and log β(101) = 29.6 ± 1, respectively. From the absorption spectrophotometry data, we report stability constant values for U(IV)DTPA, Np(IV)DTPA, and Pu(IV)DTPA of log β(101) = 31.8 ± 0.1, 32.3 ± 0.1, and 33.67 ± 0.02, respectively. From the cyclic voltammetry data, we report stability constant values for Ce(IV) and Pu(III) of log β(101) = 34.04 ± 0.04 and 20.58 ± 0.04, respectively. The values obtained in this work are compared and discussed with respect to the ionic radius of each cationic metal.

  15. Research reactors - an overview

    SciTech Connect

    West, C.D.

    1997-03-01

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs.

  16. Corrosion Behavior of Candidate Alloys for Supercritical Water Reactors

    SciTech Connect

    Sridharan, K.; Zillmer, A.; Licht, J.R.; Allen, T.R.; Anderson, M.H.; Tan, L.

    2004-07-01

    The corrosion and stress corrosion cracking behavior of metallic cladding and other core internal structures is critical to the success of the Generation IV Supercritical Water-cooled Reactors (SCWR). The eventual materials selected will be chosen based on the combined corrosion, stress-corrosion, mechanical performance, and radiation stability properties. Among the materials being considered are austenitic stainless steels, ferritic/martensitic steels, and nickel-base alloys. This paper reports initial studies on the corrosion performance of the candidate alloys 316 austenitic stainless steel, Inconel 718, and Zircaloy-2, all exposed to supercritical water at 300-500 deg. C in a corrosion loop at the University of Wisconsin. Long-term corrosion performance of AISI 347, also a candidate austenitic steel, has also been examined by sectioning samples from a component that was exposed for a period of about 30 years in supercritical water at the Genoa 3 Supercritical Water fossil power plant located in Genoa, Wisconsin. (authors)

  17. NEUTRONIC REACTOR

    DOEpatents

    Ohlinger, L.A.; Wigner, E.P.; Weinberg, A.M.; Young, G.J.

    1958-09-01

    This patent relates to neutronic reactors of the heterogeneous water cooled type, and in particular to a fuel element charging and discharging means therefor. In the embodiment illustrated the reactor contains horizontal, parallel coolant tubes in which the fuel elements are disposed. A loading cart containing a magnzine for holding a plurality of fuel elements operates along the face of the reactor at the inlet ends of the coolant tubes. The loading cart is equipped with a ram device for feeding fuel elements from the magazine through the inlot ends of the coolant tubes. Operating along the face adjacent the discharge ends of the tubes there is provided another cart means adapted to receive irradiated fuel elements as they are forced out of the discharge ends of the coolant tubes by the incoming new fuel elements. This cart is equipped with a tank coataining a coolant, such as water, into which the fuel elements fall, and a hydraulically operated plunger to hold the end of the fuel element being discharged. This inveation provides an apparatus whereby the fuel elements may be loaded into the reactor, irradiated therein, and unloaded from the reactor without stopping the fiow of the coolant and without danger to the operating personnel.

  18. The web-enabled database of JRC-EC, a useful tool for managing European Gen IV materials data

    NASA Astrophysics Data System (ADS)

    Over, H. H.; Dietz, W.

    2008-06-01

    Materials and document databases are important tools to conserve knowledge and experimental materials data of European R&D projects. A web-enabled application guarantees a fast access to these data. In combination with analysis tools the experimental data are used for e.g. mechanical design, construction and lifetime predictions of complex components. The effective and efficient handling of large amounts of generic and detailed materials data with regard to properties related to e.g. fabrication processes, joining techniques, irradiation or aging is one of the basic elements of data management within ongoing nuclear safety and design related European research projects and networks. The paper describes the structure and functionality of Mat-DB and gives examples how these tools can be used for the management and evaluation of materials data of European (national or multi-national) R&D activities or future reactor types such as the EURATOM FP7 Generation IV reactor types or the heavy liquid metals cooled reactor.

  19. Control of reactor coolant flow path during reactor decay heat removal

    DOEpatents

    Hunsbedt, Anstein N.

    1988-01-01

    An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

  20. ApoA-IV promotes the biogenesis of apoA-IV-containing HDL particles with the participation of ABCA1 and LCAT.

    PubMed

    Duka, Adelina; Fotakis, Panagiotis; Georgiadou, Dimitra; Kateifides, Andreas; Tzavlaki, Kalliopi; von Eckardstein, Leonard; Stratikos, Efstratios; Kardassis, Dimitris; Zannis, Vassilis I

    2013-01-01

    The objective of this study was to establish the role of apoA-IV, ABCA1, and LCAT in the biogenesis of apoA-IV-containing HDL (HDL-A-IV) using different mouse models. Adenovirus-mediated gene transfer of apoA-IV in apoA-I(-/-) mice did not change plasma lipid levels. ApoA-IV floated in the HDL2/HDL3 region, promoted the formation of spherical HDL particles as determined by electron microscopy, and generated mostly α- and a few pre-β-like HDL subpopulations. Gene transfer of apoA-IV in apoA-I(-/-) × apoE(-/-) mice increased plasma cholesterol and triglyceride levels, and 80% of the protein was distributed in the VLDL/IDL/LDL region. This treatment likewise generated α- and pre-β-like HDL subpopulations. Spherical and α-migrating HDL particles were not detectable following gene transfer of apoA-IV in ABCA1(-/-) or LCAT(-/-) mice. Coexpression of apoA-IV and LCAT in apoA-I(-/-) mice restored the formation of HDL-A-IV. Lipid-free apoA-IV and reconstituted HDL-A-IV promoted ABCA1 and scavenger receptor BI (SR-BI)-mediated cholesterol efflux, respectively, as efficiently as apoA-I and apoE. Our findings are consistent with a novel function of apoA-IV in the biogenesis of discrete HDL-A-IV particles with the participation of ABCA1 and LCAT, and may explain previously reported anti-inflammatory and atheroprotective properties of apoA-IV. PMID:23132909

  1. Self isolating high frequency saturable reactor

    SciTech Connect

    Moore, James A.

    1998-06-23

    The present invention discloses a saturable reactor and a method for decoupling the interwinding capacitance from the frequency limitations of the reactor so that the equivalent electrical circuit of the saturable reactor comprises a variable inductor. The saturable reactor comprises a plurality of physically symmetrical magnetic cores with closed loop magnetic paths and a novel method of wiring a control winding and a RF winding. The present invention additionally discloses a matching network and method for matching the impedances of a RF generator to a load. The matching network comprises a matching transformer and a saturable reactor.

  2. POWER REACTOR

    DOEpatents

    Zinn, W.H.

    1958-07-01

    A fast nuclear reactor system ls described for producing power and radioactive isotopes. The reactor core is of the heterogeneous, fluid sealed type comprised of vertically arranged elongated tubular fuel elements having vertical coolant passages. The active portion is surrounded by a neutron reflector and a shield. The system includes pumps and heat exchangers for the primary and secondary coolant circuits. The core, primary coolant pump and primary heat exchanger are disposed within an irapenforate tank which is filled with the primary coolant, in this case a liquid metal such as Na or NaK, to completely submerge these elements. The tank is completely surrounded by a thick walled concrete shield. This reactor system utilizes enriched uranium or plutonium as the fissionable material, uranium or thorium as a diluent and thorium or uranium containing less than 0 7% of the U/sup 235/ isotope as a fertile material.

  3. NEUTRONIC REACTOR

    DOEpatents

    Metcalf, H.E.

    1957-10-01

    A reactor of the type which preferably uses plutonium as the fuel and a liquid moderator, preferably ordinary water, and which produces steam within the reactor core due to the heat of the chain reaction is described. In the reactor shown the fuel elements are essentially in the form of trays and are ventically stacked in spaced relationship. The water moderator is continuously supplied to the trays to maintain a constant level on the upper surfaces of the fuel element as it is continually evaporated by the heat. The steam passes out through the spaces between the fuel elements and is drawn off at the top of the core. The fuel elements are clad in aluminum to prevent deterioration thereof with consequent contamimation of the water.

  4. NUCLEAR REACTOR

    DOEpatents

    Christy, R.F.

    1958-07-15

    A nuclear reactor of the homogeneous liquid fuel type is described wherein the fissionable isotope is suspended or dissolved in a liquid moderator such as water. The reactor core is comprised essentially of a spherical vessel for containing the reactive composition surrounded by a reflector, preferably of beryllium oxide. The reactive composition may be an ordinary water solution of a soluble salt of uranium, the quantity of fissionable isotope in solution being sufficient to provide a critical mass in the vessel. The liquid fuel is stored in a tank of non-crtttcal geometry below the reactor vessel and outside of the reflector and is passed from the tank to the vessel through a pipe connecting the two by air pressure means. Neutron absorbing control and safety rods are operated within slots in the reflector adjacent to the vessel.

  5. Bioconversion reactor

    DOEpatents

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  6. REACTOR CONTROL

    DOEpatents

    Fortescue, P.; Nicoll, D.

    1962-04-24

    A control system employed with a high pressure gas cooled reactor in which a control rod is positioned for upward and downward movement into the neutron field from a position beneath the reactor is described. The control rod is positioned by a coupled piston cylinder releasably coupled to a power drive means and the pressurized coolant is directed against the lower side of the piston. The coolant pressure is offset by a higher fiuid pressure applied to the upper surface of the piston and means are provided for releasing the higher pressure on the upper side of the piston so that the pressure of the coolant drives the piston upwardly, forcing the coupled control rod into the ncutron field of the reactor. (AEC)

  7. NEUTRONIC REACTORS

    DOEpatents

    Wigner, E.P.; Young, G.J.

    1958-10-14

    A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

  8. Impact of thorium based molten salt reactor on the closure of the nuclear fuel cycle

    NASA Astrophysics Data System (ADS)

    Jaradat, Safwan Qasim Mohammad

    Molten salt reactor (MSR) is one of six reactors selected by the Generation IV International Forum (GIF). The liquid fluoride thorium reactor (LFTR) is a MSR concept based on thorium fuel cycle. LFTR uses liquid fluoride salts as a nuclear fuel. It uses 232Th and 233U as the fertile and fissile materials, respectively. Fluoride salt of these nuclides is dissolved in a mixed carrier salt of lithium and beryllium (FLiBe). The objective of this research was to complete feasibility studies of a small commercial thermal LFTR. The focus was on neutronic calculations in order to prescribe core design parameter such as core size, fuel block pitch (p), fuel channel radius, fuel path, reflector thickness, fuel salt composition, and power. In order to achieve this objective, the applicability of Monte Carlo N-Particle Transport Code (MCNP) to MSR modeling was verified. Then, a prescription for conceptual small thermal reactor LFTR and relevant calculations were performed using MCNP to determine the main neutronic parameters of the core reactor. The MCNP code was used to study the reactor physics characteristics for the FUJI-U3 reactor. The results were then compared with the results obtained from the original FUJI-U3 using the reactor physics code SRAC95 and the burnup analysis code ORIPHY2. The results were comparable with each other. Based on the results, MCNP was found to be a reliable code to model a small thermal LFTR and study all the related reactor physics characteristics. The results of this study were promising and successful in demonstrating a prefatory small commercial LFTR design. The outcome of using a small core reactor with a diameter/height of 280/260 cm that would operate for more than five years at a power level of 150 MWth was studied. The fuel system 7LiF - BeF2 - ThF4 - UF4 with a (233U/ 232Th) = 2.01 % was the candidate fuel for this reactor core.

  9. Fundamentals, current state of the development of, and prospects for further improvement of the new-generation thermal-hydraulic computational HYDRA-IBRAE/LM code for simulation of fast reactor systems

    NASA Astrophysics Data System (ADS)

    Alipchenkov, V. M.; Anfimov, A. M.; Afremov, D. A.; Gorbunov, V. S.; Zeigarnik, Yu. A.; Kudryavtsev, A. V.; Osipov, S. L.; Mosunova, N. A.; Strizhov, V. F.; Usov, E. V.

    2016-02-01

    The conceptual fundamentals of the development of the new-generation system thermal-hydraulic computational HYDRA-IBRAE/LM code are presented. The code is intended to simulate the thermalhydraulic processes that take place in the loops and the heat-exchange equipment of liquid-metal cooled fast reactor systems under normal operation and anticipated operational occurrences and during accidents. The paper provides a brief overview of Russian and foreign system thermal-hydraulic codes for modeling liquid-metal coolants and gives grounds for the necessity of development of a new-generation HYDRA-IBRAE/LM code. Considering the specific engineering features of the nuclear power plants (NPPs) equipped with the BN-1200 and the BREST-OD-300 reactors, the processes and the phenomena are singled out that require a detailed analysis and development of the models to be correctly described by the system thermal-hydraulic code in question. Information on the functionality of the computational code is provided, viz., the thermalhydraulic two-phase model, the properties of the sodium and the lead coolants, the closing equations for simulation of the heat-mass exchange processes, the models to describe the processes that take place during the steam-generator tube rupture, etc. The article gives a brief overview of the usability of the computational code, including a description of the support documentation and the supply package, as well as possibilities of taking advantages of the modern computer technologies, such as parallel computations. The paper shows the current state of verification and validation of the computational code; it also presents information on the principles of constructing of and populating the verification matrices for the BREST-OD-300 and the BN-1200 reactor systems. The prospects are outlined for further development of the HYDRA-IBRAE/LM code, introduction of new models into it, and enhancement of its usability. It is shown that the program of development and

  10. NUCLEAR REACTOR

    DOEpatents

    Young, G.

    1963-01-01

    This patent covers a power-producing nuclear reactor in which fuel rods of slightly enriched U are moderated by heavy water and cooled by liquid metal. The fuel rods arranged parallel to one another in a circle are contained in a large outer closed-end conduit that extends into a tank containing the heavy water. Liquid metal is introduced into the large conduit by a small inner conduit that extends within the circle of fuel rods to a point near the lower closed end of the outer conduit. (AEC) Production Reactors

  11. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.; Szilard, L.

    1957-09-24

    Reactors of the type employing plates of natural uranium in a moderator are discussed wherein the plates are um-formly disposed in parallel relationship to each other thereby separating the moderator material into distinct and individual layers. Each plate has an uninterrupted sunface area substantially equal to the cross-sectional area of the active portion of the reactor, the particular size of the plates and the volume ratio of moderator to uranium required to sustain a chain reaction being determinable from the known purity of these materials and other characteristics such as the predictable neutron losses due to the formation of radioactive elements of extremely high neutron capture cross section.

  12. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.; Weinberg, A.W.; Young, G.J.

    1958-04-15

    A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

  13. Thermal-hydraulic interfacing code modules for CANDU reactors

    SciTech Connect

    Liu, W.S.; Gold, M.; Sills, H.

    1997-07-01

    The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.

  14. REACTOR-FLASH BOILER-FLYWHEEL POWER PLANT

    DOEpatents

    Loeb, E.

    1961-01-17

    A power generator in the form of a flywheel with four reactors positioned about its rim is described. The reactors are so positioned that steam, produced in the reactor, exists tangentially to the flywheel, giving it a rotation. The reactors are incompletely moderated without water. The water enters the flywheel at its axis, under sufficient pressure to force it through the reactors, where it is converted to steam. The fuel consists of parallel twisted ribbons assembled to approximate a cylinder.

  15. Simplified modeling of liquid sodium medium with temperature and velocity gradient using real thermal-hydraulic data. Application to ultrasonic thermometry in sodium fast reactor

    NASA Astrophysics Data System (ADS)

    Massacret, N.; Moysan, J.; Ploix, M. A.; Jeannot, J. P.; Corneloup, G.

    2013-01-01

    In the framework of the French R&D program for the Generation IV reactors and specifically for the sodium cooled fast reactors (SFR), studies are carried out on innovative instrumentation methods in order to improve safety and to simplify the monitoring of fundamental physical parameters during reactor operation. The aim of the present work is to develop an acoustic thermometry method to follow up the sodium temperature at the outlet of subassemblies. The medium is a turbulent flow of liquid sodium at 550 °C with temperature inhomogeneities. To understand the effect of disturbance created by this medium, numerical simulations are proposed. A ray tracing code has been developed with Matlabin order to predict acoustic paths in this medium. This complex medium is accurately described by thermal-hydraulic data which are issued from a simulation of a real experiment in Japan. The analysis of these results allows understanding the effects of medium inhomogeneities on the further thermometric acoustic measurement.

  16. Driver for solar cell I-V characteristic plots

    NASA Technical Reports Server (NTRS)

    Turner, G. B. (Inventor)

    1980-01-01

    A bipolar voltage ramp generator which applies a linear voltage through a resistor to a solar cell for plotting its current versus voltage (I-V) characteristic between short circuit and open circuit conditions is disclosed. The generator has automatic stops at the end points. The resistor serves the multiple purpose of providing a current sensing resistor, setting the full-scale current value, and providing a load line with a slope approximately equal to one, such that it will pass through the origin and the approximate center of the I-V curve with about equal distance from that center to each of the end points.

  17. Geologic Investigation of a Potential Site for a Next-Generation Reactor Neutrino Oscillation Experiment -- Diablo Canyon, San Luis Obispo County, CA

    SciTech Connect

    Onishi, Celia Tiemi; Dobson, Patrick; Nakagawa, Seiji; Glaser, Steven; Galic, Dom

    2004-08-01

    This report provides information on the geology and selected physical and mechanical properties of surface rocks collected at Diablo Canyon, San Luis Obispo County, California as part of the design and engineering studies towards a future reactor neutrino oscillation experiment. The main objective of this neutrino project is to study the process of neutrino flavor transformation--or neutrino oscillation--by measuring neutrinos produced in the fission reactions of a nuclear power plant. Diablo Canyon was selected as a candidate site because it allows the detectors to be situated underground in a tunnel close to the source of neutrinos (i.e., at a distance of several hundred meters from the nuclear power plant) while having suitable topography for shielding against cosmic rays. The detectors have to be located underground to minimize the cosmic ray-related background noise that can mimic the signal of reactor neutrino interactions in the detector. Three Pliocene-Miocene marine sedimentary units dominate the geology of Diablo Canyon: the Pismo Formation, the Monterey Formation, and the Obispo Formation. The area is tectonically active, located east of the active Hosgri Fault and in the southern limb of the northwest trending Pismo Syncline. Most of the potential tunnel for the neutrino detector lies within the Obispo Formation. Review of previous geologic studies, observations from a field visit, and selected physical and mechanical properties of rock samples collected from the site provided baseline geological information used in developing a preliminary estimate for tunneling construction cost. Gamma-ray spectrometric results indicate low levels of radioactivity for uranium, thorium, and potassium. Grain density, bulk density, and porosity values for these rock samples range from 2.37 to 2.86 g/cc, 1.41 to 2.57 g/cc, and 1.94 to 68.5% respectively. Point load, unconfined compressive strength, and ultrasonic velocity tests were conducted to determine rock mechanical

  18. Geologic Investigation of a Potential Site for a Next-Generation Reactor Neutrino Oscillation Experiment -- Diablo Canyon, San Luis Obispo County, CA

    SciTech Connect

    Onishi, Celia Tiemi; Dobson, Patrick; Nakagawa, Seiji; Glaser, Steven; Galic, Dom

    2004-06-11

    This report provides information on the geology and selected physical and mechanical properties of surface rocks collected at Diablo Canyon, San Luis Obispo County, California as part of the design and engineering studies towards a future reactor neutrino oscillation experiment. The main objective of this neutrino project is to study the process of neutrino flavor transformation or neutrino oscillation by measuring neutrinos produced in the fission reactions of a nuclear power plant. Diablo Canyon was selected as a candidate site because it allows the detectors to be situated underground in a tunnel close to the source of neutrinos (i.e., at a distance of several hundred meters from the nuclear power plant) while having suitable topography for shielding against cosmic rays. The detectors have to be located underground to minimize the cosmic ray-related background noise that can mimic the signal of reactor neutrino interactions in the detector. Three Pliocene-Miocene marine sedimentary units dominate the geology of Diablo Canyon: the Pismo Formation, the Monterey Formation, and the Obispo Formation. The area is tectonically active, located east of the active Hosgri Fault and in the southern limb of the northwest trending Pismo Syncline. Most of the potential tunnel for the neutrino detector lies within the Obispo Formation. Review of previous geologic studies, observations from a field visit, and selected physical and mechanical properties of rock samples collected from the site provided baseline geological information used in developing a preliminary estimate for tunneling construction cost. Gamma-ray spectrometric results indicate low levels of radioactivity for uranium, thorium, and potassium. Grain density, bulk density, and porosity values for these rock samples range from 2.37 to 2.86 g/cc, 1.41 to 2.57 g/cc, and 1.94 to 68.5 percent respectively. Point load, unconfined compressive strength, and ultrasonic velocity tests were conducted to determine rock

  19. Neutronic reactor

    DOEpatents

    Wende, Charles W. J.

    1976-08-17

    A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield.

  20. Neutronic reactor

    DOEpatents

    Carleton, John T.

    1977-01-25

    A graphite-moderated nuclear reactor includes channels between blocks of graphite and also includes spacer blocks between adjacent channeled blocks with an axis of extension normal to that of the axis of elongation of the channeled blocks to minimize changes in the physical properties of the graphite as a result of prolonged neutron bombardment.

  1. NEUTRONIC REACTOR

    DOEpatents

    Creutz, E.C.; Ohlinger, L.A.; Weinberg, A.M.; Wigner, E.P.; Young, G.J.

    1959-10-27

    BS>A reactor cooled by water, biphenyl, helium, or other fluid with provision made for replacing the fuel rods with the highest plutonium and fission product content without disassembling the entire core and for promptly cooling the rods after their replacement in order to prevent build-up of heat from fission product activity is described.

  2. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1962-12-25

    A reactor is described comprising a plurality of horizontal trays containing a solution of a fissionable material, the trays being sleeved on a vertical tube which contains a vertically-reciprocable control rod, a gas-tight chamber enclosing the trays, and means for conducting vaporized moderator from the chamber and for replacing vaporized moderator in the trays. (AEC)

  3. NEUTRONIC REACTORS

    DOEpatents

    Anderson, H.L.

    1958-10-01

    The design of control rods for nuclear reactors are described. In this design the control rod consists essentially of an elongated member constructed in part of a neutron absorbing material and having tube means extending therethrough for conducting a liquid to cool the rod when in use.

  4. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1960-09-27

    A unit assembly is described for a neutronic reactor comprising a tube and plurality of spaced parallel sandwiches in the tube extending lengthwise thereof, each sandwich including a middle plate having a central opening for plutonium and other openings for fertile material at opposite ends of the plate.

  5. Steam-Reheat Option for Supercritical-Water-Cooled Reactors

    NASA Astrophysics Data System (ADS)

    Saltanov, Eugene

    SuperCritical-Water-cooled Reactors (SCWRs) are being developed as one of the Generation-IV nuclear-reactor concepts. Main objectives of the development are to increase thermal efficiency of a Nuclear Power Plant (NPP) and to decrease capital and operational costs. The first objective can be achieved by introducing nuclear steam reheat inside a reactor and utilizing regenerative feedwater heaters. The second objective can be achieved by designing a steam cycle that closely matches that of the mature supercritical fossil-fuelled power plants. The feasibility of these objectives is discussed. As a part of this discussion, heat-transfer calculations have been performed and analyzed for SuperCritical-Water (SCW) and SuperHeated-Steam (SHS) channels of the proposed reactor concept. In the calculations a uniform and three non-uniform Axial Heat Flux Profiles (AHFPs) were considered for six different fuels (UO2, ThO 2, MOX, UC2, UC, and UN) and at average and maximum channel power. Bulk-fluid, sheath, and fuel centerline temperatures as well as the Heat Transfer Coefficient (HTC) profiles were obtained along the fuel-channel length. The HTC values are within a range of 4.7--20 kW/m2·K and 9.7--10 kW/m2·K for the SCW and SHS channels respectively. The main conclusion is that while all the mentioned fuels may be used for the SHS channel, only UC2, UC, or UN are suitable for a SCW channel, because their fuel centerline temperatures are at least 1000°C below melting point, while that of UO2, ThO2 , and MOX may reach melting point.

  6. A CANDU-Based Fast Irradiation Reactor

    SciTech Connect

    Shatilla, Youssef

    2006-07-01

    A new steady-state fast neutron reactor is needed to satisfy the testing needs of Generation IV reactors, the Space Propulsion Program, and the Advanced Fuel Cycle Initiative. This paper presents a new concept for a CANDU-based fast irradiation reactor that is horizontal in orientation, with individual pressure tubes running the entire length of the scattering-medium tank (Calandria) filled with Lead-Bismuth-Eutectic (LBE). This approach for a test reactor will provide more flexibility in refueling, sample removal, and ability to completely re-configure the core to meet different users' requirements. Full core neutronic analysis of several fuel/coolant/geometry combinations showed a small hexagonal, LBE-cooled, U-Pu-10Zr fuel, with a core power of 100 MW{sub th} produced a fast flux (>0.1 MeV) of 1.5 x 10{sup 15} n/cm{sup 2} sec averaged over the whole length of six irradiation channels with a total testing volume of more than 77 liters. In-core breeding allowed the Pu-239 enrichment to be 15.3% which should result in core continuous operation for 180 effective full power days. Other coolants investigated included high pressure water steam and helium. An innovative shutdown/control system which consisted of the six outermost fuel channels was proven to be effective in shutting the core down when flooded with boric acid as a neutron absorber. The new shutdown/control system has the advantage of causing the minimum perturbation of the axial flux shape when the control channels are partially flooded with boric acid. This is because the acid is injected homogeneously along the control channel in contrast to regular control rods that are injected partially causing an axial perturbation in the core flux which in turn reduces safety analysis margins. The new shutdown/control system is not required to penetrate the core in a direction vertical to the fuel channels which allowed the freedom of changing core pitch as deemed necessary. A preliminary thermal hydraulic analysis

  7. Irradiation Facilities at the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2005-12-01

    The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC – formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950s with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world’s data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities1. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens.

  8. Survey for the presence of Naegleria fowleri amebae in lake water used to cool reactors at a nuclear power generating plant.

    PubMed

    Jamerson, Melissa; Remmers, Kenneth; Cabral, Guy; Marciano-Cabral, Francine

    2009-04-01

    Water from Lake Anna in Virginia, a lake that is used to cool reactors at a nuclear power plant and for recreational activities, was assessed for the presence of Naegleria fowleri, an ameba that causes primary amebic meningoencephalitis (PAM). This survey was undertaken because it has been reported that thermally enriched water fosters the propagation of N. fowleri and, hence, increases the risk of infection to humans. Of 16 sites sampled during the summer of 2007, nine were found to be positive for N. fowleri by a nested polymerase chain reaction assay. However, total ameba counts, inclusive of N. fowleri, never exceeded 12/50 mL of lake water at any site. No correlation was obtained between the conductivity, dissolved oxygen, temperature, and pH of water and presence of N. fowleri. To date, cases of PAM have not been reported from this thermally enriched lake. It is postulated that predation by other protozoa and invertebrates, disturbance of the water surface from recreational boating activities, or the presence of bacterial or fungal toxins, maintain the number N. fowleri at a low level in Lake Anna.

  9. The Very High Temperature Reactor

    SciTech Connect

    Hans D. Gougar; David A. Petti

    2011-06-01

    The High Temperature Reactor (HTR) and Very High Temperature Reactor (VHTR) are types of nuclear power plants that, as the names imply, operate at temperatures above those of the conventional nuclear power plants that currently generate electricity in the US and other countries. Like existing nuclear plants, heat generated from the fission of uranium or plutonium atoms is carried off by a working fluid and can be used generate electricity. The very hot working fluid also enables the VHTR to drive other industrial processes that require high temperatures not achievable by conventional nuclear plants (Figure 1). For this reason, the VHTR is being considered for non-electrical energy applications. The reactor and power conversion system are constructed using special materials that make a core meltdown virtually impossible.

  10. Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 5: Graphite PIRTs

    SciTech Connect

    Burchell, Timothy D; Bratton, Rob; Marsden, Barry; Srinivasan, Makuteswara; Penfield, Scott; Mitchell, Mark; Windes, Will

    2008-03-01

    Here we report the outcome of the application of the Nuclear Regulatory Commission (NRC) Phenomena Identification and Ranking Table (PIRT) process to the issue of nuclear-grade graphite for the moderator and structural components of a next generation nuclear plant (NGNP), considering both routine (normal operation) and postulated accident conditions for the NGNP. The NGNP is assumed to be a modular high-temperature gas-cooled reactor (HTGR), either a gas-turbine modular helium reactor (GTMHR) version [a prismatic-core modular reactor (PMR)] or a pebble-bed modular reactor (PBMR) version [a pebble bed reactor (PBR)] design, with either a direct- or indirect-cycle gas turbine (Brayton cycle) system for electric power production, and an indirect-cycle component for hydrogen production. NGNP design options with a high-pressure steam generator (Rankine cycle) in the primary loop are not considered in this PIRT. This graphite PIRT was conducted in parallel with four other NRC PIRT activities, taking advantage of the relationships and overlaps in subject matter. The graphite PIRT panel identified numerous phenomena, five of which were ranked high importance-low knowledge. A further nine were ranked with high importance and medium knowledge rank. Two phenomena were ranked with medium importance and low knowledge, and a further 14 were ranked medium importance and medium knowledge rank. The last 12 phenomena were ranked with low importance and high knowledge rank (or similar combinations suggesting they have low priority). The ranking/scoring rationale for the reported graphite phenomena is discussed. Much has been learned about the behavior of graphite in reactor environments in the 60-plus years since the first graphite rectors went into service. The extensive list of references in the Bibliography is plainly testament to this fact. Our current knowledge base is well developed. Although data are lacking for the specific grades being considered for Generation IV (Gen IV

  11. Hadronic event generation for hadron cascade calculations and detector simulation, Part IV: The application of the intranuclear cascade model to reactions of pions, nucleons, kaons, and their antiparticles with nuclei below 6 GeV/c

    SciTech Connect

    Haenbssgen, K.

    1987-02-01

    An extension of the intranuclear cascade model is described. The primary hadrons may be pions, kaons, nucleons, and their antiparticles. Secondary particles produced include hyperons or antihyperons. A large amount of experimental data is described by the model. The model is constructed via the Monte Carlo generation of complete events, based on a model of the nucleus structure and the hadron/nucleon interaction inside the nucleus. Calculated average multiplicities and single and double differential cross sections are compared with experimental data.

  12. INFLUENCE OF NATURAL AND SYNTHETIC ORGANIC LIGANDS ON THE STABILITY AND MOBILITY OF REDUCED TC(IV)

    SciTech Connect

    Nathalie A. Wall; Baohua Gu

    2012-12-20

    The primary objectives were (1) to quantify the interactions of organic ligands with Tc(IV) through the generation of thermodynamic (complexation) and kinetic parameters needed to assess and predict the mobility of reduced Tc(IV) at DOE contaminated sites; and (2) to determine the impact of organic ligands on the mobility and fate of reduced Tc(IV) under field geochemical conditions.

  13. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1959-12-15

    A reactor which is particularly adapted tu serve as a heat source for a nuclear powered alrcraft or rocket is described. The core of this reactor consists of a porous refractory modera;or body which is impregnated with fissionable nuclei. The core is designed so that its surface forms tapered inlet and outlet ducts which are separated by the porous moderator body. In operation a gaseous working fluid is circulated through the inlet ducts to the surface of the moderator, enters and passes through the porous body, and is heated therein. The hot gas emerges into the outlet ducts and is available to provide thrust. The principle advantage is that tremendous quantities of gas can be quickly heated without suffering an excessive pressure drop.

  14. NUCLEAR REACTOR

    DOEpatents

    Treshow, M.

    1958-08-19

    A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

  15. REACTOR MONITORING

    DOEpatents

    Bugbee, S.J.; Hanson, V.F.; Babcock, D.F.

    1959-02-01

    A neutron density inonitoring means for reactors is described. According to this invention a tunnel is provided beneath and spaced from the active portion of the reactor and extends beyond the opposite faces of the activc portion. Neutron beam holes are provided between the active portion and the tunnel and open into the tunnel near the middle thereof. A carriage operates back and forth in the tunnel and is adapted to convey a neutron detector, such as an ion chamber, and position it beneath one of the neutron beam holes. This arrangement affords convenient access of neutron density measuring instruments to a location wherein direct measurement of neutron density within the piles can be made and at the same time affords ample protection to operating personnel.

  16. NEUTRONIC REACTOR

    DOEpatents

    Wade, E.J.

    1958-09-16

    This patent relates to a reflector means for a neutronic reactor. A reflector comprised of a plurality of vertically movable beryllium control members is provided surrounding the sides of the reactor core. An absorber of fast neutrons comprised of natural uramum surrounds the reflector. An absorber of slow neutrons surrounds the absorber of fast neutrons and is formed of a plurality of beryllium blocks having natural uranium members distributcd therethrough. in addition, a movable body is positioned directly below the core and is comprised of a beryllium reflector and an absorbing member attached to the botiom thereof, the absorbing member containing a substance selected from the goup consisting of natural urantum and Th/sup 232/.

  17. Nuclear reactor

    DOEpatents

    Wade, Elman E.

    1979-01-01

    A nuclear reactor including two rotatable plugs and a positive top core holddown structure. The top core holddown structure is divided into two parts: a small core cover, and a large core cover. The small core cover, and the upper internals associated therewith, are attached to the small rotating plug, and the large core cover, with its associated upper internals, is attached to the large rotating plug. By so splitting the core holddown structures, under-the-plug refueling is accomplished without the necessity of enlarging the reactor pressure vessel to provide a storage space for the core holddown structure during refueling. Additionally, the small and large rotating plugs, and their associated core covers, are arranged such that the separation of the two core covers to permit rotation is accomplished without the installation of complex lifting mechanisms.

  18. REACTOR UNLOADING

    DOEpatents

    Leverett, M.C.

    1958-02-18

    This patent is related to gas cooled reactors wherein the fuel elements are disposed in vertical channels extending through the reactor core, the cooling gas passing through the channels from the bottom to the top of the core. The invention is a means for unloading the fuel elements from the core and comprises dump values in the form of flat cars mounted on wheels at the bottom of the core structure which support vertical stacks of fuel elements. When the flat cars are moved, either manually or automatically, for normal unloading purposes, or due to a rapid rise in the reproduction ratio within the core, the fuel elements are permtted to fall by gravity out of the core structure thereby reducing the reproduction ratio or stopping the reaction as desired.

  19. Neutronic reactor

    DOEpatents

    Lewis, Warren R.

    1978-05-30

    A graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels.

  20. NUCLEAR REACTORS

    DOEpatents

    Koch, L.J.; Rice, R.E. Jr.; Denst, A.A.; Rogers, A.J.; Novick, M.

    1961-12-01

    An active portion assembly for a fast neutron reactor is described wherein physical distortions resulting in adverse changes in the volume-to-mass ratio are minimized. A radially expandable locking device is disposed within a cylindrical tube within each fuel subassembly within the active portion assembly, and clamping devices expandable toward the center of the active portion assembly are disposed around the periphery thereof. (AEC)

  1. NUCLEAR REACTORS

    DOEpatents

    Long, E.; Ashby, J.W.

    1958-09-16

    ABS>A graphite moderator structure is presented for a nuclear reactor compriscd of an assembly of similarly orientated prismatic graphite blocks arranged on spaced longitudinal axes lying in common planes wherein the planes of the walls of the blocks are positioned so as to be twisted reintive to the planes of said axes so thatthe unlmpeded dtrect paths in direction wholly across the walls of the blocks are limited to the width of the blocks plus spacing between the blocks.

  2. NEUTRONIC REACTORS

    DOEpatents

    Anderson, J.B.

    1960-01-01

    A reactor is described which comprises a tank, a plurality of coaxial steel sleeves in the tank, a mass of water in the tank, and wire grids in abutting relationship within a plurality of elongated parallel channels within the steel sleeves, the wire being provided with a plurality of bends in the same plane forming adjacent parallel sections between bends, and the sections of adjacent grids being normally disposed relative to each other.

  3. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1961-01-24

    A core structure for neutronic reactors adapted for the propulsion of aircraft and rockets is offered. The core is designed for cooling by gaseous media, and comprises a plurality of hollow tapered tubular segments of a porous moderating material impregniated with fissionable fuel nested about a common axis. Alternate ends of the segments are joined. In operation a coolant gas passes through the porous structure and is heated.

  4. Confirmatory factor analysis of the WAIS-IV/WMS-IV.

    PubMed

    Holdnack, James A; Xiaobin Zhou; Larrabee, Glenn J; Millis, Scott R; Salthouse, Timothy A

    2011-06-01

    The Wechsler Adult Intelligence Scale-fourth edition (WAIS-IV) and the Wechsler Memory Scale-fourth edition (WMS-IV) were co-developed to be used individually or as a combined battery of tests. The independent factor structure of each of the tests has been identified; however, the combined factor structure has yet to be determined. Confirmatory factor analysis was applied to the WAIS-IV/WMS-IV Adult battery (i.e., age 16-69 years) co-norming sample (n = 900) to test 13 measurement models. The results indicated that two models fit the data equally well. One model is a seven-factor solution without a hierarchical general ability factor: Verbal Comprehension, Perceptual Reasoning, Processing Speed, Auditory Working Memory, Visual Working Memory, Auditory Memory, and Visual Memory. The second model is a five-factor model composed of Verbal Comprehension, Perceptual Reasoning, Processing Speed, Working Memory, and Memory with a hierarchical general ability factor. Interpretative implications for each model are discussed.

  5. Spherical torus fusion reactor

    DOEpatents

    Peng, Yueng-Kay M.

    1989-04-04

    A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

  6. Spherical torus fusion reactor

    DOEpatents

    Peng, Yueng-Kay M.

    1989-01-01

    A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

  7. Fusion reactor pumped laser

    DOEpatents

    Jassby, D.L.

    1987-09-04

    A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam. 10 figs.

  8. REACTOR CONTROL

    DOEpatents

    Ruano, W.J.

    1957-12-10

    This patent relates to nuclear reactors of the type which utilize elongited rod type fuel elements immersed in a liquid moderator and shows a design whereby control of the chain reaction is obtained by varying the amount of moderator or reflector material. A central tank for containing liquid moderator and fuel elements immersed therein is disposed within a surrounding outer tank providing an annular space between the two tanks. This annular space is filled with liquid moderator which functions as a reflector to reflect neutrons back into the central reactor tank to increase the reproduction ratio. Means are provided for circulating and cooling the moderator material in both tanks and additional means are provided for controlling separately the volume of moderator in each tank, which latter means may be operated automatically by a neutron density monitoring device. The patent also shows an arrangement for controlling the chain reaction by injecting and varying an amount of poisoning material in the moderator used in the reflector portion of the reactor.

  9. Unique features of space reactors

    SciTech Connect

    Buden, D.

    1990-01-01

    Space reactors are designed to meet a unique set of requirements; they must be sufficiently compact to be launched in a rocket to their operational location, operate for many years without maintenance and servicing, operate in extreme environments, and reject heat by radiation to space. To meet these restrictions, operating temperatures are much greater than in terrestrial power plants, and the reactors tend to have a fast neutron spectrum. Currently, a new generation of space reactor power plants is being developed. The major effort is in the SP-100 program, where the power plant is being designed for seven years of full power, and no maintenance operation at a reactor outlet operating temperature of 1350 K. 8 refs., 3 figs., 1 tab.

  10. Overview of fusion reactor safety

    NASA Astrophysics Data System (ADS)

    Cohen, S.; Crocker, J. G.

    Use of deuterium-tritium fusion reactors requires examination of several major safety and environmental issues: (1) tritium inventory control; (2) neutron activation of structural materials, fluid streams and reactor hall environment; (3) release of radioactivity from energy sources including lithium spill reactions, superconducting magnet stored energy release, and plasma disruptions; (4) high magnetic and electromagnetic fields associated with fusion reactor superconducting magnets and radio frequency heating devices; and (5) handling and disposal of radioactive waste. Early recognition of potential safety problems with fusion reactors provides the opportunity for improvement in design and materials to eliminate or greatly reduce these problems. With an early start in this endeavor, fusion should be among the lower risk technologies for generation of commercial electrical power.

  11. Proliferation resistance for fast reactors and related fuel cycles: issues and impacts

    SciTech Connect

    Pilat, Joseph F

    2010-01-01

    The prospects for a dramatic growth in nuclear power may depend to a significant degree on the effectiveness of, and the resources devoted to, plans to develop and implement technologies and approaches that strengthen proliferation resistance and nuclear materials accountability. The challenges for fast reactors and related fuel cycles are especially critical. They are being explored in the Generation IV Tnternational Forum (GIF) and the Tnternational Atomic Energy Agency's (IAEA's) International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) initiative, as well as by many states that are looking to these systems for the efficient lise of uranium resources and long-term energy security. How do any proliferation risks they may pose compare to other reactors, both existing and under development, and their fuel cycles? Can they be designed with intrinsic (technological) features to make these systems more proliferation resistant? What roles can extrinsic (institutional) features play in proliferation resistance? What are the anticipated safeguards requirements, and will new technologies and approaches need to be developed? How can safeguards be facilitated by the design process? These and other questions require a rethinking of proliferation resistance and the prospects for new technologies and other intrinsic and extrinsic features being developed that are responsive to specific issues for fast reactors and related fuel cycles and to the broader threat environment in which these systems will have to operate. There are no technologies that can wholly eliminate the risk of proliferation by a determined state, but technology and design can playa role in reducing state threats and perhaps in eliminating non-state threats. There will be a significant role for extrinsic factors, especially the various measures - from safeguards and physical protection to export controls - embodied in the international nuclear nonproliferation regime. This paper will offer

  12. Advanced PPA Reactor and Process Development

    NASA Technical Reports Server (NTRS)

    Wheeler, Raymond; Aske, James; Abney, Morgan B.; Miller, Lee A.; Greenwood, Zachary

    2012-01-01

    Design and development of a second generation Plasma Pyrolysis Assembly (PPA) reactor is currently underway as part of NASA s Atmosphere Revitalization Resource Recovery effort. By recovering up to 75% of the hydrogen currently lost as methane in the Sabatier reactor effluent, the PPA helps to minimize life support resupply costs for extended duration missions. To date, second generation PPA development has demonstrated significant technology advancements over the first generation device by doubling the methane processing rate while, at the same time, more than halving the required power. One development area of particular interest to NASA system engineers is fouling of the PPA reactor with carbonaceous products. As a mitigation plan, NASA MSFC has explored the feasibility of using an oxidative plasma based upon metabolic CO2 to regenerate the reactor window and gas inlet ports. The results and implications of this testing are addressed along with the advanced PPA reactor development work.

  13. Biofilm Community Dynamics in Bench-Scale Annular Reactors Simulating Arrestment of Chloraminated Drinking Water Nitrification

    EPA Science Inventory

    Annular reactors (ARs) were used to study biofilm community succession and provide an ecological insight during nitrification arrestment through simultaneously increasing monochloramine (NH2Cl) and chlorine to nitrogen mass ratios, resulting in four operational periods (I to IV)....

  14. Scram signal generator

    DOEpatents

    Johanson, Edward W.; Simms, Richard

    1981-01-01

    A scram signal generating circuit for nuclear reactor installations monitors a flow signal representing the flow rate of the liquid sodium coolant which is circulated through the reactor, and initiates reactor shutdown for a rapid variation in the flow signal, indicative of fuel motion. The scram signal generating circuit includes a long-term drift compensation circuit which processes the flow signal and generates an output signal representing the flow rate of the coolant. The output signal remains substantially unchanged for small variations in the flow signal, attributable to long term drift in the flow rate, but a rapid change in the flow signal, indicative of a fast flow variation, causes a corresponding change in the output signal. A comparator circuit compares the output signal with a reference signal, representing a given percentage of the steady state flow rate of the coolant, and generates a scram signal to initiate reactor shutdown when the output signal equals the reference signal.

  15. Scram signal generator

    DOEpatents

    Johanson, E.W.; Simms, R.

    A scram signal generating circuit for nuclear reactor installations monitors a flow signal representing the flow rate of the liquid sodium coolant which is circulated through the reactor, and initiates reactor shutdown for a rapid variation in the flow signal, indicative of fuel motion. The scram signal generating circuit includes a long-term drift compensation circuit which processes the flow signal and generates an output signal representing the flow rate of the coolant. The output signal remains substantially unchanged for small variations in the flow signal, attributable to long term drift in the flow rate, but a rapid change in the flow signal, indicative of a fast flow variation, causes a corresponding change in the output signal. A comparator circuit compares the output signal with a reference signal, representing a given percentage of the steady state flow rate of the coolant, and generates a scram signal to initiate reactor shutdown when the output signal equals the reference signal.

  16. Research Program of a Super Fast Reactor

    SciTech Connect

    Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie; Terai, Takayuki; Nagasaki, Shinya; Muroya, Yusa; Abe, Hiroaki; Akiba, Masato; Akimoto, Hajime; Okumura, Keisuke; Akasaka, Naoaki; GOTO, Shoji

    2006-07-01

    Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is not breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)

  17. Mars, the Moon, and the Ends of the Earth: Autonomy for Small Reactor Power Systems

    SciTech Connect

    Wood, Richard Thomas

    2008-01-01

    . Additionally, many Generation IV (Gen IV) reactor concepts have goals for optimizing investment recovery and economic efficiency that promote significant reductions in plant operations and maintenance staff over current-generation nuclear power plants. To accomplish these Gen IV goals and also address the SRPS remote-siting challenges, higher levels of automation, fault tolerance, and advanced diagnostic capabilities are needed to provide nearly autonomous operations with anticipatory maintenance. Essentially, the SRPS control system for several anticipated terrestrial applications can benefit from the kind of operational autonomy that is necessary for deep space and planetary SRPS-enabled missions. Investigation of the state of the technology for autonomous control confirmed that control systems with varying levels of autonomy have been employed in robotic, transportation, spacecraft, and manufacturing applications. As an example, NASA has pursued autonomy for spacecraft and surface exploration vehicles (e.g., rovers) to reduce mission costs, increase efficiency for communications between ground control and the vehicle, and enable independent operation of the vehicle during times of communications blackout. However, autonomous control has not been implemented for an operating terrestrial nuclear power plant nor has there been any experience beyond automating simple control loops for space reactors. Current automated control technologies for nuclear power plants are reasonably mature, and fully automated control of normal SRPS operations is clearly feasible. However, the space-based and remote terrestrial applications of SRPS modules require autonomous capabilities that can accommodate nonoptimum operations when degradation, failure, and other off-normal events challenge the performance of the reactor while immediate human intervention is not possible. The independent action provided by autonomous control, which is distinct from the more limited self action of automated

  18. MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR

    DOEpatents

    Balent, R.

    1963-03-12

    This patent describes a method of obtaining a flatter flux and more uniform power generation across the core of a nuclear reactor. The method comprises using moderator elements having differing moderating strength. The elements have an increasing amount of the better moderating material as a function of radial and/or axial distance from the reactor core center. (AEC)

  19. BJN Awards 2016: IV therapy.

    PubMed

    Rickard, Claire

    2016-07-28

    Claire Rickard Professor of Nursing, National Health and Medical Research Council (NHMRC) Centre of Research Excellence in Nursing, Griffith University, was awarded second place in the BJN Awards 2016 for IV Therapy Nurse of the Year. Here she talks about the she has done to be recognised in this field. PMID:27467655

  20. The PLATO IV Communications System.

    ERIC Educational Resources Information Center

    Sherwood, Bruce Arne; Stifle, Jack

    The PLATO IV computer-based educational system contains its own communications hardware and software for operating plasma-panel graphics terminals. Key echoing is performed by the central processing unit: every key pressed at a terminal passes through the entire system before anything appears on the terminal's screen. Each terminal is guaranteed…

  1. Baseline Concept Description of a Small Modular High Temperature Reactor

    SciTech Connect

    Gougar, Hans D.

    2014-10-01

    Generation IV program and its specific R&D needs will be included in this report when appropriate for comparison. The distinguishing features of the HTGR are the refractory (TRISO) coated particle fuel, the low-power density, graphite-moderated core, and the high outlet temperature of the inert helium coolant. The low power density and fuel form effectively eliminate the possibility of core melt, even upon a complete loss of coolant pressure and flow. The graphite, which constitutes the bulk of the core volume and mass, provides a large thermal buffer that absorbs fission heat such that thermal transients occur over a timespan of hours or even days. As chemically-inert helium is already a gas, there is no coolant temperature or void feedback on the neutronics and no phase change or corrosion product that could degrade heat transfer. Furthermore, the particle coatings and interstitial graphite retain fission products such that the source terms at the plant boundary remain well below actionable levels under all anticipated nominal and off-normal operating conditions. These attributes enable the reactor to supply process heat to a collocated industrial plant with negligible risk of contamination and minimal dynamic coupling of the facilities (Figure 1). The exceptional retentive properties of coated particle fuel in a graphite matrix were first demonstrated in the DRAGON reactor, a European research facility that began operation in 1964.

  2. Baseline Concept Description of a Small Modular High Temperature Reactor

    SciTech Connect

    Hans Gougar

    2014-05-01

    Generation IV program and its specific R&D needs will be included in this report when appropriate for comparison. The distinguishing features of the HTGR are the refractory (TRISO) coated particle fuel, the low-power density, graphite-moderated core, and the high outlet temperature of the inert helium coolant. The low power density and fuel form effectively eliminate the possibility of core melt, even upon a complete loss of coolant pressure and flow. The graphite, which constitutes the bulk of the core volume and mass, provides a large thermal buffer that absorbs fission heat such that thermal transients occur over a timespan of hours or even days. As chemically-inert helium is already a gas, there is no coolant temperature or void feedback on the neutronics and no phase change or corrosion product that could degrade heat transfer. Furthermore, the particle coatings and interstitial graphite retain fission products such that the source terms at the plant boundary remain well below actionable levels under all anticipated nominal and off-normal operating conditions. These attributes enable the reactor to supply process heat to a collocated industrial plant with negligible risk of contamination and minimal dynamic coupling of the facilities (Figure 1). The exceptional retentive properties of coated particle fuel in a graphite matrix were first demonstrated in the DRAGON reactor, a European research facility that began operation in 1964.

  3. Design Option of Heat Exchanger for the Next Generation Nuclear Plant

    SciTech Connect

    Eung Soo Kim; Chang Oh

    2008-09-01

    The Next Generation Nuclear Plant (NGNP), a very High temperature Gas-Cooled Reactor (VHTGRS) concept, will provide the first demonstration of a closed-loop Brayton cycle at a commercial scale of a few hundred megawatts electric and hydrogen production. The power conversion system (PCS) for the NGNP will take advantage of the significantly higher reactor outlet temperatures of the VHTGRS to provide higher efficiencies than can be achieved in the current generation of light water reactors. Besides demonstrating a system design that can be used directly for subsequent commercial deployment, the NGNP will demonstrate key technology elements that can be used in subsequent advanced power conversion systems for other Generation IV reactors. In anticipation of the design, development and procurement of an advanced power conversion system for the NGNP, the system integration of the NGNP and hydrogen plant was initiated to identify the important design and technology options that must be considered in evaluating the performance of the proposed NGNP. As part of the system integration of the VHTGRS and hydrogen production plant, the intermediate heat exchanger is used to transfer the process heat from VHTGRS to hydrogen plant. Therefore, the design and configuration of the intermediate heat exchanger are very important. This paper will include analysis of one stage versus two stage heat exchanger design configurations and thermal stress analyses of a printed circuit heat exchanger, helical coil heat exchanger, and shell/tube heat exchanger.

  4. Hydrogen generation via anaerobic fermentation of paper mill wastes.

    PubMed

    Valdez-Vazquez, Idania; Sparling, Richard; Risbey, Derek; Rinderknecht-Seijas, Noemi; Poggi-Varaldo, Héctor M

    2005-11-01

    The objective of this work was to determine the hydrogen production from paper mill wastes using microbial consortia of solid substrate anaerobic digesters. Inocula from mesophilic, continuous solid substrate anaerobic digestion (SSAD) reactors were transferred to small lab scale, batch reactors. Milled paper (used as a surrogate paper waste) was added as substrate and acetylene or 2-bromoethanesulfonate (BES) was spiked for methanogenesis inhibition. In the first phase of experiments it was found that acetylene at 1% v/v in the headspace was as effective as BES in inhibiting methanogenic activity. Hydrogen gas accumulated in the headspace of the bottles, reaching a plateau. Similar final hydrogen concentrations were obtained for reactors spiked with acetylene and BES. In the second phase of tests the headspace of the batch reactors was flushed with nitrogen gas after the first plateau of hydrogen was reached, and subsequently incubated, with no further addition of inhibitor nor substrate. It was found that hydrogen production resumed and reached a second plateau, although somewhat lower than the first one. This procedure was repeated a third time and an additional amount of hydrogen was obtained. The plateaux and initial rates of hydrogen accumulation decreased in each subsequent incubation cycle. The total cumulative hydrogen harvested in the three cycles was much higher (approx. double) than in the first cycle alone. We coined this procedure as IV-SSAH (intermittently vented solid substrate anaerobic hydrogen generation). Our results point out to a feasible strategy for obtaining higher hydrogen yields from the fermentation of industrial solid wastes, and a possible combination of waste treatment processes consisting of a first stage IV-SSAH followed by a second SSAD stage. Useful products of this approach would be hydrogen, organic acids or methane, and anaerobic digestates that could be used as soil amenders after post-treatment.

  5. Nuclear Reactors. Revised.

    ERIC Educational Resources Information Center

    Hogerton, John F.

    This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: How Reactors Work; Reactor Design; Research, Teaching, and Materials Testing; Reactors (Research, Teaching and Materials); Production Reactors; Reactors for Electric Power…

  6. Nuclear reactor

    DOEpatents

    Pennell, William E.; Rowan, William J.

    1977-01-01

    A nuclear reactor in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assmblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters and the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters and the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance.

  7. Monitoring and control requirement definition study for dispersed storage and generation (DSG). Volume IV. Final report, Appendix C: identification from utility visits of present and future approaches to integration of DSG into distribution networks

    SciTech Connect

    Not Available

    1980-10-01

    A major aim of the US National Energy Policy, as well as that of the New York State Energy Research and Development Authority, is to conserve energy and to shift from oil to more abundant domestic fuels and renewable energy sources. Dispersed Storage and Generation (DSG) is the term that characterizes the present and future dispersed, relatively small (<30 MW) energy systems, such as solar thermal electric, photovoltaic, wind, fuel cell, storage battery, hydro, and cogeneration, which can help achieve these national energy goals and can be dispersed throughout the distribution portion of an electric utility system. As a result of visits to four utilities concerned with the use of DSG power sources on their distribution networks, some useful impressions of present and future approaches to the integration of DSGs into electrical distribution network have been obtained. A more extensive communications and control network will be developed by utilities for control of such sources for future use. Different approaches to future utility systems with DSG are beginning to take shape. The new DSG sources will be in decentralized locations with some measure of centralized control. The utilities have yet to establish firmly the communication and control means or their organization. For the present, the means for integrating the DSGs and their associated monitoring and control equipment into a unified system have not been decided.

  8. REACTOR COMPONETN

    DOEpatents

    Creutz, E.C.

    1959-10-27

    A reactor fuel element comprised of a slug of fissionable material disposed in a sheath of corrosion resistantmaterial is described. The sheath is in the form of a tubular container closed at one end and is in tight-fitting engagement with the peripheral sunface of the slug. An inner cap is insented into the open end of the sheath against the slug, which end is then bent around the inner cap and welded thereto. An outer cap is then welded around its peripheny to the bent portion of the container.

  9. NUCLEAR REACTORS

    DOEpatents

    Long, E.; Ashley, J.W.

    1958-12-16

    A graphite moderator structure is described for a gas-cooled nuclear reactor having a vertical orlentation wherein the structure is physically stable with regard to dlmensional changes due to Wigner growth properties of the graphite, and leakage of coolant gas along spaces in the structure is reduced. The structure is comprised of stacks of unlform right prismatic graphite blocks positioned in layers extending in the direction of the lengths of the blocks, the adjacent end faces of the blocks being separated by pairs of tiles. The blocks and tiles have central bores which are in alignment when assembled and are provided with cooperatlng keys and keyways for physical stability.

  10. Italian hybrid and fission reactors scenario analysis

    NASA Astrophysics Data System (ADS)

    Ciotti, M.; Manzano, J.; Sepielli, M.

    2012-06-01

    Italy is a country where a long tradition of studies both in the fission and fusion field is consolidated; nevertheless a strong public opinion concerned with the destination of the Spent Nuclear Fuel hinders the development of nuclear power. The possibility to a severe reduction of the NSF mass generated from a fleet of nuclear reactors employing an hypothetical fusionfission hybrid reactor has been investigated in the Italian framework. The possibility to produce nuclear fuel for the fission nuclear reactors with the hybrid reactor was analyzed too.

  11. Italian hybrid and fission reactors scenario analysis

    SciTech Connect

    Ciotti, M.; Manzano, J.; Sepielli, M.

    2012-06-19

    Italy is a country where a long tradition of studies both in the fission and fusion field is consolidated; nevertheless a strong public opinion concerned with the destination of the Spent Nuclear Fuel hinders the development of nuclear power. The possibility to a severe reduction of the NSF mass generated from a fleet of nuclear reactors employing an hypothetical fusionfission hybrid reactor has been investigated in the Italian framework. The possibility to produce nuclear fuel for the fission nuclear reactors with the hybrid reactor was analyzed too.

  12. Facile Routes to Th(IV), U(IV), and Np(IV) Phosphites and Phosphates

    SciTech Connect

    Villa, Eric M.; Wang, Shuao; Alekseev, Evgeny V.; Depmeier, Wulf; Albrecht-Schmitt, Thomas E.

    2011-08-05

    Three actinide(IV) phosphites and a NpIV phosphate, AnIV(HPO₃)₂(H₂O)₂ (An = Th, U, Np) and Cs[Np(H1.5PO₄)(PO₄)]₂, respectively, were synthesized using mild hydrothermal conditions. The first three phases are isotypic and were obtained using similar reaction conditions. Cs[Np(H1.5PO₄)(PO₄)]₂ was synthesized using an analogous method to that of Np(HPO₃)₂(H₂O)₂. However, this fourth phase is quite different in comparison to the other phases in both composition and structure. The structure of Cs[Np(H1.5PO₄)(PO₄)]₂ is constructed from double layers of neptunium(IV) phosphate with caesium cations in the interlayer region. In contrast, An(HPO₃)₂(H₂O)₂ (An = Th, U, Np) form dense 3D networks. The actinide contraction is detected in variety of metrics obtained from single-crystal X-ray diffraction data. Changes in the oxidation state of the neptunium starting materials yield different products.

  13. Complex oscillator and Painlevé IV equation

    SciTech Connect

    Fernández C, David J. González, J.C.

    2015-08-15

    Supersymmetric quantum mechanics is a powerful tool for generating exactly solvable potentials departing from a given initial one. In this article the first- and second-order supersymmetric transformations will be used to obtain new exactly solvable potentials departing from the complex oscillator. The corresponding Hamiltonians turn out to be ruled by polynomial Heisenberg algebras. By applying a mechanism to reduce to second the order of these algebras, the connection with the Painlevé IV equation is achieved, thus giving place to new solutions for the Painlevé IV equation.

  14. BOILER-SUPERHEATED REACTOR

    DOEpatents

    Heckman, T.P.

    1961-05-01

    A nuclear power reactor of the type in which a liquid moderator-coolant is transformed by nuclear heating into a vapor that may be used to drive a turbo- generator is described. The core of this reactor comprises a plurality of freely suspended tubular fuel elements, called fuel element trains, within which nonboiling pressurized liquid moderator-coolant is preheated and sprayed through orifices in the walls of the trains against the outer walls thereof to be converted into vapor. Passage of the vapor ovcr other unwetted portions of the outside of the fuel elements causes the steam to be superheated. The moderatorcoolant within the fuel elements remains in the liqUid state, and that between the fuel elements remains substantiaily in the vapor state. A unique liquid neutron-absorber control system is used. Advantages expected from the reactor design include reduced fuel element failure, increased stability of operation, direct response to power demand, and circulation of a minimum amount of liquid moderatorcoolant. (A.G.W.)

  15. PINCHED PLASMA REACTOR

    DOEpatents

    Phillips, J.A.; Suydam, R.; Tuck, J.L.

    1961-07-01

    BS>A plasma confining and heating reactor is described which has the form of a torus with a B/sub 2/ producing winding on the outside of the torus and a helical winding of insulated overlapping tunns on the inside of the torus. The inner helical winding performs the double function of shielding the plasma from the vitreous container and generating a second B/sub z/ field in the opposite direction to the first B/sub z/ field after the pinch is established.

  16. Advanced light water reactor requirements document: Chapter 3, Reactor coolant system and reactor non-safety auxiliary systems

    SciTech Connect

    Not Available

    1987-06-01

    The purpose of this chapter of the Advanced Light Water Reactor (ALWR) Plant Requirements Document is to establish utility requirements for the design of the Reactor Coolant System and the Reactor Non-safety Auxiliary Systems of Advanced LWR plants consistent with the objectives and principles of the ALWR program. The scope of this chapter covers the reactor coolant system and reactor non-safety auxiliary systems for Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). Non-safety auxiliaries include systems which are required for normal operation of the plant but are not required to operate for accident mitigation or to bring the plant to a safe shutdown condition. For PWRs, the reactor coolant system, steam generator system, chemical and volume control system and boron recycle system are included. For BWRs, the reactor coolant system and reactor water cleanup system are included. The chapter also includes requirements for the above systems which are common to BWRs and PWRs and requirements for process sampling for BWRs and PWRs.

  17. Nozzle for electric dispersion reactor

    DOEpatents

    Sisson, W.G.; Harris, M.T.; Scott, T.C.; Basaran, O.A.

    1996-04-02

    A nozzle for an electric dispersion reactor includes two coaxial cylindrical bodies, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode. 5 figs.

  18. Nozzle for electric dispersion reactor

    DOEpatents

    Sisson, W.G.; Basaran, O.A.; Harris, M.T.

    1995-11-07

    A nozzle for an electric dispersion reactor includes two concentric electrodes, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode. 4 figs.

  19. Nozzle for electric dispersion reactor

    DOEpatents

    Sisson, Warren G.; Basaran, Osman A.; Harris, Michael T.

    1995-01-01

    A nozzle for an electric dispersion reactor includes two concentric electrodes, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode.

  20. Nozzle for electric dispersion reactor

    DOEpatents

    Sisson, Warren G.; Harris, Michael T.; Scott, Timothy C.; Basaran, Osman A.

    1996-01-01

    A nozzle for an electric dispersion reactor includes two coaxial cylindrical bodies, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode.

  1. Nozzle for electric dispersion reactor

    DOEpatents

    Sisson, Warren G.; Basaran, Osman A.; Harris, Michael T.

    1998-01-01

    A nozzle for an electric dispersion reactor includes two concentric electrodes, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode.

  2. Nozzle for electric dispersion reactor

    DOEpatents

    Sisson, Warren G.; Harris, Michael T.; Scott, Timothy C.; Basaran, Osman A.

    1998-01-01

    A nozzle for an electric dispersion reactor includes two coaxial cylindrical bodies, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode.

  3. Nozzle for electric dispersion reactor

    DOEpatents

    Sisson, W.G.; Basaran, O.A.; Harris, M.T.

    1998-04-14

    A nozzle for an electric dispersion reactor includes two concentric electrodes, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode. 4 figs.

  4. Nozzle for electric dispersion reactor

    DOEpatents

    Sisson, W.G.; Harris, M.T.; Scott, T.C.; Basaran, O.A.

    1998-06-02

    A nozzle for an electric dispersion reactor includes two coaxial cylindrical bodies, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode. 5 figs.

  5. Thermal criteria to compare fast reactors coolants for the intermediate loop

    SciTech Connect

    Saez, Manuel; Rodriguez, Gilles

    2007-07-01

    Fast Breeder Reactors (FBR) are typically using a liquid metal as the primary coolant. Up to now, sodium is the referenced coolant for all large-scale FBR, but lead and sodium-potassium alloy have both also been used successfully for smaller rigs. The French Atomic Energy Commission (CEA) has an extensive experience and significant expertise in Sodium cooled Fast Reactors (SFR) over the past 40 years of R and D and feedback experiments. Some improvements are needed on the SFR to meet the Generation IV goals, and in particular the safety and the reliability through the intermediate loop coolant. As sodium reacts exo-thermically with air and water and to eliminate the drawback of the water-sodium interaction when a steam generator tube is ruptured, CEA is involved in a substantial effort in order to investigate the interest to use an alternative coolant than sodium in the intermediate loop. This paper presents the main thermal criteria to compare Fast Reactors coolants for the intermediate loop under natural and forced convection. Neutronics considerations are not taken into account for the intermediate loop coolant. Transport, transfer and energetic criteria are analysed in the field of turbulent flows. Criteria are applied to the following potential coolant candidates: sodium, lithium, tin, bismuth, lead, lead-bismuth alloy, lead-lithium alloy, gallium, indium, potassium and sodium-potassium alloy. According to this thermal analysis, the gallium as heat transfer agent for the intermediate loop is considered as a promising candidate. For the discussion of the applicability of the gallium as heat transfer agent for the intermediate loop, a limited thermal hydraulic pre-sizing of a steam generator is undertaken using simple engineering methods implemented in COPERNIC code, a CEA tool dedicated to reactor systems pre-sizing. (authors)

  6. Zirconium Hydride Space Power Reactor design.

    NASA Technical Reports Server (NTRS)

    Asquith, J. G.; Mason, D. G.; Stamp, S.

    1972-01-01

    The Zirconium Hydride Space Power Reactor being designed and fabricated at Atomics International is intended for a wide range of potential applications. Throughout the program a series of reactor designs have been evaluated to establish the unique requirements imposed by coupling with various power conversion systems and for specific applications. Current design and development emphasis is upon a 100 kilowatt thermal reactor for application in a 5 kwe thermoelectric space power generating system, which is scheduled to be fabricated and ground tested in the mid 70s. The reactor design considerations reviewed in this paper will be discussed in the context of this 100 kwt reactor and a 300 kwt reactor previously designed for larger power demand applications.

  7. Hybrid adsorptive membrane reactor

    NASA Technical Reports Server (NTRS)

    Tsotsis, Theodore T. (Inventor); Sahimi, Muhammad (Inventor); Fayyaz-Najafi, Babak (Inventor); Harale, Aadesh (Inventor); Park, Byoung-Gi (Inventor); Liu, Paul K. T. (Inventor)

    2011-01-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  8. Hybrid adsorptive membrane reactor

    DOEpatents

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  9. Radiation Damage In Reactor Cavity Concrete

    SciTech Connect

    Field, Kevin G; Le Pape, Yann; Naus, Dan J; Remec, Igor; Busby, Jeremy T; Rosseel, Thomas M; Wall, Dr. James Joseph

    2015-01-01

    License renewal up to 60 years and the possibility of subsequent license renewal to 80 years has established a renewed focus on long-term aging of nuclear generating stations materials, and recently, on concrete. Large irreplaceable sections of most nuclear generating stations include concrete. The Expanded Materials Degradation Analysis (EMDA), jointly performed by the Department of Energy, the Nuclear Regulatory Commission and Industry, identified the urgent need to develop a consistent knowledge base on irradiation effects in concrete. Much of the historical mechanical performance data of irradiated concrete does not accurately reflect typical radiation conditions in NPPs or conditions out to 60 or 80 years of radiation exposure. To address these potential gaps in the knowledge base, The Electric Power Research Institute and Oak Ridge National Laboratory are working to disposition radiation damage as a degradation mechanism. This paper outlines the research program within this pathway including: (i) defining the upper bound of the neutron and gamma dose levels expected in the biological shield concrete for extended operation (80 years of operation and beyond), (ii) determining the effects of neutron and gamma irradiation as well as extended time at temperature on concrete, (iii) evaluating opportunities to irradiate prototypical concrete under accelerated neutron and gamma dose levels to establish a conservative bound and share data obtained from different flux, temperature, and fluence levels, (iv) evaluating opportunities to harvest and test irradiated concrete from international NPPs, (v) developing cooperative test programs to improve confidence in the results from the various concretes and research reactors, (vi) furthering the understanding of the effects of radiation on concrete (see companion paper) and (vii) establishing an international collaborative research and information exchange effort to leverage capabilities and knowledge.

  10. Control Means for Reactor

    DOEpatents

    Manley, J. H.

    1961-06-27

    An apparatus for controlling a nuclear reactor includes a tank just below the reactor, tubes extending from the tank into the reactor, and a thermally expansible liquid neutron absorbent material in the tank. The liquid in the tank is exposed to a beam of neutrons from the reactor which heats the liquid causing it to expand into the reactor when the neutron flux in the reactor rises above a predetermincd danger point. Boron triamine may be used for this purpose.

  11. Microchannel Reactors for ISRU Applications

    NASA Astrophysics Data System (ADS)

    Carranza, Susana; Makel, Darby B.; Blizman, Brandon; Ward, Benjamin J.

    2005-02-01

    Affordable planning and execution of prolonged manned space missions depend upon the utilization of local resources and the waste products which are formed in manned spacecraft and surface bases. Successful in-situ resources utilization (ISRU) will require component technologies which provide optimal size, weight, volume, and power efficiency. Microchannel reactors enable the efficient chemical processing of in situ resources. The reactors can be designed for the processes that generate the most benefit for each mission. For instance, propellants (methane) can be produced from carbon dioxide from the Mars atmosphere using the Sabatier reaction and ethylene can be produced from the partial oxidation of methane. A system that synthesizes ethylene could be the precursor for systems to synthesize ethanol and polyethylene. Ethanol can be used as a nutrient for Astrobiology experiments, as well as the production of nutrients for human crew (e.g. sugars). Polyethylene can be used in the construction of habitats, tools, and replacement parts. This paper will present recent developments in miniature chemical reactors using advanced Micro Electro Mechanical Systems (MEMS) and microchannel technology to support ISRU of Mars and lunar missions. Among other applications, the technology has been demonstrated for the Sabatier process and for the partial oxidation of methane. Microchannel reactors were developed based on ceramic substrates as well as metal substrates. In both types of reactors, multiple layers coated with catalytic material are bonded, forming a monolithic structure. Such reactors are readily scalable with the incorporation of extra layers. In addition, this reactor structure minimizes pressure drop and catalyst settling, which are common problems in conventional packed bed reactors.

  12. LIQUID PHASE FISCHER-TROPSCH (III & IV) DEMONSTRATION IN THE LAPORTE ALTERNATIVE FUELS DEVELOPMENT UNIT. Final Topical Report. Volume I/II: Main Report. Task 1: Engineering Modifications (Fischer-Tropsch III & IV Demonstration) and Task 2: AFDU Shakedown, Operations, Deactivation (Shut-Down) and Disposal (Fischer-Tropsch III & IV Demonstration).

    SciTech Connect

    Bharat L. Bhatt

    1999-06-01

    Slurry phase Fischer-Tropsch technology was successfully demonstrated in DOE's Alternative Fuels Development Unit (AFDU) at LaPorte, Texas. Earlier work at LaPorte, with iron catalysts in 1992 and 1994, had established proof-of-concept status for the slurry phase process. The third campaign (Fischer-Tropsch III), in 1996, aimed at aggressively extending the operability of the slurry reactor using a proprietary cobalt catalyst. Due to an irreversible plugging of catalyst-wax separation filters as a result of unexpected catalyst fines generation, the operations had to be terminated after seven days on-stream. Following an extensive post-run investigation by the participants, the campaign was successfully completed in March-April 1998, with an improved proprietary cobalt catalyst. These runs were sponsored by the U. S. Department of Energy (DOE), Air Products & Chemicals, Inc., and Shell Synthetic Fuels, Inc. (SSFI). A productivity of approximately 140 grams (gm) of hydrocarbons (HC)/ hour (hr)-liter (lit) of expanded slurry volume was achieved at reasonable system stability during the second trial (Fischer-Tropsch IV). The productivity ranged from 110-140 at various conditions during the 18 days of operations. The catalyst/wax filters performed well throughout the demonstration, producing a clean wax product. For the most part, only one of the four filter housings was needed for catalyst/wax filtration. The filter flux appeared to exceed the design flux. A combination of use of a stronger catalyst and some innovative filtration techniques were responsible for this success. There was no sign of catalyst particle attrition and very little erosion of the slurry pump was observed, in contrast to the Fischer-Tropsch III operations. The reactor operated hydrodynamically stable with uniform temperature profile and gas hold-ups. Nuclear density and differential pressure measurements indicated somewhat higher than expected gas hold-up (45 - 50 vol%) during Fischer-Tropsch IV

  13. Ratchet model for type IV pilus retraction

    NASA Astrophysics Data System (ADS)

    Lindén, Martin; Tuohimaa, Tomi; Jonsson, Ann-Beth; Wallin, Mats

    2004-03-01

    Type IV pilus rectraction is required for twitching motility in a wide range of bacteriae, including Neisseria gonorrhoeae, Myxococcus xanthus and Pseudomonas aeruginosa. The mechanism of retraction is believed to be filament disassembly mediated by PilT, a member of the AAA family of motor proteins. Recent laser tweezer measurements of the force-velocity relation of PilT in N. gonorrhoeae, reveal that single PilT complexes generate forces of over 100 pN. We assume that PilT forms a cyclic ATPase surrounding the base of the pilus and formulate a model of retraction in terms of coupled flashing ratchets. We obtain a force-velocity relation by numerical simulation of the model which is in qualitative agreement with the experimental results.

  14. NEUTRONIC REACTOR

    DOEpatents

    Stewart, H.B.

    1958-12-23

    A nuclear reactor of the type speclfically designed for the irradiation of materials is discussed. In this design a central cyllndrical core of moderating material ls surrounded by an active portlon comprlsed of an annular tank contalning fissionable material immersed ln a liquid moderator. The active portion ls ln turn surrounded by a reflector, and a well ls provided in the center of the core to accommodate the materlals to be irradiated. The over-all dimensions of the core ln at least one plane are equal to or greater than twice the effective slowing down length and equal to or less than twlce the effective diffuslon length for neutrons in the core materials.

  15. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1957-09-17

    A reactor of the type having coolant liquid circulated through clad fuel elements geometrically arranged in a solid moderator, such as graphite, is described. The core is enclosed in a pressure vessel and suitable shielding, wherein means is provided for circulating vapor through the core to superheat the same. This is accomplished by drawing off the liquid which has been heated in the core due to the fission of the fuel, passing it to a nozzle within a chamber where it flashes into a vapor, and then passing the vapor through separate tubes extending through the moderator to pick up more heat developed in the core due to the fission of the fuel, thereby producing superheated vapor.

  16. Very High-Temperature Reactor (VHTR) Proliferation Resistance and Physical Protection (PR&PP)

    SciTech Connect

    Moses, David Lewis

    2011-10-01

    This report documents the detailed background information that has been compiled to support the preparation of a much shorter white paper on the design features and fuel cycles of Very High-Temperature Reactors (VHTRs), including the proposed Next-Generation Nuclear Plant (NGNP), to identify the important proliferation resistance and physical protection (PR&PP) aspects of the proposed concepts. The shorter white paper derived from the information in this report was prepared for the Department of Energy Office of Nuclear Science and Technology for the Generation IV International Forum (GIF) VHTR Systems Steering Committee (SSC) as input to the GIF Proliferation Resistance and Physical Protection Working Group (PR&PPWG) (http://www.gen-4.org/Technology/horizontal/proliferation.htm). The short white paper was edited by the GIF VHTR SCC to address their concerns and thus may differ from the information presented in this supporting report. The GIF PR&PPWG will use the derived white paper based on this report along with other white papers on the six alternative Generation IV design concepts (http://www.gen-4.org/Technology/systems/index.htm) to employ an evaluation methodology that can be applied and will evolve from the earliest stages of design. This methodology will guide system designers, program policy makers, and external stakeholders in evaluating the response of each system, to determine each system's resistance to proliferation threats and robustness against sabotage and terrorism threats, and thereby guide future international cooperation on ensuring safeguards in the deployment of the Generation IV systems. The format and content of this report is that specified in a template prepared by the GIF PR&PPWG. Other than the level of detail, the key exception to the specified template format is the addition of Appendix C to document the history and status of coated-particle fuel reprocessing technologies, which fuel reprocessing technologies have yet to be deployed

  17. Biosynthesis of alpha2(IV) and alpha1(IV) chains of collagen IV and interactions with matrix metalloproteinase-9.

    PubMed

    Toth, M; Sado, Y; Ninomiya, Y; Fridman, R

    1999-07-01

    In vitro binding studies with latent matrix metalloproteinase-9 (pro-MMP-9) have revealed the existence of nondisulfide-bonded alpha2(IV) chains on the cell surface capable of forming a high-affinity complex with the enzyme. Here we investigated the biosynthesis and cellular distribution of alpha2(IV) and alpha1(IV) chains in breast epithelial (MCF10A and MDA-MB-231) and fibrosarcoma (HT1080) cells by pulse-chase analysis followed by immunoprecipitation with chain-specific monoclonal antibodies (mAb). These studies showed that whereas the alpha1(IV) chain remained in the intracellular compartment, nondisulfide-bonded alpha2(IV) chains were secreted into the media in a stable form. Consistently, only alpha2(IV) was detected on the cell surface by surface biotinylation or indirect immunofluorescence. In agreement with the pulse-chase analysis, media subjected to co-precipitation experiments with pro-MMP-9 or pro-MMP-9-affinity chromatography followed by immunoblotting with chain-specific mAbs resulted in the detection of alpha2(IV). A preferential secretion of nondisulfide-bonded alpha2(IV) chains was also observed in CHO-K1 cells transiently transfected with full-length mouse alpha2(IV) or alpha (IV) cDNAs. However, a complex of mouse alpha1(IV) with pro-MMP-9 was coprecipitated with exogenous enzyme from lysates of CHO-K1 cells transfected with mouse alpha1(IV), suggesting that under overexpression conditions the enzyme can also interact with the alpha1 (IV) chain. Collectively, these studies further demonstrate the interactions of pro-MMP-9 with collagen IV chains and a unique processing and targeting of nondisulfide-bonded alpha2(IV) chains that may play a role in the surface/matrix association of pro-MMP-9.

  18. Solid State Reactor Final Report

    SciTech Connect

    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 of research

  19. Nuclear reactor

    DOEpatents

    Yant, Howard W.; Stinebiser, Karl W.; Anzur, Gregory C.

    1977-01-01

    A nuclear reactor, particularly a liquid-metal breeder reactor, whose upper internals include outlet modules for channeling the liquid-metal coolant from selected areas of the outlet of the core vertically to the outlet plenum. The modules are composed of a highly-refractory, high corrosion-resistant alloy, for example, INCONEL-718. Each module is disposed to confine and channel generally vertically the coolant emitted from a subplurality of core-component assemblies. Each module has a grid with openings, each opening disposed to receive the coolant from an assembly of the subplurality. The grid in addition serves as a holdown for the assemblies of the corresponding subplurality preventing their excessive ejection upwardly from the core. In the region directly over the core the outlet modules are of such peripheral form that they nest forming a continuum over the core-component assemblies whose outlet coolant they confine. Each subassembly includes a chimney which confines the coolant emitted by its corresponding subassemblies to generally vertical flow between the outlet of the core and the outlet plenum. Each subplurality of assemblies whose emitted coolant is confined by an outlet module includes assemblies which emit lower-temperature coolant, for example, a control-rod assembly, or fertile assemblies, and assemblies which emit coolant of substantially higher temperature, for example, fuel-rod assemblies. The coolants of different temperatures are mixed in the chimneys reducing the effect of stripping (hot-cold temperature fluctuations) on the remainder of the upper internals which are composed typically of AISI-304 or AISI-316 stainless steel.

  20. Safeguards and Non-proliferation Issues as Related to Advanced Fuel Cycle and Advanced Fast Reactor Development with Processing of Reactor Fuel

    SciTech Connect

    Rahmat Aryaeinejad; Jerry D. Cole; Mark W. Drigert; Dee E. Vaden

    2006-10-01

    The goal of this work is to establish basic data and techniques to enable safeguards appropriate to a new generation of nuclear power systems that will be based on fast spectrum reactors and mixed actinide fuels containing significant quantities of "minor" actinides, possibly due to reprocessing, and determination of what new radiation signatures and parameters need to be considered. The research effort focuses on several problems associated with the use of fuel having significantly different actinide inventories that current practice and on the development of innovative techniques using new radiation signatures and other parameters useful for safeguards and monitoring. In addition, the development of new distinctive radiation signatures as an aid in controlling proliferation of nuclear materials has parallel applications to support Gen-IV and current advanced fuel cycle initiative (AFCI) goals as well as the anticipated Global Nuclear Energy Partnership (GNEP).