Science.gov

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. Benchmark Development in Support of Generation-IV Reactor Validation (IRPhEP 2010 Handbook)

    SciTech Connect

    John D. Bess; J. Blair Briggs

    2010-06-01

    The March 2010 edition of the International Reactor Physics Experiment Evaluation Project (IRPhEP) Handbook includes additional benchmark data that can be implemented in the validation of data and methods for Generation IV (GEN-IV) reactor designs. Evaluations supporting sodium-cooled fast reactor (SFR) efforts include the initial isothermal tests of the Fast Flux Test Facility (FFTF) at the Hanford Site, the Zero Power Physics Reactor (ZPPR) 10B and 10C experiments at the Idaho National Laboratory (INL), and the burn-up reactivity coefficient of Japan’s JOYO reactor. An assessment of Russia’s BFS-61 assemblies at the Institute of Physics and Power Engineering (IPPE) provides additional information for lead-cooled fast reactor (LFR) systems. Benchmarks in support of the very high temperature reactor (VHTR) project include evaluations of the HTR-PROTEUS experiments performed at the Paul Scherrer Institut (PSI) in Switzerland and the start-up core physics tests of Japan’s High Temperature Engineering Test Reactor. The critical configuration of the Power Burst Facility (PBF) at the INL which used ternary ceramic fuel, U(18)O2-CaO-ZrO2, is of interest for fuel cycle research and development (FCR&D) and has some similarities to “inert-matrix” fuels that are of interest in GEN-IV advanced reactor design. Two additional evaluations were revised to include additional evaluated experimental data, in support of light water reactor (LWR) and heavy water reactor (HWR) research; these include reactor physics experiments at Brazil’s IPEN/MB-01 Research Reactor Facility and the French High Flux Reactor (RHF), respectively. The IRPhEP Handbook now includes data from 45 experimental series (representing 24 reactor facilities) and represents contributions from 15 countries. These experimental measurements represent large investments of infrastructure, experience, and cost that have been evaluated and preserved as benchmarks for the validation of methods and collection of

  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. The SGR Multipurpose - Generation IV - Transportable Cogeneration Nuclear Reactor with Innovative Shielding

    SciTech Connect

    Pahladsingh, R.R.

    2002-07-01

    Deregulation and liberalization are changing the global energy-markets. At the same time innovative technologies are introduced in the electricity industry; often as a requirement from the upcoming Digital Society. Energy solutions for the future are more seen as a mix of energy-sources for generation-, transmission- and distribution energy-services. The Internet Energy-web based 'Virtual' enterprises are coming up and will gradually change our society. It the fast changing world we have to realize that there will be less time to look for the adequate solutions to anticipate on global developments and the way they will influence our own societies. Global population may reach 9 billion people by 2030; this will put tremendous pressure on energy-, water- and food supply in the global economy. It is time to think about some major issues as described below and come up with the right answers. These are needed on very short term to secure a humane global economic growth and the sustainable global environment. The DOE (Department of Energy - USA) has started the Generation IV initiative for the new generation of nuclear reactors that must lead to much better safety, economics and public acceptance the new reactors. The SGR (Simplified Gas-cooled Reactor) is being proposed as a Generation IV modular nuclear reactor, using graphite pebbles as fuel, whereby an attempt has been made to meet all the DOE requirements, to be used for future nuclear reactors. The focus in this paper is on the changing and emerging global energy-markets and shows some relevant criteria to the nuclear industry and how we can anticipate with improved and new designs towards the coming Digital Society. (author)

  8. 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.

  9. 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

  10. 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.

  11. A Supercritical CO{sub 2} Cycle- a Promising Power Conversion System for Generation IV Reactors

    SciTech Connect

    Hejzlar, Pavel; Dostal, Vaclav; Driscoll, Michael J.

    2006-07-01

    Advances in power conversion systems (PCS) for Generation IV power plants are of high importance because of their impact on plant specific capital cost reduction, which can be more significant than the cost savings achieved through the modifications of the nuclear island itself. One such PCS candidate, especially attractive for reactor outlet temperatures in the range of 550 to 650 deg C, is applicable to lead-alloy, sodium, or liquid salt-cooled reactors, as well as direct-cycle CO{sub 2} cooled reactors. The efficiencies achievable in this medium temperature range exceed those of conventional Brayton cycles and supercritical steam Rankine cycles and are comparable to those of conventional helium Brayton cycles at turbine inlet temperatures of 800 to 900 deg C. The S-CO{sub 2} recompression cycle under evaluation at MIT, is described with its advantages, drawbacks and R and D needs. The cycle is shown to excel in efficiency, simplicity and compactness which projects to cost savings, and in lower sensitivity of efficiency to core bypass flow, component pressure losses and flow maldistribution in recuperators. On the other hand, the cycle is highly recuperative and thus requires very compact heat exchangers, poses challenges to design of piping for large units, and its control and part load operation is more complicated. (authors)

  12. 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.

  13. 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)

  14. 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.

  15. Generation IV nuclear energy system initiative. Pin core subassembly designfor the Gas-Cooled Fast Reactor.

    SciTech Connect

    Farmer, M. T.; Hoffman, E. A.; Pfeiffer, P. F.; Therios, I. U.

    2006-07-31

    The Gas-Cooled Fast Reactor (GFR) is one of six systems selected for viability assessment in the Generation IV program. It features a closed nuclear fuel cycle, consisting of a high-temperature helium-cooled fast spectrum reactor, coupled to a direct-cycle helium turbine for electricity production. The GFR combines the advances of fast spectrum systems with those of high-temperature systems. It was clear from the very beginning that GFR design should be driven by the objective to offer a complementary approach to liquid metal cooling. On this basis, CEA and the US DOE decided to collaborate on the pre-conceptual design of a GFR. This reactor design will provide a high level of safety and full recycling of the actinides, and will also be highly proliferation resistant and economically attractive. The status of this collaborative project is that two unit sizes, 600 MWt and 2400 MWt were selected as the focus of the design and safety studies. Researchers studied fuel forms, fuel assembly/element designs, core configurations, primary and balance-of-plant layouts, and safety approaches for both of these unit sizes. Results regarding the feasibility of this GFR design are encouraging. For example, sustainability and non-proliferation goals can be met and the proposed concept has attractive safety features. These features take advantage of the helium in terms of its neutronic quasi-transparency as well as the enhanced Doppler effect in connection with candidate fuel and structural materials. The current design trend is to consider high unit power for the GFR (2400 MWt), an attractive level for the power density (100 MW/m{sup 3}), and the implementation of an innovative plate type fuel or pin type sub-assembly with carbide-based actinide compounds and SiC-based structural materials. Work is still needed to refine the safety approach, to select the main system options, and to more definitively establish economic parameters.

  16. 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.

  17. 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.

  18. 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)

  19. 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.

  20. 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

  1. 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.

  2. 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.

  3. 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.

  4. 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

  5. Generation IV nuclear energy system initiative. Large GFR core subassemblydesign for the Gas-Cooled Fast Reactor.

    SciTech Connect

    Hoffman, E. A.; Kulak, R. F.; Therios, I. U.; Wei, T. Y. C.

    2006-07-31

    Gas-cooled fast reactor (GFR) designs are being developed to meet Gen IV goals of sustainability, economics, safety and reliability, and proliferation resistance and physical protection as part of an International Generation IV Nuclear Energy System Research Initiative effort. Different organizations are involved in the development of a variety of GFR design concepts. The current analysis has focused on the evaluation of low-pressure drop, pin-core designs with favorable passive cooling properties. Initial evaluation of the passive cooling safety case for the GFR during depressurized decay heat removal accidents with concurrent loss of electric power have resulted in requirements for a reduction of core power density to the 100 w/cc level and a low core pressure drop of 0.5 bars. Additional design constraints and the implementation of their constraints are evaluated in this study to enhance and passive cooling properties of the reactor. Passive cooling is made easier by a flat radial distribution of the decay heat. One goal of this study was to evaluate the radial power distribution and determine to what extent it can be flattened, since the decay heat is nearly proportional to the fission power at shutdown. In line with this investigation of the radial power profile, an assessment was also made of the control rod configuration. The layout provided a large number of control rod locations with a fixed area provided for control rods. The number of control rods was consistent with other fast reactor designs. The adequacy of the available control rod locations was evaluated. Future studies will be needed to optimize the control rod designs and evaluate the shutdown system. The case for low pressure drop core can be improved by the minimization of pressure drop sources such as the number of required fuel spacers in the subassembly design and by the details of the fuel pin design. The fuel pin design is determined by a number of neutronic, thermal-hydraulic (gas dynamics

  6. 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.

  7. 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.

  8. 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)

  9. NUCLEAR ENERGY RESEARCH INITIATIVE (NERI) PROGRAM GRANT NUMBER DE-FG03-00SF22168 TECHNICAL PROGRESS REPORT (Nov. 15, 2001 - Feb. 15,2002) ''Design and Layout Concepts for Compact, Factory-Produced, Transportable, Generation IV Reactor Systems''

    SciTech Connect

    Fred R. Mynatt; Andy Kadak; Marc Berte; Larry Miller; Mohammed Khan; Joe McConn; Lawrence Townsend; Wesley Williams; Martin Williamson

    2002-03-15

    The objectives of this project are to develop and evaluate nuclear power plant designs and layout concepts to maximize the benefits of compact modular Generation IV reactor concepts including factory fabrication and packaging for optimal transportation and siting. Three nuclear power plant concepts are being studied representing water, helium and lead-bismuth coolants. This is the sixth quarterly progress report.

  10. 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.

  11. NUCLEAR ENERGY RESEARCH INITIATIVE (NERI) PROGRAM GRANT NUMBER DE-FG03-00SF22168 TECHNICAL PROGRESS REPORT (Aug 15, 2002 to Nov. 15, 2002) - DESIGN AND LAYOUT CONCEPTS FOR COMPACT, FACTORY-PRODUCED, TRANSPORTABLE GENERATION IV REACTOR SYSTEMS

    SciTech Connect

    Fred R. Mynatt; Andy Kadak; Marc Berte; Larry Miller; Lawrence Townsend; Martin Williamson; Rupy Sawhney; Jacob Fife

    2002-12-15

    The objectives of this project are to develop and evaluate nuclear power plant designs and layout concepts to maximize the benefits of compact modular Generation IV reactor concepts including factory fabrication and packaging for optimal transportation and siting. This report covers the ninth quarter of the project. The three reactor concept teams have completed initial plant concept development, evaluation and layout. A significant design effort has proceeded with substantial change and evolution from original ideas. The concepts have been reviewed by the industry participants and improvements have been implemented. The third phase, industrial engineering simulation of reactor fabrication has begun.

  12. 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.

  13. 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

  14. 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.

  15. 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)

  16. Regenerative Heater Optimization for Steam Turbo-Generation Cycles of Generation IV Nuclear Power Plants with a Comparison of Two Concepts for the Westinghouse International Reactor Innovative and Secure (IRIS)

    SciTech Connect

    Williams, W.C.

    2002-08-01

    The intent of this study is to discuss some of the many factors involved in the development of the design and layout of a steam turbo-generation unit as part of a modular Generation IV nuclear power plant. Of the many factors involved in the design and layout, this research will cover feed water system layout and optimization issues. The research is arranged in hopes that it can be generalized to any Generation IV system which uses a steam powered turbo-generation unit. The research is done using the ORCENT-II heat balance codes and the Salisbury methodology to be reviewed herein. The Salisbury methodology is used on an original cycle design by Famiani for the Westinghouse IRIS and the effects due to parameter variation are studied. The vital parameters of the Salisbury methodology are the incremental heater surface capital cost (S) in $/ft{sup 2}, the value of incremental power (I) in $/kW, and the overall heat transfer coefficient (U) in Btu/ft{sup 2}-degrees Fahrenheit-hr. Each is varied in order to determine the effects on the cycles overall heat rate, output, as well as, the heater surface areas. The effects of each are shown. Then the methodology is then used to compare the optimized original Famiani design consisting of seven regenerative feedwater heaters with an optimized new cycle concept, INRC8, containing four regenerative heaters. The results are shown. It can be seen that a trade between the complexity of the seven stage regenerative Famiani cycle and the simplicity of the INRC8 cycle can be made. It is desired that this methodology can be used to show the ability to evaluate modularity through the value of size a complexity of the system as well as the performance. It also shows the effectiveness of the Salisbury methodology in the optimization of regenerative cycles for such an evaluation.

  17. 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)

  18. 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.

  19. 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.

  20. 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.

  1. Decay heat removal in GEN IV gas cooled fast reactors.

    SciTech Connect

    Cheng, L. Y.; Wei, T. Y. C.

    2009-08-01

    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 were 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.

  2. The U.S. Generation IV Implementation Strategy

    SciTech Connect

    2003-09-01

    This report has been prepared by the U.S. Department of Energy (DOE) to respond to Congressional direction contained in Senate Report 107-220 from the Senate Committee on Appropriations regarding the Energy and Water Development Appropriations for 2003. In that report, the Committee instructed the Department to prepare a report regarding how it intends to carry out the results of the Generation IV Roadmap. This report is the U.S. Department of Energy's response to the Congressional directive. It summarizes results from the Generation IV Technology Roadmap and the strategy for implementing of the Generation IV program in the United States. Planning for the implementation of the Generation IV program is based on (1) the long-term outlook for nuclear energy in the United States, (2) the advice of the Nuclear Energy Research Advisory Committee during the two-year development of the Generation IV Technology Roadmap, and (3) the need for the Generation IV program to be integrated with other nuclear energy research programs of the Department. Considerable emphasis is given to developing the priorities and necessary timelines for the U.S. Generation IV Program, as well as developing international R&D cooperation that will benefit the program and strengthen U.S. leadership in nuclear technology R&D.

  3. Maize RNA polymerase IV defines trans-generational epigenetic variation.

    PubMed

    Erhard, Karl F; Parkinson, Susan E; Gross, Stephen M; Barbour, Joy-El R; Lim, Jana P; Hollick, Jay B

    2013-03-01

    The maize (Zea mays) RNA Polymerase IV (Pol IV) largest subunit, RNA Polymerase D1 (RPD1 or NRPD1), is required for facilitating paramutations, restricting expression patterns of genes required for normal development, and generating small interfering RNA (siRNAs). Despite this expanded role for maize Pol IV relative to Arabidopsis thaliana, neither the general characteristics of Pol IV-regulated haplotypes, nor their prevalence, are known. Here, we show that specific haplotypes of the purple plant1 locus, encoding an anthocyanin pigment regulator, acquire and retain an expanded expression domain following transmission from siRNA biogenesis mutants. This conditioned expression pattern is progressively enhanced over generations in Pol IV mutants and then remains heritable after restoration of Pol IV function. This unusual genetic behavior is associated with promoter-proximal transposon fragments but is independent of sequences required for paramutation. These results indicate that trans-generational Pol IV action defines the expression patterns of haplotypes using co-opted transposon-derived sequences as regulatory elements. Our results provide a molecular framework for the concept that induced changes to the heterochromatic component of the genome are coincident with heritable changes in gene regulation. Alterations of this Pol IV-based regulatory system can generate potentially desirable and adaptive traits for selection to act upon. PMID:23512852

  4. 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.

  5. 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

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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)

  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. 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.

  13. 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

  14. 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.

  15. Multiple Unit Instrumentation and Control (I and C) Systems for Generation IV Nuclear Power Systems

    SciTech Connect

    Miller, Don W.; Fiorino, Michael M.; Quinn, Edward 'Ted'; Mauck, Jerry L.

    2002-07-01

    Several Generation IV design concepts involve compact modular reactor configurations that can significantly reduce the overall cost of construction of a nuclear plant. However, the operating costs of independent smaller units are increased on a per-MW basis versus larger scale reactors. To offset this economic penalty, Generation IV nuclear plants will benefit economically from a multi-unit (or multi-module) configuration, where some facilities or power conversion system resources are shared; balance of plant systems, auxiliary systems, and the main control room are all candidates for shared or integrated implementation. However, these multi-modular configurations introduce safety and operational challenges that must be addressed at an early stage in the design process. The goal of this paper is the identification and evaluation of the regulatory, operational, and monitoring issues arising from multi-unit nuclear plant implementations. The paper will provide an overview of a research approach that uses a model of a generic module as a basis for integration of design and monitoring, alongside regulatory requirements, for these proposed configurations. (authors)

  16. 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.

  17. 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

    Energy Science and Technology Software Center (ESTSC)

    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

  18. 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)

  19. RGG: Reactor geometry (and mesh) generator

    SciTech Connect

    Jain, R.; Tautges, T.

    2012-07-01

    The reactor geometry (and mesh) generator RGG takes advantage of information about repeated structures in both assembly and core lattices to simplify the creation of geometry and mesh. It is released as open source software as a part of the MeshKit mesh generation library. The methodology operates in three stages. First, assembly geometry models of various types are generated by a tool called AssyGen. Next, the assembly model or models are meshed by using MeshKit tools or the CUBIT mesh generation tool-kit, optionally based on a journal file output by AssyGen. After one or more assembly model meshes have been constructed, a tool called CoreGen uses a copy/move/merge process to arrange the model meshes into a core model. In this paper, we present the current state of tools and new features in RGG. We also discuss the parallel-enabled CoreGen, which in several cases achieves super-linear speedups since the problems fit in available RAM at higher processor counts. Several RGG applications - 1/6 VHTR model, 1/4 PWR reactor core, and a full-core model for Monju - are reported. (authors)

  20. 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)

  1. 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.

  2. 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.

  3. Generating unstructured nuclear reactor core meshes in parallel

    DOE PAGESBeta

    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

  4. 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.

  5. 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

  6. The next generation of power reactors - safety characteristics

    SciTech Connect

    Modro, S.M.

    1995-01-01

    The next generation of commercial nuclear power reactors is characterized by a new approach to achieving reliability of their safety systems. In contrast to current generation reactors, these designs apply passive safety features that rely on gravity-driven transfer processes or stored energy, such as gas-pressurized accumulators or electric batteries. This paper discusses the passive safety system of the AP600 and Simplified Boiling Water Reactor (SBWR) designs.

  7. 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)

  8. 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.

  9. 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.

  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. Maize RNA Polymerase IV Defines trans-Generational Epigenetic Variation[W

    PubMed Central

    Erhard, Karl F.; Parkinson, Susan E.; Gross, Stephen M.; Barbour, Joy-El R.; Lim, Jana P.; Hollick, Jay B.

    2013-01-01

    The maize (Zea mays) RNA Polymerase IV (Pol IV) largest subunit, RNA Polymerase D1 (RPD1 or NRPD1), is required for facilitating paramutations, restricting expression patterns of genes required for normal development, and generating small interfering RNA (siRNAs). Despite this expanded role for maize Pol IV relative to Arabidopsis thaliana, neither the general characteristics of Pol IV–regulated haplotypes, nor their prevalence, are known. Here, we show that specific haplotypes of the purple plant1 locus, encoding an anthocyanin pigment regulator, acquire and retain an expanded expression domain following transmission from siRNA biogenesis mutants. This conditioned expression pattern is progressively enhanced over generations in Pol IV mutants and then remains heritable after restoration of Pol IV function. This unusual genetic behavior is associated with promoter-proximal transposon fragments but is independent of sequences required for paramutation. These results indicate that trans-generational Pol IV action defines the expression patterns of haplotypes using co-opted transposon-derived sequences as regulatory elements. Our results provide a molecular framework for the concept that induced changes to the heterochromatic component of the genome are coincident with heritable changes in gene regulation. Alterations of this Pol IV–based regulatory system can generate potentially desirable and adaptive traits for selection to act upon. PMID:23512852

  12. A REVIEW ON CURRENT STATUS OF ALLOYS 617 AND 230 FOR GEN IV NUCLEAR REACTOR INTERNALS AND HEAT EXCHANGERS

    SciTech Connect

    Ren, Weiju; Swindeman, Robert W

    2009-01-01

    Alloys 617 and 230 are currently identified as two leading candidate metallic materials in the down selection for applications at temperatures above 760 C in the Gen IV Nuclear Reactor Systems. Qualifying the materials requires significant information related to Codification, mechanical behavior modeling, metallurgical stability, environmental resistance, and many other aspects. In the present paper, material requirements for the Gen IV Nuclear Reactor Systems are discussed; certain available information regarding the two alloys under consideration for the intended applications are reviewed and analyzed. Suggestions are presented for further R&D activities for the materials selection.

  13. 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.

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

    DOE PAGESBeta

    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

  15. 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

  16. 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.

  17. Safety aspect concerning radiolytic gas generation in reactors.

    PubMed

    Ramshesh, V

    2001-01-01

    In water cooled and water moderated reactors (H2O in boiling water reactors/pressurised water reactors, D2O in pressurised heavy water reactors) during normal operation, radiolysis is a source of production of hydrogen/deuterium and oxygen. During the progress of a nuclear accident, while there are other important sources of hydrogen/deuterium, the oxygen availability can occur only through radiolysis or direct contact with air. In air saturated with water vapour at room temperature and pressure when H2/D2 concentration exceeds 4 vol % (a conservative estimate), a combustible mixture with oxygen can be formed. It is proposed to examine the basic principles of water radiolysis as far as they pertain to generation of H2/D2 and O2 and try to apply these concepts to reactors both under operating conditions and in accident situations. It is concluded that the possibility of an accident taking place through radiolysis is highly unlikely. PMID:11382138

  18. 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.

  19. 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

  20. 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.

  1. 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.

  2. 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.

  3. 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

  4. 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).

  5. 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

  6. 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.

  7. Generation III reactors safety requirements and the design solutions

    SciTech Connect

    Felten, P.

    2009-03-31

    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.

  8. 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.

  9. 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

  10. 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

  11. 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.

  12. Granular flow in pebble bed reactors: Dust generation and scaling

    SciTech Connect

    Rycroft, C. H.; Lind, T.; Guentay, S.; Dehbi, A.

    2012-07-01

    In experimental prototypes of pebble bed reactors, significant quantities of graphite dust have been observed due to rubbing between pebbles as they flow through the core. At the high temperatures and pressures in these reactors, little data is available to understand the frictional properties of the pebble surfaces, and as a result, the Paul Scherrer Institut (Switzerland) proposes a conceptual design of a scaled-down version of a pebble bed reactor to investigate this issue in detail. In this paper, simulations of granular flow in pebble bed reactors using the discrete-element method are presented. Simulations in the full geometry (using 440,000 pebbles) are compared to those in geometries scaled down by 3:1 and 6:1. The simulations show complex behavior due to discrete pebble packing effects, meaning that pebble flow and dust generation in a scaled-down facility may be significantly different. The differences between velocity profiles, packing geometry, and pebble wear at the different scales are discussed. The results can aid in the design of the prototypical facility to more accurately reproduce the flow in a full-size reactor. (authors)

  13. 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

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. Materials development for a fast breeder reactor steam generator concept

    SciTech Connect

    Sessions, C.E.; Reynolds, S.D. Jr.; Hebbar, M.A.; Lewis, J.F.; Kiefer, J.H.

    1981-11-01

    The progress achieved since 1977 in the important area of materials and processes development of fast reactor steam generator development is summarized. The two distinguishing features of the proposed Westinghouse-Tampa steam generator concept are the convoluted shell expansion joint (CSEJ) and the double-wall tubing with a third fluid leak detection capability. A 2/one quarter/ Cr-1 Mo low alloy steel will be used for all important parts of the generator including the CSEJ and the tubes. Other areas in which progress was made include tube-to-tubesheet (T/TS) welding, post-weld heat treatment (PWHT), tube expansion, and development of materials specifications for prototype and future plant materials. 8 refs.

  19. Versatile thin-film reactor for photochemical vapor generation.

    PubMed

    Zheng, Chengbin; Sturgeon, Ralph E; Brophy, Christine; Hou, Xiandeng

    2010-04-01

    A novel thin-film reactor is described and evaluated for its analytical performance with photochemical vapor generation (TF-PVG). The device, comprising both the generator and a gas-liquid separator, utilizes a vertical central quartz rod onto which the sample is pumped to yield a thin liquid film conducive to the rapid escape of generated hydrophobic species. The rod is housed within a concentric quartz tube through which a flow of argon carrier/stripping gas is passed to remove and transport the generated species to a detector, which in this study is an inductively coupled argon plasma optical emission spectrometer (ICP-OES). The concentric quartz tube is itself surrounded by a 78-turn 0.5 m long quartz coil low-pressure mercury discharge lamp operating at 20 W. The performance of this thin-film photoreactor was evaluated through comparison of analytical figures of merit for detection of a number of elements undergoing PVG in the presence of formic or acetic acid with those arising from conventional solution nebulization under optimized conditions. The TF-PVG reactor provided sensitivity enhancements, of 110-, 120-, 130-, 250-, 120-, 230-, 78-, 1.3-, 16-, and 32-fold for As, Sb, Bi, Se, Te, Hg, Ni, Co, Fe, and I, respectively, and detection limit enhancements of 110-, 140-, 170-, 270-, 200-, 300-, 160-, 2.7-, 50-, and 44-fold for these same elements. Vapor generation efficiencies ranged from 20-100% for this suite of analytes. The utility of this technique was demonstrated by the determination of Fe and Ni in Certified Reference Materials DORM-3 (fish protein) and DOLT-4 (dogfish liver tissue). PMID:20225824

  20. 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

  1. 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

  2. "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.

  3. Assessment methodology development for proliferation resistance and physical protection of generation IV systems.

    SciTech Connect

    Roglans, J.; Peterson, P. F.; Nuclear Engineering Division; Univ. of California Berkeley

    2003-01-01

    One of the technology goals set for future Generation IV nuclear energy systems is to be 'a very unattractive and least desirable route' for proliferation, and to provide increased physical protection against theft or sabotage. To evaluate system performance for this goal, an international Expert Group has been formed and has adopted an evaluation method that involves three elements: (1) a process to systematically identify the range of potential security challenges that could face the system-a 'threat space' that includes State diversion or undeclared production of materials for nuclear explosives (proliferation resistance), and non-State theft or radiological sabotage (physical protection robustness); (2) methods for evaluating the system response to these challenges, at a level of detail appropriate to the stage of system or facility design; and (3) a set of measures of system performance that allow assessment and comparison of how well facilities systems meet the goal of providing a 'very unattractive and least desirable route.'

  4. High-Force Generation Is a Conserved Property of Type IV Pilus Systems▿

    PubMed Central

    Clausen, Martin; Jakovljevic, Vladimir; Søgaard-Andersen, Lotte; Maier, Berenike

    2009-01-01

    The type IV pilus (T4P) system of Neisseria gonorrhoeae is the strongest linear molecular motor reported to date, but it is unclear whether high-force generation is conserved between bacterial species. Using laser tweezers, we found that the average stalling force of single-pilus retraction in Myxococcus xanthus of 149 ± 14 pN exceeds the force generated by N. gonorrhoeae. Retraction velocities including a bimodal distribution were similar between M. xanthus and N. gonorrhoeae, but force-dependent directional switching was not. Force generation by pilus retraction is energized by the ATPase PilT. Surprisingly, an M. xanthus mutant lacking PilT apparently still retracted T4P, although at a reduced frequency. The retraction velocity was comparable to the high-velocity mode in the wild type at low forces but decreased drastically when the force increased, with an average stalling force of 70 ± 10 pN. Thus, M. xanthus harbors at least two different retraction motors. Our results demonstrate that the major physical properties are conserved between bacteria that are phylogenetically distant and pursue very different lifestyles. PMID:19429611

  5. 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.

  6. 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

  7. SPECIATION OF SELENIUM(IV) AND SELENIUM(VI) USING COUPLED ION CHROMATOGRAPHY: HYDRIDE GENERATION ATOMIC ABSORPTION SPECTROMETRY

    Technology Transfer Automated Retrieval System (TEKTRAN)

    A simple method was developed to speciate inorganic selenium in the microgram per liter range using coupled ion chromatography-hydride generation atomic absorption spectrometry. Because of the differences in toxicity and adsorption behavior, determination of the redox states selenite, Se(IV), and s...

  8. Less common applications of monoliths: IV. Recent developments in immobilized enzyme reactors for proteomics and biotechnology

    PubMed Central

    Krenkova, Jana; Svec, Frantisek

    2009-01-01

    Use of monolithic supports for enzyme immobilization rapidly expanded since we published the previous part in this series concerned with this topic almost three years ago. Many groups worldwide realized the benefits of applying monolith as a support and used a variety of techniques to immobilize many different enzymes. Although some of these new developments are a refinement of the methods developed previously, some notable new approaches have also been reported. This review summarizes the literature published since 2006 and demonstrates the broad variability of reactive monolith prepared from silica as well as from organic polymers in shapes of disks, columns, and capillaries. All these monoliths were prepared using direct formation from reactive precursors or activation of preformed inactive structures. Interestingly, most of the applications of monolithic enzyme reactors targets proteolytic digestion of proteins for proteomic analysis. PMID:19194973

  9. Second generation Research Reactor Fuel Container (RRFC-II).

    SciTech Connect

    Abhold, M. E.; Baker, M. C.; Bourret, S. C.; Harker, W. C.; Pelowitz, D. G.; Polk, P. J.

    2001-01-01

    The second generation Research Reactor Fuel Counter (RRFC-II) has been developed to measure the remaining {sup 235}U content in foreign spent Material Test Reactor (MTR)-type fuel being returned to the Westinghouse Savannah River Site (WSRS) for interim storage and subsequent disposal. The fuel to be measured started as fresh fuel nominally with 93% enriched Uraniuin alloyed with A1 clad in Al. The fuel was irradiated to levels of up to 65% burnup. The RRFC-II, which will be located in the L-Basin spent fuel pool, is intended to assay the {sup 235}U content using a combination of passive neutron coincidence counting, active neutron coincidence counting, and active-multiplicity analysis. Measurements will be done underwater, eliminating the need for costly and hazardous handling operations of spent fuel out of water. The underwater portion of the RRFC-II consists of a watertight stainless steel housing containing neutron and gamma detectors and a scanning active neutron source. The portion of the system that resides above water consists of data-processing electronics; electromechanical drive electronics; a computer to control the operation of the counter, to collect, and to analyze data; and a touch screen interface located at the equipment rack. The RRFC-II is an improved version of the Los Alamos-designed RRFC already installed in the SRS Receipts Basin for Offsite Fuel. The RRFC-II has been fabricated and is scheduled for installation in late FY 2001 pending acceptance testing by Savannah River Site personnel.

  10. 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.

  11. 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.

  12. 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.

  13. 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

  14. 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.

  15. 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

  16. 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.

  17. Beam Loading Generated by the LOLA-IV Structure in TTF-II

    SciTech Connect

    Bane, Karl LF

    2003-09-17

    The LOLA-IV transverse detecting cavity [1] is used in the Sub-Picosecond Photon Source (SPPS) as a diagnostic for measuring the length of very short bunches (with rms length on the order of tens of microns). It is envisioned to use the same structure for the same purpose in the Tesla Test Facility (TTF)-II. However, unlike the SPPS, the TTF-II may also run in multi-bunch mode, and the question arises, How serious is the beam loading that would be induced? In this report we address this question and find that, for LOLA-IV in TTF-II, the variation in beam-loading induced energy is confined to the first {approx} 80 bunches, and that the total spread in induced energy--the difference in energy between the bunch in the train with the highest energy and the one with the lowest energy--is very small, {approx} 0.03%.

  18. A non-iterative technique for determination of solar cell parameters from the light generated I-V characteristic

    NASA Astrophysics Data System (ADS)

    Kumar, Gaurav; Panchal, Ashish K.

    2013-08-01

    Accurate information about electrical parameters of a photovoltaic (PV) cell is many times essential for evaluating the performance of the cell when delivering power at its full capacity. This paper presents a technique for determining the cell parameters from the light generated current-voltage (I-V) characteristic with a valid assumption for any kind of cells. The technique neither involves any initial approximations nor iteration processes. The technique is employed for various PV cell technologies such as silicon, copper indium gallium selenide, organic, dye sensitized solar cell, and organic tandem cells, previously available in the literatures. Obtained I-V characteristics for the cells using the present technique are in well agreement with those of reported in the literature. The technique is further extended for the analysis of a silicon cell and a silicon module tested in the laboratory and the results obtained are very close to those of the experimental data.

  19. Kinetics of phosphodiester cleavage by differently generated cerium(IV) hydroxo species in neutral solutions.

    PubMed

    Maldonado, Ana L; Yatsimirsky, Anatoly K

    2005-08-01

    Neutral aqueous solutions of cerium ammonium nitrate obtained by dilution of their acetonitrile stock solution with imidazole buffer show high catalytic activity in the hydrolysis of bis(p-nitrophenyl) phosphate (BNPP) and better reproducibility than other similar systems, but suffer from low stability. The kinetics of catalytic hydrolysis is second-order in Ce(IV), independent of pH in the range 5-8 and tentatively involves the Ce2(OH)7+ species as the active form. Attempts to stabilize the active species by different types of added ligands failed, but the use of Ce(IV) complexes pre-synthesized in an organic solvent with potentially stabilizing ligands as precursors of active hydroxo species appeared to be more successful. Three new Ce(IV) complexes, [Ce(Phen)2O(NO3)2], [Ce(tris)O(NO3)(OH)] and [Ce(BTP)2(NO3)4].2H2O (BTP = bis-tris propane, 1,3-bis[tris(hydroxymethyl)methylamino]propane), were prepared by reacting cerium ammonium nitrate with the respective ligands in acetonitrile and were characterized by analytical and spectroscopic techniques. Aqueous solutions of these complexes undergo rapid hydrolysis producing nearly neutral polynuclear Ce(IV) oxo/hydroxo species with high catalytic activity in BNPP hydrolysis. Potentiometric titrations of the solutions obtained from the complex with BTP revealed the formation of Ce4(OH)15+ species at pH > 7, which are protonated affording Ce4(OH)14(2+) and then Ce4(OH)13(3+) on a decrease in pH from 7 to 5. The catalytic activity increases strongly on going to species with a higher positive charge. The reaction mechanism involves first- and second-order in catalyst paths as well as intermediate complex formation with the substrate for higher charged species. PMID:16032364

  20. 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.

  1. Automatic control system by power distribution in a power-generating reactor

    SciTech Connect

    Aleksakov, A.N.; Podlazov, L.N.; Ryabov, V.I.; Shevchenko, V.V.; Postnikov, V.V.

    1980-12-01

    The development of the theoretical principles of construction of these systems and of sufficiently detailed nonlinear dynamic numerical models of a power-generation unit with an RBMK reactor have allowed a consistent procedure to be produced for the engineering synthesis of an (local automated control) LAC-LEP (local emergency protection) system. The LAC system facilitates the shaping and maintenance of the desired power distribution in the whole volume of the reactor. In emergency situations, the LAC-LEP system qualitatively reduces the power to a safe level and effectively suppresses the power warpings in one-half of the reactor, which are characteristic for these reactors.

  2. 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.

  3. 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.

  4. 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.

  5. 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...

  6. 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).

  7. Identification of Zr(iv)-based architectures generated from ligands incorporating the 2,2'-biphenolato unit.

    PubMed

    Miao, Chengrui; Khalil, Georges; Chaumont, Alain; Mobian, Pierre; Henry, Marc

    2016-05-10

    The structural identification in solution of the Zr(iv) complexes involving two 2,2-biphenol-based proligands is reported. The proligand L(1)H2 contains one 2,2-biphenol unit whereas L(2)H4 incorporates two 2,2-biphenol units linked by a para-phenylene bridge. Diffusion Ordered Spectroscopy (DOSY) combined with electrospray mass spectrometry analysis and density functional theory (DFT) allowed for determining the molecular structures of such Zr(iv)-based architectures. It is proposed that [Zr(OPr(i))4(HOPr(i))] in the presence of L(1)H2 generates an octahedral complex formulated as [ZrL(1)3H2]. Concerning the self-assembled architecture incorporating the L(2) ligand, the analytical data highlight the formation of an unprecedented neutral Zr(iv) triple-stranded helicate ([Zr2L(2)3H4]). Insight into the geometry of these complexes is obtained via DFT calculations. Remarkably, the helicate structure characterized in solution strongly contrasts with the triple-stranded structure of the complex that crystallizes. PMID:27070916

  8. 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

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

    PubMed

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

    2016-03-01

    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 nonclassical 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 nonclassical platinum(II) complexes with trans geometry or with a monofunctional coordination mode, polynuclear platinum(II) compounds, platinum(IV) prodrugs, dual-threat agents, and photoactivatable platinum(IV) complexes. Nanoparticles 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 will reflect our optimism that the next generation of approved platinum cancer drugs is about to arrive. PMID:26865551

  10. 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.

  11. 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.

  12. 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.

  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. 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.

  15. Challenges to Integration of Safety and Reliability with Proliferation Resistance and Physical Protection for Generation IV Nuclear Energy Systems

    SciTech Connect

    H. Khalil; P. F. Peterson; R. Bari; G. -L. Fiorini; T. Leahy; R. Versluis

    2012-07-01

    The optimization of a nuclear energy system's performance requires an integrated consideration of multiple design goals - sustainability, safety and reliability (S&R), proliferation resistance and physical protection (PR&PP), and economics - as well as careful evaluation of trade-offs for different system design and operating parameters. Design approaches motivated by each of the goal areas (in isolation from the other goal areas) may be mutually compatible or in conflict. However, no systematic methodology approach has yet been developed to identify and maximize synergies and optimally balance conflicts across the possible design configurations and operating modes of a nuclear energy system. Because most Generation IV systems are at an early stage of development, design, and assessment, designers and analysts are only beginning to identify synergies and conflicts between PR&PP, S&R, and economics goals. The close coupling between PR&PP and S&R goals has motivated early attention within the Generation IV International Forum to their integrated consideration to facilitate the optimization of their effects and the minimization of potential conflicts. This paper discusses the status of this work.

  16. 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

  17. A Compact Torus Fusion Reactor Utilizing a Continuously Generated Strings of CT's. The CT String Reactor, CTSR.

    SciTech Connect

    Hartman, C W; Reisman, D B; McLean, H S; Thomas, J

    2007-05-30

    A fusion reactor is described in which a moving string of mutually repelling compact toruses (alternating helicity, unidirectional Btheta) is generated by repetitive injection using a magnetized coaxial gun driven by continuous gun current with alternating poloidal field. An injected CT relaxes to a minimum magnetic energy equilibrium, moves into a compression cone, and enters a conducting cylinder where the plasma is heated to fusion-producing temperature. The CT then passes into a blanketed region where fusion energy is produced and, on emergence from the fusion region, the CT undergoes controlled expansion in an exit cone where an alternating poloidal field opens the flux surfaces to directly recover the CT magnetic energy as current which is returned to the formation gun. The CT String Reactor (CTSTR) reactor satisfies all the necessary MHD stability requirements and is based on extrapolation of experimentally achieved formation, stability, and plasma confinement. It is supported by extensive 2D, MHD calculations. CTSTR employs minimal external fields supplied by normal conductors, and can produce high fusion power density with uniform wall loading. The geometric simplicity of CTSTR acts to minimize initial and maintenance costs, including periodic replacement of the reactor first wall.

  18. 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.

  19. 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.

  20. Nonuniform steam generator U-tube flow distribution during natural circulation tests in ROSA-IV large scale test facility

    SciTech Connect

    Kukita, Y.; Nakamura, H.; Tasaka, K. ); Chauliac, C. )

    1988-08-01

    Natural circulation experiments were conducted in a large-scale (1/48 scale in volume) full-height simulator of a Westinghouse-type pressurized water reactor. This facility has two steam generators each containing 141 full-size U-tubes of 9 different heights. Transition of the natural circulation mode was observed in the experiments as the primary of side mass inventory was decreased. Three major circulation modes were observed: single-phase liquid natural circulation, two-phase natural circulation, and reflux condensation. For all these circulation modes, and during the transitions between the modes, the mass flow distribution among the steam generator U-tubes was significantly nonuniform. The longer U-tubes indicated reversed flow at higher primary side mass inventories and also tended to empty earlier than the shorter U-tubes when the primary side mass inventory was decreased.

  1. 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.

  2. 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

  3. 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.

  4. Calculation of stellar structure. IV. Results using a detailed energy generation subroutine.

    NASA Astrophysics Data System (ADS)

    Rouse, C. A.

    1995-12-01

    The results from two solar model calculations using the "energy.for" energy generation and neutrino flux code (Bahcall & Pinsonneault 1992) are presented. The models of the present Sun were generated using the program described in the first three papers of this series and using only the helium abundance profile from the Bahcall & Ulrich (1988) (BU) standard model in the present model structure calculations. One model is a simulation of the BU model and yields a ^37^Cl solar neutrino counting rate of 7.0SNU (compared to 7.9SNU for the BU model) and a ^71^Ga neutrino experiment counting rate between 112 and 137SNU (compared to 132SNU for the BU model). The second model has a postulated small high-Z core (Rouse 1983) and yields a ^37^Cl neutrino experiment counting rate of 2.45SNU that is within one sigma of the Homestake Collaboration observed rate of (2.55+/-0.25)SNU (see Parke 1995). It yields a ^71^Ga neutrino experiment counting rate between 89 and 103SNU that is within one sigma of the GALLEX Collaboration neutrino experiment observed rate of (79+/-12)SNU (see Parke 1995). The theoretical ^8^B solar neutrino flux and the observed Kamiokande ^8^B flux (Hirata et al. 1989) are discussed regarding the puzzle of explaining both the chlorine experiment results and the Kamiokande results. The modification of the energy.for code for use in the current Rouse program is described. Consistency of a high-Z core solar model with theories of star formation from pre-stellar nuclei (Krat 1952; Urey 1956; Huang 1957) is suggested.

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

    SciTech Connect

    S. Blaine Grover

    2009-09-01

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

  6. 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

  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. 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

  9. 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.

  10. Hydrogen generation from hydrides in millimeter scale reactors for micro proton exchange membrane fuel cell applications

    NASA Astrophysics Data System (ADS)

    Zhu, L.; Kim, D.; Kim, H.; Masel, R. I.; Shannon, M. A.

    This paper introduces and discusses the feasibility of millimeter scale powder packed-bed reactors using high energy density chemical hydrides for micro-PEM fuel cell applications. Two different reactors were designed and tested using LiBH 4, LiAlH 4, NaAlH 4 and CaH 2 hydride fuels. The mechanisms that limit the total yield of H 2 generated and impact the performance of the hydrogen generator are investigated in this paper, including density, solubility and porosity of the reaction byproducts. The volume expansion of the byproducts has significant effect on the total energy density of micro-hydrogen generators because the byproducts are left in the millimeter scale reactor after the reactions and the micro-hydrogen generators have limited fuel storage space. The SEM images of the reaction byproducts indicates that the byproducts of LiBH 4 can form a single solid mass that clogged the reaction vessel and limit the full utilization of the hydride. However, the byproducts of LiAlH 4 and CaH 2 reactions are non-agglomerated and they do not form impermeable mass. The experimental results show that the highest yield of hydrogen generation was achieved with LiAlH 4 and CaH 2 fuels.

  11. Source-Term and building-Wake Consequence Modeling for the Godiva IV Reactor at Los Alamos National Laboratory

    SciTech Connect

    Letellier, B.C.; McClure, P.; Restrepo, L.

    1999-06-13

    The objectives of this work were to evaluate the consequences of a postulated accident to onsite security personnel stationed near the facility during operations of the Godiva IV critical assembly and to identify controls needed to protect these personnel in case of an extreme criticality excursion equivalent to the design-basis accident (DBA). This paper presents the methodology and results of the source-term calculations, building ventilation rates, air concentrations, and consequence calculations that were performed using a multidisciplinary approach with several phenomenology models. Identification of controls needed to mitigate the consequences to near-field receptors is discussed.

  12. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    SciTech Connect

    Koch, M.; Kazimi, M.S.

    1991-04-01

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

  13. 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.

  14. 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

  15. 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.

  16. Confirmatory Survey Results for the Reactor Building Dome Upper Surfaces, Rancho Saco Nuclear Generating Station

    SciTech Connect

    Wade C. Adams

    2006-10-25

    Results from a confirmatory survey of the upper structural surfaces of the Reactor Building Dome at the Rancho Seco Nuclear Generating Station (RSNGS) performed by the Oak Ridge Institute for Science and Education for the NRC. Also includes results of interlaboratory comparison analyses on several archived soil samples that would be provided by RSNGS personnel. The confirmatory surveys were performed on June 7 and 8, 2006.

  17. 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.

  18. 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.

  19. 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)

  20. 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).

  1. 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.

  2. 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.

  3. Investigations of Alternative Steam Generator Location and Flatter Core Geometry for Lead-Cooled Fast Reactors

    SciTech Connect

    Carlsson, Johan; Tucek, Kamil; Wider, Hartmut

    2006-07-01

    This paper concerns two independent safety investigations on critical and sub-critical heavy liquid metal cooled fast reactors using simple flow paths. The first investigation applies to locating the steam generators in the risers instead of the down-comers of a simple flow path designed sub-critical reactor of 600 MW{sub th} power. This was compared to a similar design, but with the steam generators located in the downcomers. The transients investigated were Total-Loss-of-Power and unprotected Loss-Of-Flow. It is shown that this reactor peaks at 1041 K after 29 hours during a Total-Loss-Of-Power accident. The difference between locating the steam generators in the risers and the downcomers is insignificant for this accident type. During an unprotected Loss-Of-Flow accident at full power, the core outlet temperature stabilizes at 1010 K, which is 337 K above nominal outlet temperature. The second investigation concerns a 1426 MW{sub th} critical reactor where the influence of the core height versus the core outlet temperature is studied during an unprotected Loss-Of-Flow and Total-Loss-Of-Power accident. A pancake type core geometry of 1.0 m height and 5.8 m diameter, is compared to a compact core of 2 m height and 4.5 m diameter. Moderators, like BeO and hydrides, and their influence on safety coefficients and burnup swings are also presented. Both cores incinerate transuranics from spent LWR fuel with minor actinide fraction of 5%. We show that LFRs can be designed both to breed and burn transuranics from LWRs. It is shown that the hydrides lead to the most favorable reactivity feedbacks, but the poorest reactivity swing. The computational fluid dynamics code STAR-CD was used for all thermal hydraulic calculations, and the MCNP and MCB for neutronics, and burn-up calculations. (authors)

  4. Physico-chemical properties of the new generation IV iron preparations ferumoxytol, iron isomaltoside 1000 and ferric carboxymaltose.

    PubMed

    Neiser, Susann; Rentsch, Daniel; Dippon, Urs; Kappler, Andreas; Weidler, Peter G; Göttlicher, Jörg; Steininger, Ralph; Wilhelm, Maria; Braitsch, Michaela; Funk, Felix; Philipp, Erik; Burckhardt, Susanna

    2015-08-01

    The advantage of the new generation IV iron preparations ferric carboxymaltose (FCM), ferumoxytol (FMX), and iron isomaltoside 1000 (IIM) is that they can be administered in relatively high doses in a short period of time. We investigated the physico-chemical properties of these preparations and compared them with those of the older preparations iron sucrose (IS), sodium ferric gluconate (SFG), and low molecular weight iron dextran (LMWID). Mössbauer spectroscopy, X-ray diffraction, and Fe K-edge X-ray absorption near edge structure spectroscopy indicated akaganeite structures (β-FeOOH) for the cores of FCM, IIM and IS, and a maghemite (γ-Fe2O3) structure for that of FMX. Nuclear magnetic resonance studies confirmed the structure of the carbohydrate of FMX as a reduced, carboxymethylated, low molecular weight dextran, and that of IIM as a reduced Dextran 1000. Polarography yielded significantly different fingerprints of the investigated compounds. Reductive degradation kinetics of FMX was faster than that of FCM and IIM, which is in contrast to the high stability of FMX towards acid degradation. The labile iron content, i.e. the amount of iron that is only weakly bound in the polynuclear iron core, was assessed by a qualitative test that confirmed decreasing labile iron contents in the order SFG ≈ IS > LMWID ≥ FMX ≈ IIM ≈ FCM. The presented data are a step forward in the characterization of these non-biological complex drugs, which is a prerequisite to understand their cellular uptake mechanisms and the relationship between the structure and physiological safety as well as efficacy of these complexes. PMID:25801756

  5. 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

  6. Synergetic Action of Domain II and IV Underlies Persistent Current Generation in Nav1.3 as revealed by a tarantula toxin

    PubMed Central

    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

  7. 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)

  8. Production of tungsten-188 and osmium-194 in a nuclear reactor for new clinical generators

    SciTech Connect

    Mirzadeh, S.; Knapp, F.F. Jr.; Callahan, A.P.

    1991-01-01

    Rhenium-188 and iridium-194 are potential candidates for radioimmunotherapy with monoclonal antibodies directed against tumor-associated antigens. Both nuclei are short-lived and decay by high energy {Beta}{minus} emission. In addition, both nuclei emit {gamma}-rays with energy suitable for imaging. An important characteristics is availability of {sup 188}Re and {sup 194}Ir from decay of reactor-produced parents ({sup 188}W and {sup 194}Os, respectively) in convenient generator systems. The {sup 188}W and {sup 194}Os are produced by double neutron capture of {sup 186}W and {sup 192}Os, respectively. The large scale production yields of {sup 188}W in several nuclear reactors will be presented. We also report a new management for the cross-section of {sup 193}Os(n,{gamma}){sup 194}Os reaction and discuss the feasibility of producing sufficient quantities of {sup 194}Os. 17 refs., 1 fig., 2 tabs.

  9. 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.

  10. 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

  11. 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.

  12. 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)

  13. 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.

  14. 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

  15. 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.

  16. 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.

  17. 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

  18. 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.

  19. 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.

  20. Radwaste generation survey update: Volume 2, Pressurized water reactors: Final report

    SciTech Connect

    Daloisio, G.S.; Deltete, C.P.

    1988-02-01

    The Electric Power Research Institute (EPRI) commissioned an operations-related project (RP1557-26) in mid-1986 to update the project data base developed for EPRI Report NP-3370, ''Identification of Radwaste Sources and Reduction Techniques,'' which was published in January 1984. An update was deemed particularly desirable in order to assess the impact on power reactor low level radioactive waste generation of 10 CFR 61, the recent implementation of the 1985 Amendment to the Low Level Waste Policy Act of 1980 (and its potential effects on accelerated waste shipment programs), and the efforts of several plants to implement waste minimization program over the past several years. These events, as reflected in waste generation rates from 1982 through 1985, should help NP-3370 continue to be a useful document for a plant's radwaste manager in the future. Furthermore, the trends of the past several years presented herein should help to more accurately define utility waste source terms for use in planning on-site storage and developing regional burial facilities. A new data base was developed that includes 1982 through 1986 information, as well as pertinent portions of the 1978 through 1981 data base. The result of the project is a two volume report comprising radwaste related information from more than 95% of the nuclear power plants in commerical operation as of 1986. Volume 1 contains all information pertaining to boiling water reactors (BWRs), while Volume 2 contains information for pressurized water reactors (PWRs). The computerized data base of waste volumes, sources and characteristics for each plant type (BWR or PWR) is included as an appendix in each respective volume. 36 figs., 26 tabs.

  1. Radwaste generation survey update: Volume 1, Boiling water reactors: Final report

    SciTech Connect

    Daloisio, G.S.; Deltete, C.P.

    1988-02-01

    The Electric Power Research Institute (EPRI) commissioned an operations-related project (RP1557-26) in mid-1986 to update the project data base developed for EPRI Report NP-3370, ''Identification of Radwaste Sources and Reduction Techniques,'' which was published in January 1984. An update was deemed particularly desirable in order to assess the impact on power reactor low level radioactive waste generation of 10 CFR 61, the recent implementation of the 1985 Amendment to The Low Level Waste Policy Act of 1980 (and its potential effects on accelerated waste shipment programs), and the efforts of several plants to implement waste minimization programs over the past several years. These events, as reflected in waste generation rates from 1982 through 1986, should help NP-3370 continue to be a useful document for a plant's radwaste manager in the future. Furthermore, the trends of the past several years presented herein should help to more accurately define utility waste source terms for use in planning on-site storage and developing regional burial facilities. A new data base was developed that includes 1982 through 1986 information, as well as pertinent portions of the 1978 through 1981 data base. The result of the project is a two volume report comprising radwaste related information from more than 95% of the nuclear power plants in commercial operation as of 1986. Volume 1 contains all information pertaining to boiling water reactors (BWRs), while Volume 2 contains information for pressurized water reactors (PWRs). The computerized data base of waste volumes, sources and characteristics for each plant type (BWR or PWR) is included as an appendix in each respective volume.

  2. Improvement of Dielectric Barrier Discharge Plasma Reactor for Ozone Generation by Electrode Shape

    NASA Astrophysics Data System (ADS)

    Shimizu, Masaki; Sato, Tohru; Kato, Shoji; Mukaigawa, Seiji; Takaki, Koichi; Fujiwara, Tamiya

    An effect of electrode shape on ozone generation in dielectric barrier discharge reactor is described in this article. Three different shape electrodes were employed as ground electrodes. A plane electrode is 6 cm in width, and 20 cm in length. A trench electrode has large number of knife-edge rails. A multipoint electrode has large number of four-sided pyramid projections on the plane. A high voltage plane electrode is covered with 0.5 mm thickness alumina layer worked as dielectric barrier. The experimental results show that the breakdown for the multipoint electrode occurs at 7.0 kVpp. This value is lower than 8.4 kVpp that is the breakdown voltage of the plane electrode. The ozone yield increases from 80 g/kWh to 130 g/kWh by changing the electrode shape from the plane to the multipoint. The ozone generation efficiency decreased with increase of the ozone concentration.

  3. 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

  4. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  5. 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

  6. 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.

  7. 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

  8. Ultrasound pressure distributions generated by high frequency transducers in large reactors.

    PubMed

    Leong, Thomas; Coventry, Michael; Swiergon, Piotr; Knoerzer, Kai; Juliano, Pablo

    2015-11-01

    The performance of an ultrasound reactor chamber relies on the sound pressure level achieved throughout the system. The active volume of a high frequency ultrasound chamber can be determined by the sound pressure penetration and distribution provided by the transducers. This work evaluated the sound pressure levels and uniformity achieved in water by selected commercial scale high frequency plate transducers without and with reflector plates. Sound pressure produced by ultrasonic plate transducers vertically operating at frequencies of 400 kHz (120 W) and 2 MHz (128 W) was characterized with hydrophones in a 2 m long chamber and their effective operating distance across the chamber's vertical cross section was determined. The 2 MHz transducer produced the highest pressure amplitude near the transducer surface, with a sharp decline of approximately 40% of the sound pressure occurring in the range between 55 and 155 mm from the transducer. The placement of a reflector plate 500 mm from the surface of the transducer was shown to improve the sound pressure uniformity of 2 MHz ultrasound. Ultrasound at 400 kHz was found to penetrate the fluid up to 2 m without significant losses. Furthermore, 400 kHz ultrasound generated a more uniform sound pressure distribution regardless of the presence or absence of a reflector plate. The choice of the transducer distance to the opposite reactor wall therefore depends on the transducer plate frequency selected. Based on pressure measurements in water, large scale 400 kHz reactor designs can consider larger transducer distance to opposite wall and larger active cross-section, and therefore can reach higher volumes than when using 2 MHz transducer plates. PMID:26186816

  9. REACTOR

    DOEpatents

    Spitzer, L. Jr.

    1962-01-01

    The system conteraplates ohmically heating a gas to high temperatures such as are useful in thermonuclear reactors of the stellarator class. To this end the gas is ionized and an electric current is applied to the ionized gas ohmically to heat the gas while the ionized gas is confined to a central portion of a reaction chamber. Additionally, means are provided for pumping impurities from the gas and for further heating the gas. (AEC)

  10. 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.

  11. 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

  12. CHLORINE ABSORPTION IN S(IV) SOLUTIONS

    EPA Science Inventory

    The rate of chlorine (Cl2) absorption into aqueous sulfite/bisulfite [S(IV)] solutions was measured at ambient temperature using a highly characterized stirred cell reactor. The reactor media were 0 to 10 mM S(IV) with pH ranging from 3.5 to 8.5. Experiments were performed using ...

  13. 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...

  14. 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.

  15. 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.

  16. Recovery of tritium dissolved in sodium at the steam generator of fast breeder reactor

    SciTech Connect

    Oya, Y.; Oda, T.; Tanaka, S.; Okuno, K.

    2008-07-15

    The tritium recovery technique in steam generators for fast breeder reactors using the double pipe concept was proposed. The experimental system for developing an effective tritium recovery technique was developed and tritium recovery experiments using Ar gas or Ar gas with 10-10000 ppm oxygen gas were performed using D{sub 2} gas instead of tritium gas. It was found that deuterium permeation through two membranes decreased by installing the double pipe concept with Ar gas. By introducing Ar gas with 10000 ppm oxygen gas, the concentration of deuterium permeation through two membranes decreased by more than 1/200, compared with the one pipe concept, indicating that most of the deuterium was scavenged by Ar gas or reacted with oxygen to form a hydroxide. However, most of the hydroxide was trapped at the surface of the membranes because of the short duration of the experiment. (authors)

  17. Evaluation of cracking in steam generator feedwater piping in pressurized water reactor plants

    SciTech Connect

    Goldberg, A.; Streit, R.D.

    1981-05-01

    Cracking in feedwater piping was detected near the inlet to steam generators in 15 pressurized water reactor plants. Sections with cracks from nine plants are examined with the objective of identifying the cracking mechanism and assessing various factors that might contribute to this cracking. Using transmission electron microscopy, fatigue striations are observed on replicas of cleaned crack surfaces. Calculations based on the observed striation spacings gave a cyclic stress value of 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 and it is concluded that the overriding factor in the cracking problem was the presence of such undocumented cyclic loads.

  18. 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.

  19. Structure interaction due to thermal bowing of shrouds in steam generator of gas-cooled reactor

    SciTech Connect

    Woo, H.H.

    1981-01-01

    The design of the gas-cooled reactor steam generators includes a tube bundle support plate system which restrains and supports the helical tubes in the steam generator. The support system consists of an array of radially oriented, perforated plates through which the helical tube coils are wound. These support plates have tabs on their edges which fit into vertical slots in the inner and outer shrouds. When the helical tube bundle and support plates are installed in the steam generator, they most likely cannot fit evenly between the inner and outer shrouds. This imperfection leads to different gaps between two extreme sides of the tube bundle and the shrouds. With different gaps through the tube bundle height, the helium flow experiences different cooling effects from the tube bundle. Hence, the temperature distribution in the shrouds will be non-uniform circumferentially since their surrounding helium flow temperatures are varied. These non-uniform temperatures in the shrouds result in the phenomenon of thermal bowing of shrouds.

  20. Lessons Learned From Gen I Carbon Dioxide Cooled Reactors

    SciTech Connect

    David E. Shropshire

    2004-04-01

    This paper provides a review of early gas cooled reactors including the Magnox reactors originating in the United Kingdom and the subsequent development of the Advanced Gas-cooled Reactors (AGR). These early gas cooled reactors shared a common coolant medium, namely carbon dioxide (CO2). A framework of information is provided about these early reactors and identifies unique problems/opportunities associated with use of CO2 as a coolant. Reactor designers successfully rose to these challenges. After years of successful use of the CO2 gas cooled reactors in Europe, the succeeding generation of reactors, called the High Temperature Gas Reactors (HTGR), were designed with Helium gas as the coolant. Again, in the 21st century, with the latest reactor designs under investigation in Generation IV, there is a revived interest in developing Gas Cooled Fast Reactors that use CO2 as the reactor coolant. This paper provides a historical perspective on the 52 CO2 reactors and the reactor programs that developed them. The Magnox and AGR design features and safety characteristics were reviewed, as well as the technologies associated with fuel storage, reprocessing, and disposal. Lessons-learned from these programs are noted to benefit the designs of future generations of gas cooled nuclear reactors.

  1. New generation of NPP with boiling water reactor of improved safety

    SciTech Connect

    Adamov, E.O.; Kuklin, A.N.; Mityaev, Yu.I.; Mikhan, V.I.; Tokarev, Yu.I.; Cherkashov, Yu.M.; Sokolov, I.N.; Iljin, Yu.V.; Pakh, E.E.; Abramov, V.I.

    1993-12-31

    The nuclear power plants with boiling water reactors of improved safety are being developed. There is 26 years of operating experience with the plant VK-50 in Dimitrovgrad. The design and operation of the BWR reactors are described.

  2. Measures for ensuring reliable operation of the welded joint connecting the reactor coolant circuit's header to the shell of a steam generator used at a VVER-1000 reactor-based nuclear power station

    NASA Astrophysics Data System (ADS)

    Kharchenko, S. A.; Trunov, N. B.; Korotaev, N. F.; Lyakishev, S. L.

    2011-03-01

    Problems that arose around the weld joint connecting the reactor coolant circuit's header to the steam generator shell during operation of steam generators at nuclear power stations equipped with VVER-1000 reactors are considered. Works on studying the defects occurred in the header's metal are described, and ways for preventing their development are determined.

  3. 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

  4. 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.

  5. Testing the improved method for calculating the radiation heat generation at the periphery of the BOR-60 reactor core

    SciTech Connect

    Varivtsev, A. V. Zhemkov, I. Yu.

    2014-12-15

    The application of the improved method for calculating the radiation heat generation in the elements of an experimental device located at the periphery of the BOR-60 reactor core results in a significant reduction in the discrepancies between the calculated and the experimental data. This allows us to conclude that the improved method has an advantage over the one used earlier.

  6. 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.

  7. 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.

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

    DOE PAGESBeta

    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

  9. 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.

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

    Energy Science and Technology Software Center (ESTSC)

    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

  11. A Compact Torus Fusion Reactor Utilizing a Continuously Generated String of CT's. The CT String Reactor, CTSR

    NASA Astrophysics Data System (ADS)

    Hartman, Charles W.; Reisman, David B.; McLean, Harry S.; Thomas, John

    2008-06-01

    A fusion reactor is described in which a moving string of mutually repelling compact toruses (alternating helicity, unidirectional Btheta) is generated by repetitive injection using a magnetized coaxial gun driven by continuous gun current with alternating poloidal field. An injected CT relaxes to a minimum magnetic energy equilibrium, moves into a compression cone, and enters a conducting cylinder where the plasma is heated to fusion-producing temperature. The CT then passes into a blanketed region where fusion energy is produced and, on emergence from the fusion region, the CT undergoes controlled expansion in an exit cone where an alternating poloidal field opens the flux surfaces to directly recover the CT magnetic energy as current which is returned to the formation gun. The CT String Reactor (CTSTR) reactor satisfies all the necessary MHD stability requirements and is based on extrapolation of experimentally achieved formation, stability, and plasma confinement. It is supported by extensive 2D, MHD calculations. CTSTR employs minimal external fields supplied by normal conductors, and can produce high fusion power density with uniform wall loading. The geometric simplicity of CTSTR acts to minimize initial and maintenance costs, including periodic replacement of the reactor first wall. 60 /proj/ads/abstracts/absload/springerxml.pl* /proj/ads/abstracts/sources/SPRINGER/files/JOU=10894/VOL=2008.27/ISU=3/ART=9121: total 28 28 10894_2007_Article_9121.xml.meta /proj/ads/abstracts/sources/SPRINGER/files/JOU=10894/VOL=2008.27/ISU=3/ART=9122: total 16 16 10894_2007_Article_9122.xml.meta /proj/ads/abstracts/sources/SPRINGER/files/JOU=10894/VOL=2008.27/ISU=3/ART=9123: total 12 12 10894_2007_Article_9123.xml.meta /proj/ads/abstracts/sources/SPRINGER/files/JOU=10894/VOL=2008.27/ISU=3/ART=9124: total 36 36 10894_2007_Article_9124.xml.meta /proj/ads/abstracts/sources/SPRINGER/files/JOU=10894/VOL=2008.27/ISU=3/ART=9125: total 32 32 10894_2007_Article_9125.xml.meta /proj

  12. 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

  13. 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.

  14. Sensitivity Analysis of Neutron Cross-Sections Considered for Design and Safety Studies of Lfr and SFR Generation IV Systems

    NASA Astrophysics Data System (ADS)

    Tucek, Kamil; Carlsson, Johan; Wider, Hartmut

    2006-04-01

    We evaluated the sensitivity of several design and safety parameters with regard to five different nuclear data libraries, JEF2.2, JEFF3.0, ENDF/B-VI.8, JENDL3.2, and JENDL3.3. More specifically, the effective multiplication factor, burn-up reactivity swing and decay heat generation in available LFR and SFR designs were estimated. Monte Carlo codes MCNP and MCB were used in the analyses of the neutronic and burn-up performance of the systems. Thermo-hydraulic safety calculations were performed by the STAR-CD CFD code. For the LFR, ENDF/B-VI.8 and JEF2.2 showed to give a harder neutron spectrum than JEFF3.0, JENDL3.2, and JENDL3.3 data due to the lower inelastic scattering cross-section of lead in these libraries. Hence, the neutron economy of the system becomes more favourable and keff is higher when calculated with ENDF/B-VI.8 and JEF2.2 data. As for actinide cross-section data, the uncertainties in the keff values appeared to be mainly due to 239Pu, 240Pu and 241Am. Differences in the estimated burn-up reactivity swings proved to be significant, for an SFR as large as a factor of three (when comparing ENDF/B-VI.8 results to those of JENDL3.2). Uncertainties in the evaluation of short-term decay heat generation showed to be of the order of several per cent. Significant differences were, understandably, observed between decay heat generation data quoted in literature for LWR-UOX and those calculated for an LFR (U,TRU)O2 spent fuel. A corresponding difference in calculated core parameters (outlet coolant temperature) during protected total Loss-of-Power was evaluated.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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…

  2. 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)

  3. 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

  4. A method of examining iron oxides speciation and transport to steam generators during nuclear power reactor startups

    NASA Astrophysics Data System (ADS)

    Sawicki, Jerzy A.; Sawicka, Barbara D.; Price, James E.

    2010-12-01

    Secondary side corrosion products (sludge) collected during one of CANDU1 reactor startups from wet layup have been examined by X-ray fluorescence and Mössbauer spectroscopy. The transport and chemical form of iron oxides and oxyhydroxides were determined in condensate, feedwater and preheater outlet as a function of temperature and time. The sludge burst and oxidation states of iron oxides were correlated with the rise of reactor power and corresponding changes in temperature, condensate vacuum and water flow rate. In particular, a sharp γ-FeOOH to Fe 3O 4 switch was observed that coincided in time with the onset of condensate vacuum. Also, it was found that the startup after wet layup is characterized by only brief and fairly small sludge burst at about 30% reactor power and which contributes only a small amount of undesirable α-Fe 2O 3 to total iron transport to steam generator. Thus, sludge burden to steam generators can be minimized with proper layup and startup practices. ™ Trademark of Atomic Energy of Canada Limited.

  5. The Results of Feasibility Study of Co-generation NPP With Innovative VK-300 Simplified Boiling Water Reactor

    SciTech Connect

    Kuznetsov, Yury N.

    2006-07-01

    The co-generation nuclear power plant (CNPP) producing electricity and district heating heat is planned to be constructed in Archangelsk Region of Russia. Following the 'Letter of Intent' signed by Governor of Archangelsk region and by Minister of the Russian Federation for atomic energy the feasibility study of the Project has been done. The NPP will be based on the four co-generation nuclear power units with the Russian VK-300 SBWR. The innovative passive VK-300 reactor facility has been designed on the basis of well-established nuclear technologies, proven major components, the operating experience of the prototype VK-50 reactor in RIAR, Dimitrovgrad, and the experience in designing such reactors as SBWR (GE) and SWR-1000 (Siemens). The CNPP's total power is planned to be 1000 MW(e) and district-heating heat production capacity 1600 Gcal/h. A detailed description of the results of the feasibility study is presented in the report. The results of the feasibility study have shown that the Archangelsk CGNP is feasible in terms of engineering, economics and production. (authors)

  6. 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.

  7. 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...

  8. 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.

  9. Next generation sequencing of stage IV squamous cell lung cancers reveals an association of PI3K aberrations and evidence of clonal heterogeneity in patients with brain metastases

    PubMed Central

    Paik, Paul K.; Shen, Ronglai; Won, Helen; Rekhtman, Natasha; Wang, Lu; Sima, Camelia S.; Arora, Arshi; Seshan, Venkatraman; Ladanyi, Marc; Berger, Michael F.; Kris, Mark G.

    2015-01-01

    Large-scale genomic characterization of squamous cell lung cancers (SQCLC) has revealed several putative oncogenic drivers. There are, however, little data to suggest that these alterations have clinical relevance. We performed comprehensive genomic profiling of 79 stage IV SQCLCs (including next-generation sequencing) and analyzed differences in the clinical characteristics of two major SQCLC subtypes: FGFR1 amplified and PI3K aberrant. Patients with PI3K aberrant tumors had aggressive disease marked by worse survival (median OS 8.6 vs. 19.1 mo, p<0.001), higher metastatic burden (>3 organs 18% vs. 3%, p=0.025), and greater incidence of brain metastases (27% vs. 0% in others, p<0.001). We performed whole-exome and RNA sequencing on paired brain metastases and primary lung cancers to elucidate the metastatic process to brain. SQCLC primaries that gave rise to brain metastases exhibited truncal PTEN loss. SQCLC brain metastases exhibited a high degree of genetic heterogeneity and evidence of clonal differences between their primary sites. PMID:25929848

  10. 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.

  11. Behaviour of palladium(II), platinum(IV), and rhodium(III) in artificial and natural waters: influence of reactor surface and geochemistry on metal recovery.

    PubMed

    Cobelo-Garcia, Antonio; Turner, Andrew; Millward, Geoffrey E; Couceiro, Fay

    2007-03-01

    The recovery of dissolved platinum group elements (PGE: Pd(II), Pt(IV) and Rh(III)) added to Milli-Q water, artificial freshwater and seawater and filtered natural waters has been studied, as a function of pH and PGE concentration, in containers of varying synthetic composition. The least adsorptive and/or precipitative loss was obtained for borosilicate glass under most of the conditions employed, whereas the greatest loss was obtained for low-density polyethylene. Of the polymeric materials tested, the adsorptive and/or precipitative loss of PGE was lowest for fluorinated ethylene propylene (Teflon). The loss of Pd(II) in freshwater was significant due to its affinity for surface adsorption and its relatively low solubility. The presence of natural dissolved organic matter increases the recovery of Pd(II) but enhances the loss of Pt(IV). The loss of Rh(III) in seawater was significant and was mainly due to precipitation, whereas Pd(II) recovery was enhanced, compared to freshwater, because of its complexation with chloride. The results have important implications regarding protocols employed for sample preservation and controlled laboratory experiments used in the study of the speciation and biogeochemical behaviour of PGE. PMID:17386666

  12. 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.

  13. 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.

  14. [Single-stage autotrophic nitrogen removal reactor with self-generated granular sludge for treating sludge dewatering effluent].

    PubMed

    Cao, Jian-ping; Du, Bing; Liu, Yin; Qin, Yong-sheng

    2009-10-15

    Single-stage autotrophic nitrogen removal (SANR) has been observed in a long-term operated nitrosation air-lift reactor for treating digested sludge dewatering effluent from sewage wastewater treatment plant. A kind of so called self-generated granular sludge which undertake the SANR reaction has oriented formed. The performance of SANR reactor cultivated above sludge for treating sludge dewatering effluent has been tested and better results have been reached. When the influent total nitrogen (TN) was kept about 350 mg/L (mainly ammonium nitrogen), the average TN removal efficiency and nitrogen removal load were 74.8% (maximum 86.92%) and 0.68 kg x (m3 x d)(-1) [maximum 0.9 kg x (m3 x d)(-1)] respectively. The operation stability and nitrogen removal efficiency have been enforced after adding a certain quantity powered activated carbon. The influent ammonium concentration, nitrogen load and aeration rate have a great effect on SANR reactor as well as the influent organic compound, pH, alkalinity have a relatively low effect. The parameters such as the ratios of aeration rate/deltaTN, aeration rate/deltaNH4+ -N, deltaALK/deltaTN can be used for better controlling the reaction. PMID:19968119

  15. Water mass control system based on artificial neural networks for the steam generator in a pressurized water reactor

    NASA Astrophysics Data System (ADS)

    Dong, Wei

    The control of water mass inventory and water level in the steam generator is important for nuclear power plant. Conventional control system designs attempt to maintain downcomer water level within a relatively narrow operational band. However, the water level measured in the downcomer can temporarily react in a manner opposite to water mass inventory changes, which is known as shrink and swell effects. As a result, automatic or manual control of water level can be difficult under these conditions and can lead to reactor trips. This research introduces a new feedwater control strategy for nuclear steam generators. By estimating the water mass inventory with neural networks, the new method directly controls water mass inventory by conventional PI controller. Since shrink and swell are eliminated in water mass control, theoretical analysis and simulation results show the new control strategy improves the operation of nuclear steam generators significantly. In the water mass control system design, the safety function of the system is still based on the Steam Generator Water Level. Thus, the conventional water level trips will protect the plant when the new control strategy fails to maintain the water level within the safety range. The water mass estimator can be embedded in the Instrumentation and Control System of a Nuclear Power Plant to open loop observe the Steam Generator water mass inventory, improving the safety of nuclear power plant operation. Closed loop water mass control for a Steam Generator can be implemented after the observed water mass shows good agreement with theoretical calculations and plant operation experiences.

  16. 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.

  17. 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

  18. Specific features of corrosion damage to heat-transfer tubes of steam generators used at nuclear power stations equipped with VVER-1000 reactors

    NASA Astrophysics Data System (ADS)

    Nemytov, D. S.; Tyapkov, V. F.

    2009-07-01

    Specific features of corrosion damage occurring to the heat-transfer tubes of steam generators used at nuclear power stations equipped with VVER-1000 reactors are considered. The results obtained from metallographic studies of flaws found in samples cut out from steam-generator tubes are analyzed. Regularities with which flaws of steam-generator tubes are distributed over the tube bundle volume are discussed. Approaches for assessing the technical state and remaining service life of steam-generator tubes are presented.

  19. 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.

  20. 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)

  1. 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.

  2. A Supercritical CO{sub 2} Gas Turbine Power Cycle for Next-Generation Nuclear Reactors

    SciTech Connect

    Dostal, Vaclav; Driscoll, Michael J.; Hejzlar, Pavel; Todreas, Neil E.

    2002-07-01

    Although proposed more than 35 years ago, the use of supercritical CO{sub 2} as the working fluid in a closed circuit Brayton cycle has so far not been implemented in practice. Industrial experience in several other relevant applications has improved prospects, and its good efficiency at modest temperatures (e.g., {approx}45% at 550 deg. C) make this cycle attractive for a variety of advanced nuclear reactor concepts. The version described here is for a gas-cooled, modular fast reactor. In the proposed gas-cooled fast breeder reactor design of present interest, CO{sub 2} is also especially attractive because it allows the use of metal fuel and core structures. The principal advantage of a supercritical CO{sub 2} Brayton cycle is its reduced compression work compared to an ideal gas such as helium: about 15% of gross power turbine output vs. 40% or so. This also permits the simplification of use of a single compressor stage without inter-cooling. The requisite high pressure ({approx}20 MPa) also has the benefit of more compact heat exchangers and turbines. Finally, CO{sub 2} requires significantly fewer turbine stages than He, its principal competitor for nuclear gas turbine service. One disadvantage of CO{sub 2} in a direct cycle application is the production of N-16, which will require turbine plant shielding (albeit much less than in a BWR). The cycle efficiency is also very sensitive to recuperator effectiveness and compressor inlet temperature. It was found necessary to split the recuperator into separate high-and low-temperature components, and to employ intermediate re-compression, to avoid having a pinch-point in the cold end of the recuperator. Over the past several decades developments have taken place that make the acceptance of supercritical CO{sub 2} systems more likely: supercritical CO{sub 2} pipelines are in use in the western US in oil-recovery operations; 14 advanced gas-cooled reactors (AGR) are employed in the UK at CO{sub 2} temperatures up to

  3. 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.

  4. 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.

  5. 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.

  6. 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

  7. 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

  8. 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)

  9. A low-cost municipal sewage treatment system with a combination of UASB and the "fourth-generation" downflow hanging sponge reactors.

    PubMed

    Tandukar, M; Uemura, S; Machdar, I; Ohashi, A; Harada, H

    2005-01-01

    This paper presents an evaluation of the process performance of a pilot-scale "fourth generation" downflow hanging sponge (DHS) post-treatment system combined with a UASB pretreatment unit treating municipal wastewater. After the successful operation of the second- and third-generation DHS reactors, the fourth-generation DHS reactor was developed to overcome a few shortcomings of its predecessors. This reactor was designed to further enhance the treatment efficiency and simplify the construction process in real scale, especially for the application in developing countries. Configuration of the reactor was modified to enhance the dissolution of air into the wastewater and to avert the possible clogging of the reactor especially during sudden washout from the UASB reactor. The whole system was operated at a total hydraulic retention time (HRT) of 8 h (UASB: 6 h and DHS: 2 h) for a period of over 600 days. The combined system was able to remove 96% of unfiltered BOD with only 9 mg/L remaining in the final effluent. Likewise, F. coli were removed by 3.45 log with the final count of 10(3) to 10(4) MPN/100 ml. Nutrient removal by the system was also satisfactory. PMID:16180445

  10. Hydrogen production by reforming of liquid hydrocarbons in a membrane reactor for portable power generation-Experimental studies

    NASA Astrophysics Data System (ADS)

    Damle, Ashok S.

    One of the most promising technologies for lightweight, compact, portable power generation is proton exchange membrane (PEM) fuel cells. PEM fuel cells, however, require a source of pure hydrogen. Steam reforming of hydrocarbons in an integrated membrane reactor has potential to provide pure hydrogen in a compact system. Continuous separation of product hydrogen from the reforming gas mixture is expected to increase the yield of hydrogen significantly as predicted by model simulations. In the laboratory-scale experimental studies reported here steam reforming of liquid hydrocarbon fuels, butane, methanol and Clearlite ® was conducted to produce pure hydrogen in a single step membrane reformer using commercially available Pd-Ag foil membranes and reforming/WGS catalysts. All of the experimental results demonstrated increase in hydrocarbon conversion due to hydrogen separation when compared with the hydrocarbon conversion without any hydrogen separation. Increase in hydrogen recovery was also shown to result in corresponding increase in hydrocarbon conversion in these studies demonstrating the basic concept. The experiments also provided insight into the effect of individual variables such as pressure, temperature, gas space velocity, and steam to carbon ratio. Steam reforming of butane was found to be limited by reaction kinetics for the experimental conditions used: catalysts used, average gas space velocity, and the reactor characteristics of surface area to volume ratio. Steam reforming of methanol in the presence of only WGS catalyst on the other hand indicated that the membrane reactor performance was limited by membrane permeation, especially at lower temperatures and lower feed pressures due to slower reconstitution of CO and H 2 into methane thus maintaining high hydrogen partial pressures in the reacting gas mixture. The limited amount of data collected with steam reforming of Clearlite ® indicated very good match between theoretical predictions and

  11. 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

  12. 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

  13. 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.

  14. Library of PWR (pressurized-water reactor) steam generator tubing samples: Final report

    SciTech Connect

    Albertin, L.; Clark, W.G. Jr.; Junker, W.R.; Kuchirka, P.J.; Madeyski, A.; Metala, M.J.; Taszarek, B.J.

    1988-01-01

    The PWR Steam Generator Tubing Sample Library is a Steam Generator Owners Group-EPRI program whose objective is to compile a library of well-characterized tubing samples to be used for performance evaluation of inspection systems and for training and qualification of signal interpretation systems. The library was created through the preparation of samples intended to replicate degradation encountered in actual field tubes. A limited number of tube segments removed from actual steam generators are included. Degradation categories include wear, pitting and fatigue cracks, as well as stress corrosion cracking (SCC) and intergranular attack (IGA). Eddy current and ultrasonic inspection techniques, along with supplementary radiography, dye penetrant, and optical techniques were used to characterize the library candidates. Advanced computer-aided NDE data collection, analysis and display techniques were used to assess test results. This report provides details of the library program, with major emphasis on the sampling protocol, characterization of degradation and recommendations for the use and future growth of the library. Also included is a compendium of steam generator tube degradation field observation, describing past destructive examinations of tubes removed for inspection from steam generators, and a description of a physical modeling approach, using mercury (metal) to assess the discontinuity characterization capabilities of a pancake-type eddy current probe. Computerized data analysis and display techniques were used to reconstruct the test results in both two-dimensional color-coded maps and three-dimensional pseudo-isometric plots.

  15. 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

  16. Stability of cement-glass packages containing sodium borate salt generated from pressurized water reactor power plants

    SciTech Connect

    Izumida, T.; Kawamura, F.; Chino, K.; Kikuchi, M.

    1987-08-01

    A new solidification technique using cement-glass, which is a mixture of sodium silicate and cement, was studied for solidification of sodium borate salt of liquid waste generated from pressurized water reactor plants. When the sodium borate salt was solidified with the cement-glass, the resulting package contained eight times more sodium borate than was found in cement because it did not interact with sodium borate. The leaching ratio of cesium ion from the cement-glass package was one-tenth that of cement. Its low leaching ratio was due to the high cesium adsorption ability of cement-glass. The ratio could be theoretically evaluated by considering the cesium adsorption-desorption equilibrium.

  17. Using A High-Temperature Hydrogen Co-Generation Reactor To Optimize Both Economic and Environmental Performance

    SciTech Connect

    Weimar, Mark R.; Wood, Thomas W.; Reichmuth, Barbara A.; Johnson, Wayne L.

    2004-07-01

    This paper reports the analysis of outcomes for a 3000 MWt HTGR plant, given price and cost assumptions, and determines what level of hydrogen and electricity production would optimize the plant economically and environmentally (carbon reduction). Coupling nuclear power with hydrogen production could fundamentally alter the character of the nuclear industry and nuclear technology’s development path. For this to happen, the hydrogen economy will have to be realized, and a new generation of reactors technically suitable to co-production of hydrogen with electric power must be developed and proven. The paper shows that the tradeoff between producing hydrogen through steam methane reformation and producing electricity is disproportionate and would require significant price increases for electricity to change the outcomes. It also found that estimate of shadow values for carbon credits was in the range now under discussion.

  18. 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

  19. 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.

  20. 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.

  1. Generation of a uranium plasma at near gaseous core reactor conditions.

    NASA Technical Reports Server (NTRS)

    Davis, J. F., III; Schnitzler, B. G.; Schneider, R. T.

    1971-01-01

    A constricted sliding spark discharge is used to generate a high density, high temperature uranium plasma. Uranium particle densities up to 10 to the 20th power per cu cm are obtained over a temperature range of 30,000 to 50,000 K. The device consists of a capillary discharge channel lined with pressed and sintered UO2. A 250 joule capacitor bank is discharged into the channel, producing a plasma of 10-20 microsec duration. Spectroscopic observations are made over the spectral range of 1300 to 2500 A.

  2. 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

  3. 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.

  4. Seismic Safety Margins Research Programs. Assessment of potential increases in risk due to degradation of steam generator and reactor coolant pump supports. [PWR

    SciTech Connect

    Bohn, M. P.; Wells, J. E.; Shieh, L. C.; Cover, L. E.; Streit, R. L.

    1983-08-01

    During the NRC licensing review for the North Anna Units 1 and 2 pressurized-water reactors (PWRs), questions were raised regarding the potential for low-fracture toughness of steam-generator and reactor-coolant-pump supports. Because other PWRs may face similar problems, this issue was incorporated into the NRC Program for Resolution of Generic Issues. The work described in this report was performed to provide the NRC with a quantitative evaluation of the value/impact implications of the various options of resolving the fracture-toughness question. This report presents an assessment of the probabilistic risk associated with nil-ductility failures of steam-generator and reactor-coolant-pump structural-support systems during seismic events, performed using the Seismic Safety Margins Research Program codes and data bases.

  5. 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.

  6. 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.

  7. 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

  8. A theoretical investigation on the Strecker reaction catalyzed by a Ti(IV)-complex catalyst generated from a cinchona alkaloid, achiral substituted 2,2'-biphenol, and tetraisopropyl titanate.

    PubMed

    Su, Zhishan; Li, Weiyi; Wang, Jun; Hu, Changwei; Feng, Xiaoming

    2013-01-28

    The mechanism and the origin of selectivity of the asymmetric Strecker reaction catalyzed by a Ti(IV)-complex catalyst generated from a cinchona alkaloid, achiral substituted 2,2'-biphenol, and tetraisopropyl titanate have been investigated by DFT and ONIOM methods. The calculations indicate that the reaction proceeds through a dual activation mechanism, in which Ti(IV) acts as Lewis acid to activate the electrophile aldimine substrate, whereas the tertiary amine in cinchona alkaloid works as Lewis base to promote the activation and isomerization of HCN. The C-C bond formation step is predicted to be the selectivity-controlling step in the reaction with an energy barrier of 9.3 kcal mol(-1). The "asymmetric activation" is achieved by the transfer of asymmetry from the chiral cinchonine ligand to the axially flexible achiral biphenol ligand through coordination interaction with the central metal Ti(IV) . The large steric hindrance from the 3,3'-position substitute of biphenol, combined with the quinoline fragment of cinchona alkaloid and the orientation of hydrogen bonding of iPrOH, play a key role in controlling the stereoselectivity, which is in good agreement with the experimental observations. PMID:23239409

  9. Gas-Fast Reactor Fuel Fabrication

    SciTech Connect

    Randall Fielding; Mitchell Meyer; Ramprashad Prabhakaran; Jim Miller; Sean McDeavitt

    2005-11-01

    The gas-cooled fast reactor is a high temperature helium cooled Generation IV reactor concept. Operating parameters for this type of reactor are well beyond those of current fuels so a novel fuel must be developed. One fuel concept calls for UC particles dispersed throughout a SiC matrix. This study examines a hybrid reaction bonding process as a possible fabrication route for this fuel. Processing parameters are also optimized. The process combines carbon and SiC powders and a carbon yielding polymer. In order to obtain dense reaction bonded SiC samples the porosity to carbon ratio in the preform must be large enough to accommodate SiC formation from the carbon present in the sample, however too much porosity reduces mechanical integrity which leads to poor infiltration properties . The porosity must also be of a suitable size to allow silicon transport throughout the sample but keep residual silicon to a minimum.

  10. 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.

  11. 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.

  12. 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.

  13. Computational Flow Predictions for the Lower Plenum of a High-Temperature, Gas-Cooled Reactor

    SciTech Connect

    Not Available

    2006-11-01

    Advanced gas-cooled reactors offer the potential advantage of higher efficiency and enhanced safety over present day nuclear reactors. Accurate simulation models of these Generation IV reactors are necessary for design and licensing. One design under consideration by the Very High Temperature Reactor (VHTR) program is a modular, prismatic gas-cooled reactor. In this reactor, the lower plenum region may experience locally high temperatures that can adversely impact the plant's structural integrity. Since existing system analysis codes cannot capture the complex flow effects occurring in the lower plenum, computational fluid dynamics (CFD) codes are being employed to model these flows [1]. The goal of the present study is to validate the CFD calculations using experimental data.

  14. Computational Flow Predictions for the Lower Plenum of a High-Temperature, Gas-Cooled Reactor

    SciTech Connect

    Donna Post Guillen

    2006-11-01

    Advanced gas-cooled reactors offer the potential advantage of higher efficiency and enhanced safety over present day nuclear reactors. Accurate simulation models of these Generation IV reactors are necessary for design and licensing. One design under consideration by the Very High Temperature Reactor (VHTR) program is a modular, prismatic gas-cooled reactor. In this reactor, the lower plenum region may experience locally high temperatures that can adversely impact the plant’s structural integrity. Since existing system analysis codes cannot capture the complex flow effects occurring in the lower plenum, computational fluid dynamics (CFD) codes are being employed to model these flows [1]. The goal of the present study is to validate the CFD calculations using experimental data.

  15. 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.

  16. 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.

  17. 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).

  18. 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 ...

  19. Conceptual Design of a Very High Temperature Pebble-Bed Reactor

    SciTech Connect

    Hans D. Gougar; A. M. Ougouag; Richard M. Moore; W. K. Terry

    2003-11-01

    Efficient electricity and hydrogen production distinguish the Very High Temperature Reactor as the leading Generation IV advanced concept. This graphite-moderated, helium-cooled reactor achieves a requisite high outlet temperature while retaining the passive safety and proliferation resistance required of Generation IV designs. Furthermore, a recirculating pebble-bed VHTR can operate with minimal excess reactivity to yield improved fuel economy and superior resistance to ingress events. Using the PEBBED code developed at the INEEL, conceptual designs of 300 megawatt and 600 megawatt (thermal) Very High Temperature Pebble-Bed Reactors have been developed. The fuel requirements of these compare favorably to the South African PBMR. Passive safety is confirmed with the MELCOR accident analysis code.

  20. 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 Central

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

    2013-01-01

    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 FeIII2-Y• cofactor in the NrdB subunit of the class Ia RNRs can be generated by self-assembly from FeII2-NrdB, O2, and a reducing equivalent. By contrast, the structurally homologous class Ib enzymes require a MnIII2-Y• cofactor in their NrdF subunit. MnII2-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 MnIII2-Y• cofactor. Here we investigate the mechanism of assembly of the MnIII2-Y• cofactor in Bacillus subtilis NrdF. Cluster assembly from MnII2-NrdF, NrdIhq, and O2 has been studied by stopped flow absorption and rapid freeze quench EPR spectroscopies. The results support a mechanism in which NrdIhq reduces O2 to O2•− (40-48 s−1, 0.6 mM O2), the O2•− channels to and reacts with MnII2-NrdF to form a MnIIIMnIV intermediate (2.2 ± 0.4 s−1), and the MnIIIMnIV 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 MnII2 cluster with O2 and satisfies the requirement for an “extra” reducing equivalent in Y• generation. PMID:23402532

  1. 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.

  2. 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)

  3. 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...

  4. Preliminary Results From High Temperature Scoping Irradiation Experiments Of Selected Gen IV Structural Metallic Materials

    SciTech Connect

    Nanstad, Randy K; McClintock, David A; Hoelzer, David T

    2007-01-01

    The performance of Generation IV reactors as a class will be determined by the behavior of advanced engineering materials. In the case of materials utilized for reactor internals and pressure vessels, the effects of irradiation are major issues. The environmental conditions for most of the Gen IV reactors are generally beyond present day reactor technology, especially as regards the combinations of operating temperatures, reactor coolant characteristics, and neutron spectra. In some of the applications, the conditions lay well beyond advanced research programs in radiation effects on materials. Therefore, new experimental data as well as analytical predictions of expected behavior of candidate materials at conditions for which there are no experimental data will be required. In the Gen IV Materials Program cross-cutting task, plans are being developed and irradiations and testing are being carried out to address the issues described above. This paper provides preliminary results for the first series of scoping irradiation experiments with selected metallic alloys, some of which are considered candidate materials for current Gen IV reactor applications, while others are considered as potential future candidate materials. The material classes represented are (1) nickel-base alloys (alloy 800H and Inconel 617; (2) advanced oxide-dispersion strengthened steels (14WT and 14YWT); and (3) commercial ferritic-martensitic steels (9Cr-1MoV). The results presented are from tensile tests using small flat tensile specimens (SS-3) in both the unirradiated and irradiated conditions. Specimens were irradiated in so-called rabbit capsules in the High-Flux Isotope Reactor (HFIR) at temperatures from 550 to 750 C and to irradiation doses from about 1.28 to 1.61 dpa. For the preliminary results from the first phase of this study, the annealed 9Cr-1MoV shows small amounts of irradiation-induced hardening. For the Alloy 800H, however, the hardening resulting from the 580 C irradiation

  5. 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.

  6. 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.

  7. 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].

  8. 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].

  9. 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.

  10. 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.

  11. 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.

  12. 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. PMID:24291312

  13. 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…

  14. IV treatment at home

    MedlinePlus

    ... 24 hours a day. If there is a problem with the IV, you can call your home health care agency for help. If the IV comes out of ... bleeding stops. Then call the home health care agency or the doctor right away.

  15. 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

  16. 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

  17. 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.

  18. Generation and Retention of Helium and Hydrogen in Austenitic Steels Irradiated in a Variety of LWR and Test Reactor Spectral Environments

    SciTech Connect

    Garner, Francis A.; Oliver, Brian M.; Greenwood, Lawrence R.; Edwards, Danny J.; Bruemmer, Stephen M.; Grossbeck, Martin L.

    2002-03-31

    In fission and fusion reactor environments stainless steels generate significant amounts of helium and hydrogen by transmutation. The primary sources of helium are boron and nickel, interacting with both fast and especially thermal neutrons. Hydrogen arises primarily from fast neutron reactions, but is also introduced into steels at often much higher levels by other environmental processes. Although essentially all of the helium is retained in the steel, it is commonly assumed that most of the hydrogen is not retained. It now appears that under some circumstances, significant levels of hydrogen can be retained, especially when helium-nucleated cavities become a significant part of the microstructure. A variety of stainless steel specimens have been examined from various test reactors, PWRs and BWRs. These specimens were exposed to a wide range of neutron spectra with different thermal/fast neutron ratios. Pure nickel and pure iron have also been examined. It is shown that all major features of the retention of helium and hydrogen can be explained in terms of the composition, thermal/fast neutron ratio and the presence or absence of helium-nucleated cavities. In some cases, the hydrogen retention is very large and can exceed that generated by transmutation, with the additional hydrogen arising from either environmental sources and/or previously unidentified radioisotope sources that may come into operation at high neutron exposures.

  19. 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. PMID:24527646

  20. 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.

  1. Research at the CEA in the field of safety in 2nd and 3rd generation light water reactors

    NASA Astrophysics Data System (ADS)

    Billot, Philippe

    2012-05-01

    The research programs at the CEA in the field of safety in nuclear reactors are carried out in a framework of international partnerships. Their purpose is to develop studies on: The methods allowing for the determination of earthquake hazards and their consequences; The behaviour of fuel in an accident situation; The comprehension of deflagration and detonation phenomena of hydrogen and the search for effective prevention methods involving an explosion risk; The cooling of corium in order to stop its progression in and outside the vessel thereby reducing the risk of perforating the basemat; The behaviour of the different fission product families according to their volatility for the UO2 and MOX fuels.

  2. Catalytic reactor

    DOEpatents

    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.

  3. 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

  4. 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 ...

  5. 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.

  6. 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.

  7. 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.

  8. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production, Progress Report for Work Through September 2003, 2nd Annual/8th Quarterly Report

    SciTech Connect

    Philip E. MacDonald

    2003-09-01

    The supercritical water-cooled reactor (SCWR) is one of the six reactor technologies selected for research and development under the Generation-IV program. SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% vs. about 33% efficiency for current Light Water Reactors, LWRs) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once-through cycle. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus the need for recirculation and jet pumps, a pressurizer, steam generators, steam separators and dryers is eliminated. The main mission of the SCWR is generation of low-cost electricity. It is built upon two proven technologies, LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil-fired boilers, a large number of which is also in use around the world.

  9. 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.

  10. 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.

  11. Final Report (1994 to 1996) Diagnostic of the Spatial and Velocity Distribution of Alpha Particles in Tokamak Fusion Reactor using Beat-wave Generated Lower Hybrid Wave

    SciTech Connect

    Hwang, D.Q.; Horton, R.D.; Evans, R.W.

    1999-06-03

    The alpha particles in a fusion reactor play a key role in the sustaining the fusion reaction. It is the heating provided by the alpha particles that help a fusion reactor operating in the ignition regime. It is, therefore, essential to understand the behavior of the alpha population both in real space and velocity space in order to design the optimal confinement device for fusion application. Moreover, the alphas represent a strong source of free energy that may generate plasma instabilities. Theoretical studies has identified the Toroidal Alfven Eigenmode (TAE) as an instability that can be excited by the alpha population in a toroidal device. Since the alpha has an energy of 3.5 MeV, a good confinement device will retain it in the interior of the plasma. Therefore, alpha measurement system need to probe the interior of a high density plasma. Due to the conducting nature of a plasma, wave with frequencies below the plasma frequency can not penetrate into the interior of the plasma where the alphas reside. This project uses a wave that can interact with the perpendicular motion of the alphas to probe its characteristics. However, this wave (the lower hybrid wave) is below the plasma frequency and can not be directly launched from the plasma edge. This project was designed to non-linearly excite the lower hybrid in the interior of a magnetized plasma and measure its interaction with a fast ion population.

  12. 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.

  13. 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

  14. 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.

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

    NASA Astrophysics Data System (ADS)

    Hübner, M.; Guaitella, O.; Rousseau, A.; Röpcke, J.

    2013-07-01

    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, CO2, O3, NO2, N2O, HCN, H2O, HNO3, CH2O, and CH2O2 has been monitored. The concentrations of the by-products range between 5 ppm, in the case of NO2, and 1200 ppm, for H2O. 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 C2H4 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-1. The value of the characteristic energy β, characterizing the energy efficiency of the ethylene destruction in the reactor, was found to be 330 J l-1. It was proven that the application of three reactor stages suppresses essentially the production of harmful by-products as formaldehyde, formic acid, and NO2 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 (RF(C2H4) = 5.5 × 1019 molecules J-1) and conversion efficiencies to the produced molecular species (RC = (0.1-3) × 1016 molecules J-1) on the discharge conditions could be estimated in the multistage plasma reactor.

  16. 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.

  17. 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.

  18. 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)

  19. Generation, utilization, and transformation of cathode electrons for bioreduction of Fe(III)EDTA in a biofilm electrode reactor related to NOx removal from flue gas.

    PubMed

    Li, Wei; Xia, Yinfeng; Zhao, Jingkai; Liu, Nan; Li, Sujing; Zhang, Shihan

    2015-04-01

    A chemical absorption-biological reduction (CABR) integrated system, which employs iron chelate as a solvent, is under development for NOx removal from flue gas. Biofilm electrode reactor (BER) is deemed as a promising bioreactor to regenerate the iron chelate. Although it has been proved that BER can significantly enhance the bioreduction of Fe(III)EDTA, the bioelectrochemistry mechanism involved in the bioreduction of Fe(III)EDTA remains unknown. This work aims to explore this mechanism via the analysis of the generation, utilization, and transformation of cathode electrons in the BER. The results indicate that the generation of cathode electrons follows Faraday's law. The generated cathode electrons were used to produce H2 and directly reduce Fe(III)EDTA in the BER. Meanwhile, the produced H2 served as an electron donor for bioreduction of Fe(III)EDTA. The excess H2 product was transformed to simple organics, e.g., methanol by the hydrogen autotrophy of Pseudomonas under the inorganic and anaerobic conditions. Overall, this work revealed that the reduction of Fe(III)EDTA in the BER was enhanced by both direct electrochemical reduction and indirect bioreduction using H2 as an intermediate. It is also interesting that the excess H2 product was transformed to methanol for microbial metabolism and energy storage in the BER. PMID:25799265

  20. 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.

  1. Missing aerobic-phase nitrogen: The potential for heterotrophic reduction of autotrophically generated nitrous oxide in a sequencing batch reactor wastewater treatment system.

    PubMed

    Shiskowskii, D M; Mavinic, D S

    2005-08-01

    Several biochemical pathways can induce nitrogen loss from aerated, aerobic wastewater treatment bioreactors. These pathways include "traditional" simultaneous nitrification-denitrification (SND) (i.e. autotrophic nitrification - heterotrophic denitrification), autotrophic denitrification, and anaerobic ammonia oxidation. An oxygen limitation, often expressed in terms of low dissolved oxygen (DO) concentration, is a common element of these pathways. The presented research investigated the effect of mixed liquor DO concentration and biomass slowly degradable carbon (SDC) utilization rate on the heterotrophic nitrous oxide (N2O) reduction rate, for biomass cultured in an anoxic/aerobic wastewater treatment bioreactor. Biomass oxygen and SDC availability-limitation, expressed in terms of DO concentration and SDC ultilization rate, respectively, were found to significantly impact the observed heterotrophic N2O reduction rate. The findings support the hypothesis that nitrogen lost from the mixed liquor of an aerobic bioreactor could result from simultaneous autotrophic N2O generation (i.e. autotrophic denitrification) and heterotrophic N2O reduction. The results also support the idea that autotrophic N2O generation could be occurring in a bioreactor, although N2O may not be measurable in the reactor off-gas. Therefore, this autotrophic N2O generation - heterotrophic N2O reduction mechanism provides an alternative explanation to nitrogen loss, when compared to "conventional" SND, where heterotrophic organisms are assumed to reduce autotrophically generated nitrite and nitrate to dinitrogen (N2). In addition, nitrogen loss speculatively attributed to N2 formation via anaerobic ammonia oxidation in oxygen-limited environments, again because of the absence of measurable N2O, may in fact be due to the autotrophic N2O generation - heterotrophic N2O reduction mechanism. PMID:16128383

  2. 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.

  3. MYRRHA a multi-purpose hybrid research reactor for high-tech applications

    SciTech Connect

    Abderrahim, H. A.; Baeten, P.

    2012-07-01

    MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is the flexible experimental accelerator driven system (ADS) in development at SCK-CEN. MYRRHA is able to work both in subcritical (ADS) as in critical mode. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material developments for generation IV (GEN IV) systems, material developments for fusion reactors, radioisotope production and industrial applications, such as Si-doping. MYRRHA will also demonstrate the ADS full concept by coupling the three components (accelerator, spallation target and subcritical reactor) at reasonable power level to allow operation feedback, scalable to an industrial demonstrator and allow the study of efficient transmutation of high-level nuclear waste. MYRRHA is based on the heavy liquid metal technology and so it will contribute to the development of lead fast reactor (LFR) technology and in critical mode, MYRRHA will play the role of European technology pilot plant in the roadmap for LFR. In this paper the historical evolution of MYRRHA and the rationale behind the design choices is presented and the latest configuration of the reactor core and primary system is described. (authors)

  4. 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.

  5. 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.

  6. 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

  7. 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…

  8. 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.

  9. 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…

  10. 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.

  11. Synthesis and Characterization of Pt(IV) Fluorescein Conjugates to Investigate Pt(IV) Intracellular Transformations

    PubMed Central

    Song, Ying; Suntharalingam, Kogularamanan; Yeung, Jessica S.; Royzen, Maksim; Lippard, Stephen J.

    2013-01-01

    Pt(IV) anticancer compounds typically operate as prodrugs that are reduced in the hypoxic environment of cancer cells, losing two axial ligands in the process to generate active Pt(II) species. Here we report the synthesis of two fluorescent Pt(IV) prodrugs of cisplatin in order to image and evaluate the Pt(IV) reduction process in simulated and real biological environments. Treatment of the complexes dissolved in PBS buffer with reducing agents typically encountered in cells, glutathione or ascorbate, afforded a 3- to 5-fold fluorescence turn-on owing to reduction and loss of their fluorescein-based axial ligands, which are quenched when bound to platinum. Both Pt(IV) conjugates displayed moderate cytotoxicity against human cancer cell lines, with IC50 values higher than that of cisplatin. Immunoblotting and DNA flow cytometry analyses of one of the complexes, Pt(IV)FL2, revealed that it damages DNA, causing cell cycle arrest in S or G2/M depending on exposure time, with ultimately triggering of apoptotic cell death. Fluorescence microscopic studies prove that Pt(IV)FL2 enters cells intact and undergoes reduction intracellularly. The results are best interpreted in terms of a model in which the axial fluorescein ligands are expelled through lysosomes, with the platinum(II) moiety generated in the process binding to genomic DNA, which results in cell death. PMID:23957697

  12. 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

  13. Evaluation of the use of color-set geometry during lattice physics constants generation for boiling water reactor simulation

    SciTech Connect

    Evans, S.; Ivanov, K.

    2013-07-01

    Current methods for BWR nuclear design and analysis consist of using lattice physics neutron transport methods to generate the two-group homogenized cross-sections that are then used in a nodal diffusion theory code. The lattice transport solutions are performed for a single assembly with reflective boundary conditions, which is a practical approximation. A method is developed to account for assembly exposure distributions (environment) in the core within the lattice transport calculations with the use of color-sets (2x2) geometry. The loading pattern is examined and an appropriate number of characteristic color-set cells are selected for analysis. Treatment of the co-resident exposed fuel within this method is also presented. The calculation process was followed for a recent BWR cycle design with comparisons being performed on both a lattice and core-wide basis to evaluate the proposed method. The lattice based comparisons show noticeable differences in the pin power distribution predictions, which require further investigation to see how this translates into core performance calculations. The core-wide comparisons show minor differences and are generally in a good agreement, which is expected with this small perturbation. A slight improvement was noticed in the reduction of the power distribution uncertainty. However, given the additional amount of work and computer run time increase, further evaluation, especially of core pin power predictions, is needed to consider this method for production level design and safety analysis calculations. (authors)

  14. Criticality and Safety Parameter Studies of a 3-MW TRIGA MARK-II Research Reactor and Validation of the Generated Cross-Section Library and Computational Method

    SciTech Connect

    Bhuiyan, S.I.; Mondal, M.A.W.; Sarker, M.M.; Rahman, M.; Shahdatullah, M.S.; Huda, M.Q.; Chakrobortty, T.K.; Khan, M.J.H

    2000-05-15

    This study deals with the analysis of some neutronics and safety parameters of the current core of a 3-MW TRIGA MARK-II research reactor and validation of the generated macroscopic cross-section library and calculational techniques by benchmarking with experimental, operational, and available Safety Analysis Report (SAR) values. The overall strategy is: (a) generation of the problem-dependent cross-section library from basic Evaluated Nuclear Data Files such as ENDF/B-VI and JENDL-3.2 with NJOY94.10+, (b) use of the WIMSD-5 package to generate a few-group neutron macroscopic cross section for all of the materials in the core and its immediate neighborhood, (c) use the three-dimensional CITATION code to perform the global analysis of the core, and (d) checking of the validity of the CITATION diffusion code with the MCNP4B2 Monte Carlo code. The ultimate objective is to establish methods for reshuffling the current core configuration to upgrade the thermal flux at irradiation locations for increased isotope production. The computational methods, tools and techniques, customization of cross-section libraries, various models for cells and supercells, and many associated utilities are standardized and established/validated for the overall neutronic analysis. The excess reactivity, neutron flux, power distribution, power peaking factors, determination of the hot spot, and fuel temperature reactivity coefficients {alpha}{sub f} in the temperature range of 45 to 1000 deg. C are studied. All the analyses are performed using the 4- and 7-group libraries of the macroscopic cross sections generated from the 69-group WIMSD-5 library. The 7-group calculations yield comparatively better agreement with the experimental value of k{sub eff} and the other core parameters. The CITATION test runs using different cross-section sets based on the different models applied in the WIMSD-5 calculations show a strong influence of those models on the final integral parameter. Some of the cells

  15. 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

  16. Analysis of N Reactor anticipated transients without scram

    SciTech Connect

    Sathyanarayana, K. ); Ogden, D.M. )

    1989-07-01

    This report documents the N Reactor thermal-hydraulic system response for the anticipated transients without SCRAM (ATWS). The N Reactor is a graphite-moderated, water-cooled pressure-tube reactor. The primary coolant is pumped through 1003 horizontal pressure-tubes which contain two concentric tubular metallic fuel elements. The calculations were performed using RELAP5/MOD2 computer code. RELAP5/MOD2 is a best-estimate, two-phase, full six-equation, nonequilibrium, and nonhomogeneous thermal-hydraulic computer code designed for pressurized water reactor transients. A point-reactor kinetics model with reactivity feedback was employed to compute the power generated in the reactor core. The transients simulated were 200% double-ended guillotine break in the cold-leg manifold and the station blackout with the simultaneous failure of SCRAM and emergency core cooling system (ECCS) and the loss of ac power. In all the transients, the loss of high pressure injection system (HPI), the complete loss of graphite and shield cooling system (GSCS) and the isolation of steam generator secondary were included. Also, the ATWS transient, due to the 200% guillotine break of the cold-leg manifold assuming the availability of ECCS from three high-lift diesel pumps operating at full capacity and the loss of ac power, was analyzed. These three transients were first evaluated using the void feedback reactivity having an absolute level of {minus}167 mk corresponding to an infinite lattice MK IV fuel. Then these transients were reevaluated using a more realistic void feedback reactivity having an absolute level of {minus}70 mk. The report includes brief descriptions of the N Reactor, the computer code, the thermal-hydraulic plant model used in the simulations and the discussion of results. 12 refs., 113 figs., 13 tabs.

  17. Gas Core Reactor with Magnetohydrodynamic Power System and Cascading Power Cycle

    SciTech Connect

    Smith, Blair M.; Anghaie, Samim

    2004-03-15

    The U.S. Department of Energy initiative Generation IV aim is to produce an entire nuclear energy production system with next-generation features for certification before 2030. A Generation IV-capable system must have superior sustainability, safety and reliability, and economic cost advantages in comparison with third generation light water reactors (LWRs). A gas core reactor (GCR) with magnetohydrodynamic (MHD) power converter and cascading power cycle forms the basis for a Generation IV concept that is expected to set the upper performance limits in sustainability and power conversion efficiency among all existing and proposed fission powered systems. A gaseous core reactor delivering thousands of megawatt fission power acts as the heat source for a high-temperature MHD power converter. A uranium tetrafluoride fuel mix, with {approx}95% mol fraction helium gas, provides a stable working fluid for the primary MHD Brayton cycle. The hot working fluid exiting a topping cycle MHD generator has sufficient heat to drive a conventional helium Brayton cycle with 35% thermal efficiency as well as a superheated steam Rankine cycle, with up to 40% efficiency, which recovers the waste heat from the intermediate Brayton cycle. A combined cycle efficiency of close to 70% can be achieved with only a modest MHD topping cycle efficiency. The high-temperature direct-energy conversion capability of an MHD dynamo combined with an already sophisticated steam-powered turbine industry knowledge base allows the cascading cycle design to achieve breakthrough first-law energy efficiencies previously unheard of in the nuclear power industry. Although simple in concept, the gas core reactor design has not achieved the state of technological maturity that established high-temperature gas-cooled reactors and high-temperature molten salt core reactors have pioneered. However, the GCR-MHD concept has considerable promise; for example, like molten salt reactors the fuel is continuously cycled

  18. 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

  19. Microwave photochemical reactor for the online oxidative decomposition of p-hydroxymercurybenzoate (pHMB)-tagged proteins and their determination by cold vapor generation-atomic fluorescence detection.

    PubMed

    Campanella, Beatrice; Rivera, Jose González; Ferrari, Carlo; Biagi, Simona; Onor, Massimo; D'Ulivo, Alessandro; Bramanti, Emilia

    2013-12-17

    A novel method is presented for the characterization and determination of thiolic proteins. After the labeling with p-hydroxymercurybenzoate, the pHMB-labeled proteins underwent on-line oxidation with a novel microwave (MW)/UV photochemical reactor, followed by cold vapor generation-atomic fluorescence spectrometry (CVG-AFS) detection. The MW/UV process led to the conversion of pHMB to Hg(II) with a yield of 89.0 ± 0.5% without using chemical oxidizing reagents and avoiding the use of toxic carcinogenic compounds. Hg(II) was reduced to Hg(0) in a knotted reaction coil with NaBH4 solution, stripped from the solution by an argon flow and detected. The chromatographic method for labeled thiolic peptides was linear in the 0.2-100 μmol L(-1) range, with a LOD as mercury of 57 nmol L(-1). This system has proven to be a useful interface for liquid chromatography coupled with CVG-AFS in the determination and characterization of thiolic proteins. This method has been applied to the determination of thiolic peptides after tryptic digestion of serum albumins from different species (human, bovine, rat, horse, and sheep). PMID:24266505

  20. Site-Specific, Covalent Immobilization of Dehalogenase ST2570 Catalyzed by Formylglycine-Generating Enzymes and Its Application in Batch and Semi-Continuous Flow Reactors.

    PubMed

    Jian, Hui; Wang, Yingwu; Bai, Yan; Li, Rong; Gao, Renjun

    2016-01-01

    Formylglycine-generating enzymes can selectively recognize and oxidize cysteine residues within the sulfatase sub motif at the terminus of proteins to form aldehyde-bearing formylglycine (FGly) residues, and are normally used in protein labeling. In this study, an aldehyde tag was introduced to proteins using formylglycine-generating enzymes encoded by a reconstructed set of the pET28a plasmid system for enzyme immobilization. The haloacid dehalogenase ST2570 from Sulfolobus tokodaii was used as a model enzyme. The C-terminal aldehyde-tagged ST2570 (ST2570CQ) exhibited significant enzymological properties, such as new free aldehyde groups, a high level of protein expression and improved enzyme activity. SBA-15 has widely been used as an immobilization support for its large surface and excellent thermal and chemical stability. It was functionalized with amino groups by aminopropyltriethoxysilane. The C-terminal aldehyde-tagged ST2570 was immobilized to SBA-15 by covalent binding. The site-specific immobilization of ST2570 avoided the chemical denaturation that occurs in general covalent immobilization and resulted in better fastening compared to physical adsorption. The site-specific immobilized ST2570 showed 3-fold higher thermal stability, 1.2-fold higher catalytic ability and improved operational stability than free ST2570. The site-specific immobilized ST2570 retained 60% of its original activity after seven cycles of batch operation, and it was superior to the ST2570 immobilized to SBA-15 by physical adsorption, which loses 40% of its original activity when used for the second time. It is remarkable that the site-specific immobilized ST2570 still retained 100% of its original activity after 10 cycles of reuse in the semi-continuous flow reactor. Overall, these results provide support for the industrial-scale production and application of site-specific, covalently immobilized ST2570. PMID:27409601

  1. 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.

  2. 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

  3. Higher temperature reactor materials workshop sponsored by the Department of Energy Office of Nuclear Energy, Science, and Technology (NE) and the Office of Basic Energy Sciences (BES).

    SciTech Connect

    Allen, T.; Bruemmer, S.; Kassner, M.; Odette, R.; Stoller, R.; Was, G.; Wolfer, W.; Zinkle, S.; Elmer, J.; Motta, A.

    2002-08-12

    On March 18-21, 2002, the Department of Energy, Office of Nuclear Energy, Science, and Technology (NE) and the Office of Basic Energy Sciences (BES) sponsored a workshop to identify needs and opportunities for materials research aimed at performance improvements of structural materials in higher temperature reactors. The workshop focused discussion around the reactor concepts proposed as part of the Generation IV Nuclear Energy System Roadmap. The goal of the Generation IV initiative is to make revolutionary improvements in nuclear energy system design in the areas of sustainability, economics, safety and reliability. The Generation IV Nuclear Energy Systems Roadmap working groups have identified operation at higher temperature as an important step in improving economic performance and providing a means for nuclear energy to support thermochemical production of hydrogen. However, the move to higher operating temperatures will require the development and qualification of advanced materials to perform in the more challenging environment. As part of the process of developing advanced materials for these reactor concepts, a fundamental understanding of materials behavior must be established and the data-base defining critical performance limitations of these materials under irradiation must be developed. This workshop reviewed potential reactor designs and operating regimes, potential materials for application in high-temperature reactor environments, anticipated degradation mechanisms, and research necessary to understand and develop reactor materials capable of satisfactory performance while subject to irradiation damage at high temperature. The workshop brought together experts from the reactor materials and fundamental materials science communities to identify research and development needs and opportunities to provide optimum high temperature nuclear energy system structural materials.

  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. 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

  6. 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

  7. A Mn(IV)/Fe(IV) Intermediate in Assembly of the Mn(IV)/Fe(III) Cofactor of Chlamydia trachomatis Ribonucleotide Reductase†

    PubMed Central

    Jiang, Wei; Hoffart, Lee M.; Krebs, Carsten; Bollinger, J. Martin

    2008-01-01

    We recently showed that the class Ic ribonucleotide reductase from the human pathogen, Chlamydia trachomatis, uses a MnIV/FeIII cofactor to generate protein and substrate radicals in its catalytic mechanism [Jiang, W., Yun, D., Saleh, L., Barr, E. W., Xing, G., Hoffart, L. M., Maslak, M.-A., Krebs, C., and Bollinger, J. M., Jr. (2007) Science 316, 1188-1191]. Here, we have dissected the mechanism of formation of this novel heterobinuclear redox cofactor from the MnII/FeII cluster and O2. An intermediate with a g = 2 EPR signal that shows hyperfine coupling to both 55Mn and 57Fe accumulates almost quantitatively in a second order reaction between O2 and the reduced R2 complex. The otherwise slow decay of the intermediate to the active MnIV/FeIII-R2 complex is accelerated by the presence of the one-electron reductant, ascorbate, implying that the intermediate is more oxidized than MnIV/FeIII. Mössbauer spectra show that the intermediate contains a high-spin FeIV center. Its chemical and spectroscopic properties establish that the intermediate is a MnIV/FeIV-R2 complex with an S = 1/2 electronic ground state arising from antiferromagnetic coupling between the MnIV (SMn = 3/2) and high-spin FeIV (SFe = 2) sites. PMID:17616152

  8. Initial Requirements for Gas-Cooled Fast Reactor (GFR) System Design, Performance, and Safety Analysis Models

    SciTech Connect

    Kevan D. Weaver; Thomas Y. C. Wei

    2004-08-01

    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.

  9. 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.

  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. IVS contribution to ITRF2014

    NASA Astrophysics Data System (ADS)

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

    2016-04-01

    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

  12. Optimized Battery-Type Reactor Primary System Design Utilizing Lead

    SciTech Connect

    Yu, Yong H.; Son, Hyoung M.; Lee, Il S.; Suh, Kune Y.

    2006-07-01

    A number of small and medium size reactors are being developed worldwide as well as large electricity generation reactors for co-generation, district heating or desalination. The Seoul National University has started to develop 23 MWth BORIS (Battery Optimized Reactor Integral System) as a multi-purpose reactor. BORIS is an integral-type optimized fast reactor with an ultra long life core. BORIS is being designed to meet the Generation IV nuclear energy system goals of sustainability, safety, reliability and economics. Major features of BORIS include 20 consecutive years of operation without refueling; elimination of an intermediate heat transport loop and main coolant pump; open core without individual subassemblies; inherent negative reactivity feedback; and inherent load following capability. Its one mission is to provide incremental electricity generation to match the needs of developing nations and especially remote communities without major electrical grid connections. BORIS consists of a reactor module, heat exchanger, coolant module, guard vessel, reactor vessel auxiliary cooling system (RVACS), secondary system, containment and the seismic isolation. BORIS is designed to generate 10 MWe with the resulting thermal efficiency of 45 %. BORIS uses lead as the primary system coolant because of the inherent safety of the material. BORIS is coupled with a supercritical carbon dioxide Brayton cycle as the secondary system to gain a high cycle efficiency in the range of 45 %. The reference core consists of 757 fuel rods without assembly with an active core height of 0.8 m. The BORIS core consists of single enrichment zone composed of a Pu-MA (minor actinides)-U-N fuel and a ferritic-martensitic stainless steel clad. This study is intended to set up appropriate reactor vessel geometry by performing thermal hydraulic analysis on RVACS using computational fluid dynamics codes; to examine the liquid metal coolant behavior along the subchannels; to find out whether the

  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. IVS contribution to the next ITRF

    NASA Astrophysics Data System (ADS)

    Bachmann, Sabine; Messerschmitt, Linda; Thaller, Daniela

    2015-04-01

    Generating the contribution of the International VLBI Service (IVS) to the next ITRF (ITRF2013 or ITRF2014) was the main task of the IVS Combination Center at the Federal Agency for Cartography and Geodesy (BKG, Germany) in 2014. Starting with the ITRF2005, the IVS contribution to the ITRF is an intra-technique combined solution using multiple individual contributions from different institutions. For the upcoming ITRF ten international institutions submitted data files for a combined solution. The data files contain 24h VLBI sessions from the late 1970s until the end of 2014 in SINEX file format containing datum free normal equations with station coordinates and Earth Orientation Parameters (EOP). All contributions have to meet the IVS standards for ITRF contribution in order to guarantee a consistent combined solution. In the course of the generation of the intra-technique combined solution, station coordinate time series for each station as well as a Terrestrial Reference Frame based on the contributed VLBI data (VTRF) were generated and analyzed. Preliminary results using data until the end of 2013 show a scaling factor of -0.47 ppb resulting from a 7-parameter Helmert transformation of the VTRF w.r.t. ITRF2008, which is comparable to the scaling factor that was determined in the precedent ITRF generation. An internal comparison of the EOPs between the combined solution and the individual contributions as well as external comparisons of the EOP series were carried out in order to assure a consistent quality of the EOPs. The data analyses, the combination procedure and results of the combined solution for station coordinates and EOP will be presented.

  15. 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.

  16. 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)

  17. 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)

  18. The Illiac IV

    SciTech Connect

    Hord, M.

    1982-01-01

    The story of the Illiac IV is also in part the story of the Institute for Advanced Computation. This is the government organization formed in 1971 by the Defense Advanced Research Projects Agency and the National Aeronautics and Space Administration Ames Research Center to develop and operate this computer. The Institute provides access to the Illiac through a connection to the ARPANET, a national communication network. The Institute also performs software development, maintenance, and research in various advanced computation topics. Considerable effort has been invested by the Institute in documenting the evolution of the Illiac system and providing those publications to the user community. This material has experienced quite limited circulation and to most of the computer world the Illiac remains mysterious. This attitude is fostered by the lack of a thorough summary of the Illiac's environment, design and capabilities. It is in response to that information gap that this book is addressed.

  19. Status on R and D Planning for Supercritical Water Cooled Reactor Systems in the 6. European Framework Programme

    SciTech Connect

    Starflinger, J.; Schulenberg, T.; Aksan, N.; Bittermann, D.; Heikinheimo, L.; Rimpault, G.

    2004-07-01

    On July 30, 2003, EURATOM signed the charter to join the Generation IV International Forum and thus to contribute to innovative reactor design and development. Among other concepts, supercritical water cooled reactor systems shall be foreseen as their contribution. In order to support this international forum, a dedicated budget for R and D of innovative concepts is planned for the 6. European Framework Programme of the European Commission. Currently, a detailed work plan for supercritical water cooled reactor systems is being worked out, in order to be presented and decided by the European Commission thereafter. It shall include: - Design studies of a thermal reactor core, its reactor pressure vessel internals and of the balance of plant. - Study of a fast reactor option for sustainable use of fuel and for Plutonium management. - Study of the corrosion behavior and other performances of candidate materials at supercritical pressures. - Detailed investigations of heat transfer and pressure drop at supercritical pressures and at part load operation conditions of the reactor. - Design code improvements and verifications. - Conceptual design and analyses of a suitable safety system. As a result, the program shall enable a thorough assessment of the supercritical water cooled reactor system with a view to determine its future potential. This summary report shall give an overview of the contributions, which are planned to be provided by the EURATOM partners in France, Finland, Germany, Hungary, Switzerland, The Netherlands and The Czech Republic. (authors)

  20. 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.

  1. 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.

  2. Microstructural analysis of candidate steels pre-selected for new advanced reactor systems

    NASA Astrophysics Data System (ADS)

    Zeman, A.; Debarberis, L.; Kočík, J.; Slugeň, V.; Keilová, E.

    2007-05-01

    Many of candidate materials for Generation IV concept have been developed in past, however they have not been tested yet for new conditions and therefore, a huge effort must be carried out for their characterization and validation in the R&D phase. Moreover, the issue concerning the structural integrity of materials is going to have a more and more dominant role due to the degradation mechanisms, as hardening, thermal fatigue, creep, swelling, phase instability and embrittlement, which must be understood in order to estimate the component lifetime. This paper presents some interesting results on few candidate materials selected for Generation IV reactor concepts. The study is based on application of non-destructive spectroscopic techniques to investigate materials as EUROFER'97, P91 and T91.

  3. 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…

  4. 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)

  5. 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.

  6. Future reactor experiments

    NASA Astrophysics Data System (ADS)

    Wen, Liangjian

    2015-07-01

    The non-zero neutrino mixing angle θ13 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.

  7. 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.

  8. Reactor safety assessment system

    SciTech Connect

    Sebo, D.E.; Bray, M.A.; King, M.A.

    1987-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSA is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category.

  9. 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.

  10. 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, ...

  11. 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

  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. 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.

  14. 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.

  15. RACEWAY REACTOR FOR MICROALGAL BIODIESEL PRODUCTION

    EPA Science Inventory

    The proposed mathematical model incorporating mass transfer, hydraulics, carbonate/aquatic chemistry, biokinetics, biology and reactor design will be calibrated and validated using the data to be generated from the experiments. The practical feasibility of the proposed reactor...

  16. Nuclear Data Measurements for 21st Century Reactor Physics Applications

    SciTech Connect

    Rahmat Aryaeinejad; Jerald D. Cole; Mark W. Drigert; James K. Jewell; Christopher A. McGrath; David W. Nigg; Edward L. Reber

    2003-03-01

    The United States Department of Energy (DOE), Office of Nuclear Energy (NE) has embarked on a long-term program to significantly advance the science and technology of nuclear energy. This is in response to the overall national plan for accelerated development of domestic energy resources on several fronts, punctuated by recent dramatic events that have emphasized the need for the US to reduce its dependence on foreign petroleum supplies. Key aspects of the DOE-NE agenda are embodied in the Generation-IV (Gen-IV) advanced nuclear energy systems development program and in the Advanced Fuel Cycle (AFC) program. The planned efforts involve near-term and intermediate-term improvements in fuel utilization and recycling in current nuclear power reactor systems as well as the longer-term development of new nuclear energy systems that offer much improved fuel utilization and proliferation resistance, along with continued advances in operational safety. The success of the overall NE effort will depend not only on sophisticated system development and engineering, but also on the advances in the supporting sciences and technologies. Of these, one of the most important is the improvement of the relevant fundamental nuclear science data bases, especially the evaluated neutron interaction cross section files that serve as the foundation of all reactor system designs, operating strategies, and fuel cycle engineering activities. The new concepts for reactors and fuel cycles involve the use of transuranic nuclides that were previously of little interest, and where experimentally measured information is lacking. The current state of the cross section database for some of these nuclides is such that design computations for advanced fast-spectrum reactor systems and fuel cycles that incorporate such materials in significant quantities are meaningful only for approximate conceptual applications. No actual system could reliably be designed according to currently accepted standards, nor

  17. 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)

  18. Fission energy: The integral fast reactor

    SciTech Connect

    Chang, Yoon I.

    1989-01-01

    The Integral Fast Reactor (IFR) is an innovative reactor concept being developed at Argonne National Laboratory as a such next- generation reactor concept. The IFR concept has a number of specific technical advantages that collectively address the potential difficulties facing the expansion of nuclear power deployment. In particular, the IFR concept can meet all three fundamental requirements needed in a next-generation reactor as discussed below. This document discusses these requirements.

  19. ARIES-IV Nested Shell Blanket Design

    SciTech Connect

    Wong, C.P.C.; Redler, K.; Reis, E.E.; Will, R.; Cheng, E.; Hasan, C.M.; Sharafat, S.

    1993-11-01

    The ARIES-IV Nested Shell Blanket (NSB) Design is an alternate blanket concept of the ARIES-IV low activation helium-cooled reactor design. The reference design has the coolant routed in the poloidal direction and the inlet and outlet plena are located at the top and bottom of the torus. The NSB design has the high velocity coolant routed in the toroidal direction and the plena are located behind the blanket. This is of significance since the selected structural material is SiC-composite. The NSB is designed to have key high performance components with characteristic dimensions of no larger than 2 m. These components can be brazed to form the blanket module. For the diverter design, we eliminated the use of W as the divertor coating material by relying on the successful development of the gaseous divertor concept. The neutronics and thermal-hydraulic performance of both blanket concepts are similar. The selected blanket and divertor configurations can also meet all the projected structural, neutronics and thermal-hydraulics design limits and requirements. With the selected blanket and divertor materials, the design has a level of safety assurance rate of I (LSA-1), which indicates an inherently safe design.

  20. Advanced In-Service Inspection Approaches Applied to the Phenix Fast Breeder Reactor

    SciTech Connect

    Guidez, J.; Martin, L.; Dupraz, R.

    2006-07-01

    The safety upgrading of the Phenix plant undertaken between 1994 and 1997 involved a vast inspection programme of the reactor, the external storage drum and the secondary sodium circuits in order to meet the requirements of the defence-in-depth safety approach. The three lines of defence were analysed for every safety related component: demonstration of the quality of design and construction, appropriate in-service inspection and controlling the consequences of an accident. The in-service reactor block inspection programme consisted in controlling the core support structures and the high-temperature elements. Despite the fact that limited consideration had been given to inspection constraints during the design stage of the reactor in the 1960's, as compared to more recent reactor projects such as the European Fast Reactor (EFR), all the core support line elements were able to be inspected. The three following main operations are described: Ultrasonic inspection of the upper hangers of the main vessel, using small transducers able to withstand temperatures of 130 deg. C, Inspection of the conical shell supporting the core dia-grid. A specific ultrasonic method and a special implementation technique were used to control the under sodium structure welds, located up to several meters away from the scan surface. Remote inspection of the hot pool structures, particularly the core cover plug after partial sodium drainage of the reactor vessel. Other inspections are also summarized: control of secondary sodium circuit piping, intermediate heat exchangers, primary sodium pumps, steam generator units and external storage drum. The pool type reactor concept, developed in France since the 1960's, presents several favourable safety and operational features. The feedback from the Phenix plant also shows real potential for in-service inspection. The design of future generation IV sodium fast reactors will benefit from the experience acquired from the Phenix plant. (authors)

  1. 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

  2. BOILING REACTORS

    DOEpatents

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  3. 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)

  4. Complete primary structure of the sixth chain of human basement membrane collagen, alpha 6(IV). Isolation of the cDNAs for alpha 6(IV) and comparison with five other type IV collagen chains.

    PubMed

    Zhou, J; Ding, M; Zhao, Z; Reeders, S T

    1994-05-01

    Basement membranes were previously believed to contain five distinct type IV collagen subunits. We have recently isolated part of the cDNA for a novel type IV collagen, alpha 6(IV), and shown that COL4A6, the gene encoding this new chain, is deleted in Alport syndrome-associated leiomyomatosis (Zhou, J., Mochizuki, T., Smeets, H., Antignac, C., Laurila, P., de Paepe, A., Tryggvason, K., and Reeders, S. T. (1993) Science 261, 1167-1169). Here, we describe the entire human alpha 6(IV) cDNA and show that the gene encodes a classical type IV collagen with homology throughout its length to all the other five chains. There is a 21-residue signal peptide, a 1417-residue collagenous domain interrupted at 25 points, and a 228-residue carboxyl-terminal non-collagenous domain. When the complete primary structure of this new chain was compared with all the other known chains, it became clear that alpha 6(IV) has the most resemblance to alpha 2(IV) and alpha 4(IV). The evolution of the six chains was deduced, allowing a new classification of the type IV collagen family. The alpha 6(IV) chain is a candidate gene for X-linked Alport syndrome; knowledge of the complete structure of the chain will permit us to screen systematically for mutations in patients and to generate recombinant proteins and synthetic peptides for further study of cell-matrix interactions involving the alpha 6(IV) chain. PMID:8175748

  5. 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.

  6. 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. PMID:22738207

  7. ApoA-IV promotes the biogenesis of apoA-IV-containing HDL particles with the participation of ABCA1 and LCAT[S

    PubMed Central

    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

  8. 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.

  9. 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)

  10. 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.

  11. 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.

  12. 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.

  13. EVALUATION OF THE PILLS IV

    EPA Science Inventory

    The report gives results of theoretical and experimental investigations of the operating characteristics of the PILLS IV (Particulate Instrumentation by Laser Light Scattering) in situ particle sizing instrument. Results of both investigations show large errors in sizing particle...

  14. 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.

  15. 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.

  16. Testing of Passive Safety System Performance for Higher Power Advanced Reactors

    SciTech Connect

    brian G. Woods; Jose Reyes, Jr.; John Woods; John Groome; Richard Wright

    2004-12-31

    This report describes the results of NERI research on the testing of advanced passive safety performance for the Westinghouse AP1000 design. The objectives of this research were: (a) to assess the AP1000 passive safety system core cooling performance under high decay power conditions for a spectrum of breaks located at a variety of locations, (b) to compare advanced thermal hydraulic computer code predictions to the APEX high decay power test data and (c) to develop new passive safety system concepts that could be used for Generation IV higher power reactors.

  17. J-Integral characterization of the nozzle steels from intermediate test vessels IV-5 and IV-9

    SciTech Connect

    Auten, T.A.; Macdonald, B.D.; Scavone, D.W.; Bozik, D.

    1994-10-01

    Reported here are the results of elastic-plastic fracture toughness tests performed on low alloy steels from the nozzles of the intermediate test vessels IV-5 and IV-9 from the Heavy Steel Section Technology Program at Oak Ridge National Laboratory. These vessels had been given prototypic nozzle corner flaw tests prior to the development of the ASTM E-813 standard test procedure for J-integral testing. The objective of this work is to provide J-integral material test support for future elastic-plastic fracture mechanics analysis of the nozzles. J-integral tests at 88{degrees}C (190{degrees}F) of the IV-5 nozzle material produced stable ductile tearing. The tearing resistance data are expected to support analysis of the observed similar stable tearing response of the nozzle corner flaw. J-integral tests at 24{degrees}C (75{degrees}F) of the IV-9 nozzle produced elastic-plastic fracture instability preceded by stable tearing. A similar response was observed in the IV-9 nozzle corner flaw test. It will be a major and important challenge to develop a fracture mechanics rationale that reconciles these small specimen and nozzle corner flaw test results. These test results are being made available to allow their use by a wide variety of organizations in developing such a rationale, which would be a significant contribution to quantifying the flaw tolerance of reactor pressure vessels.

  18. 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.

  19. Gas Core Reactor-MHD Power System with Cascading Power Cycle

    SciTech Connect

    Smith, Blair M.; Anghaie, Samim; Knight, Travis W.

    2002-07-01

    The US Department of Energy initiative Gen-IV aim is to produce an entire nuclear energy production system with next generation features for certification before 2030. A Generation 4 capable system must have superior sustainability, safety and reliability, and economic cost advantages in comparison with third generation light water reactors. A gas core reactor (GCR) with magnetohydrodynamic (MHD) power converter and cascading power cycle forms the basis for a Generation IV concept that is expected to set the upper performance limits in sustainability and power conversion efficiency among all existing and proposed fission powered systems. A gaseous core reactor delivering 1000's MW fission power acts as the heat source for a high temperature magnetohydrodynamic power converter. A uranium tetrafluoride fuel mix, with {approx}95% mole fraction helium gas, provides a stable working fluid for the primary MHD-Brayton cycle. A helium Brayton cycle extracts waste heat from the MHD generator with about 20% energy efficiency, but the low temperature side is still hot enough ({approx}1600 K) to drive a second conventional helium Brayton cycle with about 35% efficiency. There is enough heat at the low temperature side of the He-Brayton cycle to generate steam, and so another heat recovery cycle can be added, this time a Rankine steam cycle with up to 40% efficiency. The proof of concept does not require a tremendously efficient (first law) MHD cycle, the high temperature direct energy conversion capability of an MHD dynamo, combined with already sophisticated steam powered turbine industry knowledge base allows the cascading cycle design to achieve break-through first law energy efficiencies previously unheard of in the nuclear power industry. Although simple in concept, the gas core reactor design has not achieved the state of technological maturity that, say, molten salt or high-temperature gas-cooled reactors have pioneered. However, even on paper the GCR-MHD concept holds

  20. 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.

  1. Liquid fuel molten salt reactors for thorium utilization

    DOE PAGESBeta

    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

  2. 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

  3. Cross section generation and physics modeling in a feasibility study of the conversion of the high flux isotope reactor core to use low-enriched uranium fuel

    SciTech Connect

    Ellis, R. J.; Gehin, J. C.; Primm Iii, R. T.

    2006-07-01

    A computational study has been initiated at ORNL to examine the feasibility of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. The current study is limited to steady-state, nominal operation and are focused on the determination of the fuel requirements, primarily density, that are required to maintain the performance of the reactor. Reactor physics analyses are reported for a uranium-molybdenum alloy that would be substituted for the current fuel - U{sub 3}O{sub 8} mixed with aluminum. An LEU core design has been obtained and requires an increase in {sup 235}U loading of a factor of 1.9 over the current HEU fuel. These initial results indicate that the conversion from HEU to LEU results in a reduction of the thermal fluxes in the central flux trap region of approximately 9 % and in the outer beryllium reflector region of approximately 15%. Ongoing work is being performed to improve upon this initial design to further minimize the impact of conversion to LEU fuel. (authors)

  4. 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.

  5. 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)

  6. 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.

  7. Electro-regeneration of Ce(IV) in real spent Cr-etching solutions.

    PubMed

    Chen, Te-San; Huang, Kuo-Lin

    2013-11-15

    This paper presents the electro-regeneration of Ce(IV) in real (hazardous) spent thin-film transistor liquid-crystal display (TFT-LCD) Cr-etching solutions. In addition to Ce(III)>Ce(IV) in diffusivity, a quasi-reversible behavior of Ce(III)/Ce(IV) was observed at both boron-doped diamond (BDD) and Pt disk electrodes. The Ce(IV) yield on Pt increased with increasing current density, and the best current efficiency (CE) was obtained at 2A/2.25 cm(2). The performance in terms of Ce(IV) yield and CE of tested anodes was in order BDD>Pt>dimensional stable anode (DSA). At 2A/2.25 cm(2) on Pt and 40 °C for 90 min, the Ce(IV) yield, CE and apparent rate constant (k) for Ce(III) oxidation were 81.4%, 21.8% and 3.17 × 10(-4) s(-1), respectively. With the increase of temperature, the Ce(IV) yield, CE, and k increased (activation energy = 10.7 kJ/mol), but the specific electricity consumption decreased. The Neosepta CMX membrane was more suitable than Nafion-117 and Nafion-212 to be used as the separator of the Ce(IV) regeneration process. The obtained parameters are useful to design divided batch reactors for the Ce(IV) electro-regeneration in real spent Cr-etching solutions. PMID:24140527

  8. 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.

  9. Characterization of Caramel Colour IV.

    PubMed

    Licht, B H; Shaw, K; Smith, C; Mendoza, M; Orr, J; Myers, D V

    1992-05-01

    A large number of commercial Caramel Colour IV samples were characterized in order to assess the uniformity of the class and to provide data to be used in specifications development. Owing to the chemical and physical complexity of caramel colour it was not feasible to perform detailed analysis of all constituents for assessment of uniformity. Instead, selected parameters were evaluated and judgements were made with respect to compositional uniformity based on the similarities of these parameters among the various samples. As Caramel Colour IV is required by the food industry in a range of colour intensities, there must be a range of properties that differ from sample to sample, but that are sufficiently similar for the material to still be considered as part of the Caramel Colour IV class. Fractions as well as whole caramel were analysed using selected spectrophotometric, chromatographic and chemical techniques. Samples were fractionated based on molecular weight and polarity. The data presented here provide evidence for the uniformity in composition of Caramel Colour IV with respect to molecular weight distribution, to nitrogen and sulphur content and their distribution throughout the fractions, to absorbance properties and to specific low molecular weight compounds. Thus, it can be concluded that Caramel Colour IV exhibits compositional uniformity within the range of colour intensity required by the food industry worldwide. PMID:1644377

  10. 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.

  11. Styrene Aziridination by Iron(IV) Nitrides.

    PubMed

    Muñoz, Salvador B; Lee, Wei-Tsung; Dickie, Diane A; Scepaniak, Jeremiah J; Subedi, Deepak; Pink, Maren; Johnson, Michael D; Smith, Jeremy M

    2015-09-01

    Thermolysis of the iron(IV) nitride complex [PhB(tBuIm)3Fe≡N] with styrene leads to formation of the high-spin iron(II) aziridino complex [PhB(tBuIm)3Fe-N(CH2CHPh)]. Similar aziridination occurs with both electron-rich and electron-poor styrenes, while bulky styrenes hinder the reaction. The aziridino complex [PhB(tBuIm)3Fe-N(CH2CHPh)] acts as a nitride synthon, reacting with electron-poor styrenes to generate their corresponding aziridino complexes, that is, aziridine cross-metathesis. Reaction of [PhB(tBuIm)3Fe-N(CH2CHPh)] with Me3SiCl releases the N-functionalized aziridine Me3SiN(CH2CHPh) while simultaneously generating [PhB(tBuIm)3FeCl]. This closes a synthetic cycle for styrene azirdination by a nitride complex. While the less hindered iron(IV) nitride complex [PhB(MesIm)3Fe≡N] reacts with styrenes below room temperature, only bulky styrenes lead to tractable aziridino products. PMID:26179563

  12. 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.

  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. Experimental Substantiation, Testing, and Commissioning of a Novel Separation System for Steam Generators of Nuclear Power Plants with VVER-1000 Reactors

    SciTech Connect

    Nigmatulin, B. I.; Ageev, A. G.; Blinkov, V. N.; Vasil'eva, R. V.; Korol'kov, B. M.; Dragunov, Yu. G.; Trunov, N. B.; Nekrasov, A. V.; Ilyushin, V. F.

    2003-03-15

    The separation system of a PGV-1000 steam generator is modernized on the basis of bench tests of model PGV-1000 steam generators, full-scale tests of steam generators, and computational analysis of the results of these tests. The changes concern the configuration of the submerged plate and replacement of the louver separator by a receiver baffle. These measures increase the marginal evaporative capacity and the permissible range of variation of the water level, decrease the moisture content at the outlet from the steam generator, and improve the conditions for control and repair of its internal surface.

  15. Standard Missile Block IV battery

    SciTech Connect

    Martin, J.

    1996-11-01

    During the 1980`s a trend in automatic primary battery technologies was the replacement of silver-zinc batteries by thermal battery designs. The Standard missile (SM 2) Block IV development is a noteworthy reversal of this trend. The SM2, Block IV battery was originally attempted as a thermal battery with multiple companies attempting to develop a thermal battery design. These attempts resulted in failure to obtain a production thermal battery. A decision to pursue a silver-zinc battery design resulted in the development of a battery to supply the SM 2, Block IV (thermal battery design goal) and also the projected power requirements of the evolving SM 2, Block IVA in a single silver-zinc battery design. Several advancements in silver-zinc battery technology were utilized in this design that improve the producibility and extend the boundaries of silver-zinc batteries.

  16. 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.

  17. 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.

  18. 96. ARAIV. Typical view of ML1 reactor field test operations. ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    96. ARA-IV. Typical view of ML-1 reactor field test operations. Reactor has been hoisted onto truck at door of test assembly building. May 17, 1963. Ineel photo no. 63-3257. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

  19. Dipeptidyl Peptidase IV-Inhibitory Peptides Derived from Silver Carp (Hypophthalmichthys molitrix Val.) Proteins.

    PubMed

    Zhang, Ying; Chen, Ran; Chen, Xiling; Zeng, Zhu; Ma, Huiqin; Chen, Shangwu

    2016-02-01

    The dipeptidyl peptidase IV (DPP-IV)-inhibitory bioactivity of silver carp protein (SCP) hydrolysates were investigated, and their containing efficacious DPP-IV-inhibitory peptides were explored by in silico hydrolysis analysis, peptide separation combined with liquid chromatography-tandem mass spectrometry (LC-MS/MS) identification, and chemical synthesis. SCP hydrolysates generated by six proteases all showed efficient DPP-IV-inhibitory activities, and Neutrase-generated hydrolysates had the greatest DPP-IV inhibition (IC50 of 1.12 mg/mL). In silico Neutrase hydrolysis revealed hundreds of fragments released from myosin, actin, and collagen of SCPs, which include different Pro-motif peptides but only three reported peptidic DPP-IV inhibitors with moderate or weak bioactivity. In addition, three new DPP-IV-inhibitory peptides were identified using LC-MS/MS; in particular, LPIIDI and APGPAGP showed high DPP-IV-inhibitory activity with IC50 of 105.44 and 229.14 μM, respectively, and behaved in competitive/non-competitive mixed-type DPP-IV inhibition mode. The results indicate that the SCP-derived DPP-IV-inhibitory peptides could be potential functional ingredients in the diabetic diet. PMID:26758401

  20. 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.

  1. 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.

  2. Development and study of an acoustic method of steam content measurement at the fuel assembly outlet of a reactor of an atomic heat generating station

    SciTech Connect

    Mel'nikov, V.I.; Khokhlov, V.N.; Usynin, G.B.; Baidakov, A.G.; Trusov, B.A.; Kut'in, L.N.

    1987-01-01

    When considering the applicability of acoustic methods to steam content determination in reactor conditions, several factors must be taken into account. Among these are a low sound conduction of two-phase steam-and-liquid medium, a high level of technological (parasitic) noise that arises from the equipment operation and coolant flow, and the extreme conditions that reduce the reliability of acoustic pickups or even rule out their durable stable functioning. Hence, in case of an NPP, traditional methods of exposure of a tested medium volume to ultrasound as well as passive methods involving detection of acoustic noise that arises from the coolant boiling turned out to be inefficient. Therefore, in recent years new methods of acoustic probing and acoustic impedance have been proposed for resolving the problem. The first one is based on indication of the sound conduction of a small medium volume (a local method). The second method involves determination of ultrasound attenuation in a waveguide that is passed over by the coolant flow (an integrated method). This work presents the results of a study of an acoustic impedance method and reports on the development of measuring equipment based on this method for steam content determination at the FA outlet of an AHGS reactor.

  3. 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.

  4. 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.

  5. 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.

  6. 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)

  7. Erosion-corrosion entrainment of iron-containing compounds as a source of deposits in steam generators used at nuclear power plants equipped with VVER reactors

    NASA Astrophysics Data System (ADS)

    Tomarov, G. V.; Shipkov, A. A.

    2011-03-01

    The main stages and processes through which deposits are generated, migrate, and precipitate in the metal-secondary coolant system of power units at nuclear power plants are analyzed and determined. It is shown that substances produced by the mechanism of general erosion-corrosion are the main source of the ionic-colloid form of iron, which is the main component of deposits in a steam generator. Ways for controlling the formation of deposits in a nuclear power plant's steam generator are proposed together with methods for estimating their efficiency.

  8. 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…

  9. Title IV: Improving Indian Education.

    ERIC Educational Resources Information Center

    Barker, Kipp A.

    The Indian Education Act of 1972, Title IV, has improved Native American education by emphasizing Native American control; it comes after 400 years of Euro-American involvement in Indian education during which assimilation was the primary goal. In 1568 Jesuit priests began "civilizing" and Christianizing the "savage" Indians; in 1794 the first…

  10. Development of a methodology for the assessment of shallow-flaw fracture in nuclear reactor pressure vessels: Generation of biaxial shallow-flaw fracture toughness data

    SciTech Connect

    McAfee, W.J.; Bass, B.R.; Bryson, J.W.

    1998-07-01

    A technology to determine shallow-flaw fracture toughness of reactor pressure vessel (RPV) steels is being developed for application to the safety assessment of RPVs containing postulated shallow-surface flaws. Shallow-flaw fracture toughness of RPV material has been shown to be higher than that for deep flaws, because of the relaxation of crack-tip constraint. This report describes the preliminary test results for a series of cruciform specimens with a uniform depth surface flaw. These specimens are all of the same size with the same depth flaw. Temperature and biaxial load ratio are the independent variables. These tests demonstrated that biaxial loading could have a pronounced effect on shallow-flaw fracture toughness in the lower transition temperature region for RPV materials. Through that temperature range, the effect of full biaxial (1:1) loading on uniaxial, shallow-flaw toughness varied from no effect near the lower shelf to a reduction of approximately 58% at higher temperatures.

  11. Pressing problems of managing the service life of tube bundles used in steam generators at nuclear power stations equipped with VVER reactors

    NASA Astrophysics Data System (ADS)

    Trunov, N. B.; Popadchuk, V. S.; Davidenko, S. E.; Zhukov, R. Yu.

    2010-05-01

    Optimal approaches for monitoring the state and blanking of the tube bundles of horizontal tube generators are considered, and pressing problems associated with managing their service life are discussed.

  12. Solar photocatalytic oxidation of pretreated wastewaters: laboratory scale generation of design data for technical-scale double-skin sheet reactors.

    PubMed

    Gulyas, H; Jain, H B; Susanto, A L; Malekpur, M; Harasiuk, K; Krawczyk, I; Choromanski, P; Furmanska, M

    2005-05-01

    Batchwise heterogeneous photocatalytic oxidation of model wastewater (solutions of the azo dye "Acid Orange 7" in tap water) has been performed in a laboratory-scale stirred vessel reactor with non-submerged UV-A lamps using titanium dioxide "P25" as photocatalyst. Comparison to results of solar pilot-scale Plexiglass double-skin sheet reactor (DSSR) experiments indicates that the lab-scale method may predict area demand for technical-scale DSSR design. Characteristic UV-A fluences leading to TOC or COD reduction to e(-1) of the initial concentrations were determined in lab-scale stirred vessel experiments for treated effluents of seven different industrial branches, secondary municipal effluent and biologically treated greywater. Predicted areas for solar photocatalytic oxidation of these effluents in DSSRs yielding mineralization of 95% of organics in 100 m3 of the respective effluents for a TiO2 concentration of 2 g l(-1) and a sky and solar radiation of 3.9kWh m(-2) d(-1) within one day greatly varied from below 6,000 m2 (biologically treated lubricating oil refinery effluent) to more than 100,000 m2 (highly saline biologically treated effluent of chemical industry). Especially municipal and refinery effluents (except oil reclaiming) have been identified as promising candidates for reuse after solar photocatalytic oxidation. Mineralization efficiency was decreasing with increasing alkalinity of effluents. This was interpreted by competition of hydrogen carbonate anions with organics for binding sites on photocatalyst surface and by OH radical scavenging by hydrogen carbonate. Dependence on alkalinity was superimposed by salinity influence as some effluents with high alkalinity also exhibited high salt concentrations (especially chloride). PMID:15974268

  13. 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.

  14. 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

  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. 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.

  18. 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.

  19. Evaluation of the PLATO IV Computer-based Education System in the Community College. Final Report.

    ERIC Educational Resources Information Center

    Murphy, Richard T.; Appel, Lola Rhea

    PLATO IV (Programmed Logic for Automatic Teaching Operations) is the fourth generation of a computer assisted instructional system developed at the University of Illinois. The use of PLATO IV at five community colleges, and an evaluation of its educational impact on participating students, instructors, and colleges are described. The PLATO system…

  20. Solid tags for identifying failed reactor components

    DOEpatents

    Bunch, Wilbur L.; Schenter, Robert E.

    1987-01-01

    A solid tag material which generates stable detectable, identifiable, and measurable isotopic gases on exposure to a neutron flux to be placed in a nuclear reactor component, particularly a fuel element, in order to identify the reactor component in event of its failure. Several tag materials consisting of salts which generate a multiplicity of gaseous isotopes in predetermined ratios are used to identify different reactor components.

  1. 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

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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)

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. A comparison of the RELAP5/MOD3 and PARET/ANL codes with the experimental transient data from the SPERT-IV D-12/25 series.

    SciTech Connect

    Woodruff, W. L.

    1998-01-16

    The results from the RELAP5/MOD3 and PARET/ANL codes are compared with the SPERT-IV series of experimental reactivity insertion transients. The PARET/ANL code provides conservative estimates of SPERT-IV experimental data for the midrange transients and for the more severe transients. The PARET results are similar to the results obtained earlier for the SPERT-I D-12/25 series of experiments. The RELAP5/MOD3 code (including the developmental version 3.2.1.2) gives results comparable to PARET for some midrange transients, but seriously diverges from the experimental data when significant boiling is present. Based on the results of this study, the use of the RELAP5 code for research reactor applications should be limited to transients that do not generate substantial boiling and voids. We hope to be able to resolve these differences in further work with the NRC staff and its contractors. The RELAP5 code would be a more useful tool for the analyses research reactor transients with the addition of suitable correlations for low pressures and plate type geometry.

  13. 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)

  14. 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.

  15. 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

  16. 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

  17. 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.

  18. 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. PMID:19043740

  19. Excitation of heliumlike B IV

    NASA Astrophysics Data System (ADS)

    Kolk, K.-H.; Koenig, R.; Kunze, H.-J.

    1986-01-01

    Measurements are reported for line-intensity ratios emitted from heliumlike B IV ions in a plasma with electron density, N(e) = 1.5 x 10 to the 16th/cu cm and electron temperature, kT(e) = 175 eV. The plasma was produced in a theta-pinch discharge. The analysis supports theoretical excitation rates calculated in the distorted-wave approximation, which include the effect of resonances.

  20. 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.

  1. Characterization of a Grape Class IV Chitinase

    PubMed Central

    2015-01-01

    A chitinase was purified from Vitis vinifera Manzoni Bianco grape juice and characterized. On the basis of proteomic analysis of tryptic peptides, a significant match identified the enzyme as a type IV grape chitinase previously found in juices of other V. vinifera varieties. The optimal pH and temperature for activity toward colloidal chitin were found to be 6 and 30 °C, respectively. The enzyme was found to hydrolyze chitin and oligomers of N-acetylglucosamine, generating N,N′-diacetylchitobiose and N-acetylglucosamine as products, but was inactive toward N,N′-diacetylchitobiose. The enzyme exhibited both endo- and exochitinase activities. Because yeast contains a small amount of chitin in the cell wall, the possibility of growth inhibition was tested. At a concentration and pH expected in ripe grapes, no inhibition of wine yeast growth by the chitinase was observed. PMID:24845689

  2. 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.

  3. Biological nutrient removal with low nitrous oxide generation by cancelling the anaerobic phase and extending the idle phase in a sequencing batch reactor.

    PubMed

    Chen, Yinguang; Wang, Dongbo; Zheng, Xiong; Li, Xiang; Feng, Leiyu; Chen, Hong

    2014-08-01

    Although wastewater biological nutrient removal can be achieved by alternating the anaerobic-oxic-anoxic phases, significant amount of nitrous oxide (N2O) is generated in oxic phases, where ammonia-oxidizing bacteria (AOB) rather than heterotrophic denitrifiers are the main contributors. Here a new efficient strategy to remarkably reduce N2O generation was reported. It was found that by cancelling the anaerobic phase and extending the idle phase the N2O generation was reduced by 42% using synthetic wastewater, whereas the total nitrogen and phosphorus removals were unaffected. The mechanistic investigations revealed that the cancelling of anaerobic phase benefited heterotrophic denitrifiers instead of AOB to be responsible for nitrogen removal in the oxic phases, increased the ratio of total nitrogen removal driven by external carbon source, and decreased nitrite accumulation. Quantitative real-time polymerase chain reaction and fluorescence in situ hybridization analyses further showed that the new strategy increased the number of N2O reducing bacteria but decreased the abundance of glycogen accumulating organisms, with N2O as their primary denitrification product. It was also determined that the ratio of nitric oxide reductase activity to N2O reductase activity was significantly decreased after anaerobic phase was cancelled. All these observations were in accord with the reduction of N2O production. The feasibility of this strategy to minimize the generation of N2O was finally confirmed for a real municipal wastewater. The results reported in this paper provide a new viewpoint to reduce N2O generation from wastewater biological nutrient removal. PMID:24873707

  4. Endonuclease IV of Escherichia coli is induced by paraquat

    SciTech Connect

    Chan, E.; Weiss, B.

    1987-05-01

    The addition of paraquat (methyl viologen) to a growing culture of Escherichia coli K-12 led within 1 hr to a 10- to 20-fold increase in the level of endonuclease IV, a DNase for apurinic/apyrimidinic sites. The induction was blocked by chloramphenicol. Increases of 3-fold or more were also seen with plumbagin, menadione, and phenazine methosulfate. H/sub 2/O/sub 2/ produced no more than a 2-fold increase in endonuclease IV activity. The following agents had no significant effect: streptonigrin, nitrofurantoin, tert-butyl hydroperoxide, ..gamma.. rays, 260-nm UV radiation, methyl methanesulfonate, mitomycin C, and ascorbate. Paraquat, plumbagin, menadione, and phenazine methosulfate are known to generate superoxide radical anions via redox cycling in vivo. A mutant lacking superoxide dismutase was unusually sensitive to induction by paraquat. In addition, endonuclease IV could be induced by merely growing the mutant in pure O/sub 2/. The levels of endonuclease IV in uninduced or paraquat-treated cells were unaffected by mutations of oxyR, a H/sub 2/O/sub 2/-inducible gene that governs an oxidative-stress regulon. The results indicate that endonuclease IV is an inducible DNA-repair enzyme and that its induction can be mediated via the production of superoxide radicals.

  5. 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.

  6. Self isolating high frequency saturable reactor

    DOEpatents

    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.

  7. 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.

  8. Review of Reactor Neutrino Experiments

    NASA Astrophysics Data System (ADS)

    Kim, Soo-Bong

    New generation of reactor neutrino experiments, Daya Bay and RENO, have made definitive measurements of the smallest neutrino mixing angle θ13 in 2012, based on the disappearance of electron antineutrinos. More precise measurements of the mixing angle and reactor neutrino spectra have been made and presented. A rather large value of θ13 has opened a new window to find the CP violation phase and to determine the neutrino mass hierarchy. Future reactor experiments, JUNO and RENO-50, are proposed to determine the neutrino mass hierarchy and to make highly precise measurements of θ12, Δm212, and Δm312.

  9. 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.

  10. 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.

  11. 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

  12. COBRA-IV PC: A personal computer version of COBRA-IV-I for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

    SciTech Connect

    Webb, B.J.

    1988-01-01

    COBRA-IV PC is a modified version of COBRA-IV-I, adapted for use with most IBM PC and PC-compatible desktop computers. Like COBRA-IV-I, COBRA-IV PC uses the subchannel analysis approach to determine the enthalpy and flow distribution in rod bundles for both steady-state and transient conditions. The steady-state and transient solution schemes used in COBRA-IIIC are still available in COBRA-IV PC as the implicit solution scheme option. An explicit solution scheme is also available, allowing the calculation of severe transients involving flow reversals, recirculations, expulsions, and reentry flows, with a pressure or flow boundary condition specified. In addition, several modifications have been incorporated into COBRA-IV PC to allow the code to run on the PC. These include a reduction in the array dimensions, the removal of the dump and restart options, and the inclusion of several code modifications by Oregon State University, most notably, a critical heat flux correlation for boiling water reactor fuel and a new solution scheme for cross-flow distribution calculations. 7 refs., 8 figs., 1 tab.

  13. Test Review: Advanced Clinical Solutions for WAIS-IV and WMS-IV

    ERIC Educational Resources Information Center

    Chu, Yiting; Lai, Mark H. C.; Xu, Yining; Zhou, Yuanyuan

    2012-01-01

    The authors review the "Advanced Clinical Solutions for WAIS-IV and WMS-IV". The "Advanced Clinical Solutions (ACS) for the Wechsler Adult Intelligence Scale-Fourth Edition" (WAIS-IV; Wechsler, 2008) and the "Wechsler Memory Scale-Fourth Edition" (WMS-IV; Wechsler, 2009) was published by Pearson in 2009. It is a clinical tool for extending the…

  14. 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.

  15. 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.

  16. 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.

  17. Bioconversion reactor

    SciTech Connect

    McCarty, P.L.; Bachmann, A.

    1992-02-25

    A bioconversion reactor is described 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. 7 figs.

  18. 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.

  19. 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)

  20. 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.

  1. 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.

  2. Stage IV-S neuroblastoma. Results with definitive therapy

    SciTech Connect

    Stokes, S.H.; Thomas, P.R.; Perez, C.A.; Vietti, T.J.

    1984-05-15

    The results of management of 14 patients with Stage IV-S neuroblastoma are reported. The treatment policy, although not consistent over this time span, in general used a combination of radiotherapy and chemotherapy or infrequently one modality alone. Twelve of 14 (86%) survived more than 6 years. One patient, with a solitary mediastinal primary tumor, died of rapidly progressive disease at three months. The other death occurred in a 4.5-year-old presenting with hepatomegaly at diagnosis followed by skeletal dissemination 2.5 years later. Thirteen of the patients were younger than 1 year of age. Of the 11 patients that received radiotherapy, 4 experienced mild asymptomatic scoliosis or kyphoscoliosis at 3 to 12 years after initial therapy. A review of the literature indicates that spontaneous regression in this tumor is very frequent; therefore, it is recommended that for the common presentation of massive hepatomegaly in an infant, close observation is warranted, unless life threatening complications occur. However, initial therapeutic intervention may be indicated in those patients with life threatening presentations. This data did not substantiate the necessity for complete surgical excision of the primary tumor, as has been suggested by others.

  3. 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

  4. 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.

  5. Strategies for the release of dipeptidyl peptidase IV (DPP-IV) inhibitory peptides in an enzymatic hydrolyzate of α-lactalbumin.

    PubMed

    Nongonierma, Alice B; Le Maux, Solène; Hamayon, Joël; FitzGerald, Richard J

    2016-08-10

    Bovine α-lactalbumin (α-La) contains numerous dipeptidyl peptidase IV (DPP-IV) inhibitory peptide sequences within its primary structure. In silico analysis indicated that the targeted hydrolysis of α-La with elastase should release DPP-IV inhibitory peptides. An α-La isolate was hydrolysed with elastase under different conditions using an experimental design approach incorporating 3 factors (temperature, pH and enzyme to substrate ratio (E : S) ratio) at 2 levels. The hydrolyzate generated at pH 8.5, 50 °C, E : S 2.0% (w/w) (H9) displayed the highest mean DPP-IV inhibition value at 3.1 mg mL(-1) of 75.8 ± 3.7% and had a half maximal DPP-IV inhibitory concentration (IC50) value of 1.20 ± 0.12 mg mL(-1). Five α-La-derived DPP-IV inhibitory peptides (GY, GL, GI, NY and WL) predicted to be released in silico were identified by liquid-chromatography tandem mass spectrometry (LC-MS/MS) within H9 and its simulated gastrointestinal digestion (SGID) sample. This preliminary study demonstrated the benefit of using a targeted approach combined with an experimental design in the generation of dietary protein hydrolyzates with DPP-IV inhibitory properties. PMID:27410260

  6. 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)....

  7. 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

  8. 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

  9. 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

  10. Transport reactor development status

    SciTech Connect

    Rush, R.E.; Fankhanel, M.O.; Campbell, W.M.

    1994-10-01

    This project is part of METC`s Power Systems Development Facility (PSDF) located at Wilsonville, Alabama. The primary objective of the Advanced Gasifier module is to produce vitiated gases for intermediate-term testing of Particulate Control Devices (PCDs). The Transport reactor potentially allows particle size distribution, solids loading, and particulate characteristics in the off-gas stream to be varied in a number of ways. Particulates in the hot gases from the Transport reactor will be removed in the PCDs. Two PCDs will be initially installed in the module; one a ceramic candle filter, the other a granular bed filter. After testing of the initial PCDs they will be removed and replaced with PCDs supplied by other vendors. A secondary objective is to verify the performance of a Transport reactor for use in advanced Integrated Gasification Combined Cycle (IGCC), Integrated Gasification Fuel Cell (IG-FC), and Pressurized Combustion Combined Cycle (PCCC) power generation units. This paper discusses the development of the Transport reactor design from bench-scale testing through pilot-scale testing to design of the Process Development Unit (PDU-scale) facility at Wilsonville.

  11. Neutronic reactor

    DOEpatents

    Wende, Charles W. J.; Babcock, Dale F.; Menegus, Robert L.

    1983-01-01

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. Advantages of co-located spent fuel reprocessing, repository and underground reactor facilities

    SciTech Connect

    Mahar, James M.; Kunze, Jay F.; Wes Myers, Carl; Loveland, Ryan

    2007-07-01

    The purpose of this work is to extend the discussion of potential advantages of the underground nuclear park (UNP) concept by making specific concept design and cost estimate comparisons for both present Generation III types of reactors and for some of the modular Gen IV or the GNEP modular concept. For the present Gen III types, we propose co-locating reprocessing and (re)fabrication facilities along with disposal facilities in the underground park. The goal is to determine the site costs and facility construction costs of such a complex which incorporates the advantages of a closed fuel cycle, nuclear waste repository, and ultimate decommissioning activities all within the UNP. Modular power generation units are also well-suited for placement underground and have the added advantage of construction using current and future tunnel boring machine technology. (authors)

  17. A Spectrum of IV and V Modeling Techniques

    NASA Technical Reports Server (NTRS)

    Heimdahl, Mats; Owen, David

    2004-01-01

    The aerospace industry in general and NASA in particular is using more (semi-formal) model-based software development. Model-based development produces a collection of artifacts, for example, state diagrams, module diagrams (such as class diagrams), control-block diagrams, etc. These artifacts may than be used as a basis for auto code generation for production use. Therefore, these models must be properly evaluated in the IV and V process. IV and V practitioners know how assess standard procedural systems. But what can we du about IV and V of model-based systems? The goal of the work outlined in this proposal is to use cost effective automated techniques to the largest extent possible during the IV and V process. Our working hypotheses are: 1. There exists a range of validation techniques that can assess models built using a range of modeling techniques of increasing cost and complexity. Specifically, we hypotesize that the "cheaper" techniques can find faults cheaply and early in a project. These early results are then used to predict if this is a problem system and if a more elaborate and expensive IV and V effort is justified. 2. There exists a set of migration procedures that let us seamlessly move from simple models using cheaper techniques into more elaborate models suitable for a more expensive and detailed analysis. 3. We further hypothesize that this migration process is much cheaper than simply remodeling the system under investigation from scratch when moving to models needed for the more detailed and expensive IV and V assessments.

  18. Role of RNA polymerase IV in plant small RNA metabolism.

    PubMed

    Zhang, Xiaoyu; Henderson, Ian R; Lu, Cheng; Green, Pamela J; Jacobsen, Steven E

    2007-03-13

    In addition to the three RNA polymerases (RNAP I-III) shared by all eukaryotic organisms, plant genomes encode a fourth RNAP (RNAP IV) that appears to be specialized in the production of siRNAs. Available data support a model in which dsRNAs are generated by RNAP IV and RNA-dependent RNAP 2 (RDR2) and processed by DICER (DCL) enzymes into 21- to 24-nt siRNAs, which are associated with different ARGONAUTE (AGO) proteins for transcriptional or posttranscriptional gene silencing. However, it is not yet clear what fraction of genomic siRNA production is RNAP IV-dependent, and to what extent these siRNAs are preferentially processed by certain DCL(s) or associated with specific AGOs for distinct downstream functions. To address these questions on a genome-wide scale, we sequenced approximately 335,000 siRNAs from wild-type and RNAP IV mutant Arabidopsis plants by using 454 technology. The results show that RNAP IV is required for the production of >90% of all siRNAs, which are faithfully produced from a discrete set of genomic loci. Comparisons of these siRNAs with those accumulated in rdr2 and dcl2 dcl3 dcl4 and those associated with AGO1 and AGO4 provide important information regarding the processing, channeling, and functions of plant siRNAs. We also describe a class of RNAP IV-independent siRNAs produced from endogenous single-stranded hairpin RNA precursors. PMID:17360559

  19. Taming Tin(IV) Polyazides.

    PubMed

    Campbell, Rory; Davis, Martin F; Fazakerley, Mathew; Portius, Peter

    2015-12-14

    The first charge-neutral Lewis base adducts of tin(IV) tetraazide, [Sn(N3)4(bpy)], [Sn(N3)4(phen)] and [Sn(N3)4(py)2], and the salt bis{bis(triphenylphosphine)iminium} hexa(azido)stannate [(PPN)2Sn(N3)6] (bpy = 2,2'-bipyridine; phen = 1,10-phenanthroline; py = pyridine; PPN = N(PPh3)2) have been prepared using covalent or ionic azide-transfer reagents and ligand-exchange reactions. The azides were isolated on the 0.3 to 1 g scale and characterized by IR and NMR spectroscopies, microanalytical and thermal methods and their molecular structures determined by single-crystal XRD. All complexes have a distorted octahedral Sn[N]6 coordination geometry and possess greater thermal stability than their Si and Ge homologues. The nitrogen content of the adducts of up to 44% exceed any Sn(IV) compound known hitherto. PMID:26767331

  20. MECHANISM OF CHEMICAL MUTAGENESIS IV.

    PubMed Central

    Lorkiewicz, Z.; Szybalski, Waclaw

    1961-01-01

    Lorkiewicz, Z. (University of Wisconsin, Madison), and Waclaw Szybalski. Mechanism of chemical mutagenesis. IV. Reaction between triethylene melamine and nucleic acid components. J. Bacteriol. 82: 195–201. 1961.—Triethylene melamine interacts primarily with phosphorylated intracellular deoxyribonucleic acid (DNA) precursors and not with DNA. It was found by direct chemical and chromatographic analysis that only pyrimidine precursors of nucleic acids are attacked by triethylene melamine. In the course of the triethylene melamine-deoxycytidine reaction the mutagenicity of the reaction mixture is lost, but the mutagenicity of the triethylene melamine-thymidine reaction products significantly increases above that of the reaction substrates. Several steps are postulated to explain the mechanism of the triethylene melamine-initiated mutagenic reaction: (i) Reaction I, semireversible uptake of triethylene melamine; (ii) reaction II, chemical interaction between triethylene melamine and intracellular thymidine mono- or triphosphate with the production of a functional analogue of the latter; (iii) incorporation of this fraudulent analogue into the newly formed DNA strand; (iv) occurrence of self-perpetuating errors in the sequence of natural bases during subsequent rounds of replication of the analogue-containing DNA strand. It is postulated that the mechanism of mutagenic responses to different types of mutagens can fit either a simplified (mutagenic base analogues) or extended version (radiation) of this schema. PMID:16561917