Sample records for transport reactor integrated

  1. An Advanced Integrated Diffusion/Transport Method for the Design, Analysis and Optimization of the Very-High-Temperature Reactors

    SciTech Connect (OSTI)

    Farzad Rahnema; Dingkang Zhang; Abderrafi Ougouag; Frederick Gleicher

    2011-04-04T23:59:59.000Z

    The main objective of this research is to develop an integrated diffusion/transport (IDT) method to substantially improve the accuracy of nodal diffusion methods for the design and analysis of Very High Temperature Reactors (VHTR). Because of the presence of control rods in the reflector regions in the Pebble Bed Reactor (PBR-VHTR), traditional nodal diffusion methods do not accurately model these regions, within which diffusion theory breaks down in the vicinity of high neutron absorption and steep flux gradients. The IDT method uses a local transport solver based on a new incident flux response expansion method in the controlled nodes. Diffusion theory is used in the rest of the core. This approach improves the accuracy of the core solution by generating transport solutions of controlled nodes while maintaining computational efficiency by using diffusion solutions in nodes where such a treatment is sufficient. The transport method is initially developed and coupled to the reformulated 3-D nodal diffusion model in the CYNOD code for PBR core design and fuel cycle analysis. This method is also extended to the prismatic VHTR. The new method accurately captures transport effects in highly heterogeneous regions with steep flux gradients. The calculations of these nodes with transport theory avoid errors associated with spatial homogenization commonly used in diffusion methods in reactor core simulators

  2. Spent fuel integrity during transportation

    SciTech Connect (OSTI)

    Funk, C.W.; Jacobson, L.D.

    1980-01-01T23:59:59.000Z

    The conditions of recent shipments of light water reactor spent fuel were surveyed. The radioactivity level of cask coolant was examined in an attempt to find the effects of transportation on LWR fuel assemblies. Discussion included potential cladding integrity loss mechanisms, canning requirements, changes of radioactivity levels, and comparison of transportation in wet or dry media. Although integrity loss or degradation has not been identified, radioactivity levels usually increase during transportation, especially for leaking assemblies.

  3. Integral fast reactor safety features

    SciTech Connect (OSTI)

    Cahalan, J.E.; Kramer, J.M.; Marchaterre, J.F.; Mueller, C.J.; Pedersen, D.R.; Sevy, R.H.; Wade, D.C.; Wei, T.Y.C.

    1988-01-01T23:59:59.000Z

    The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. In addition to liquid metal cooling, the principal design features that distinguish the IFR are: (1) a pool-type primary system, (2) an advanced ternary alloy metallic fuel, and (3) an integral fuel cycle with on-site fuel reprocessing and fabrication. This paper focuses on the technical aspects of the improved safety margins available in the IFR concept. This increased level of safety is made possible by (1) the liquid metal (sodium) coolant and pool-type primary system layout, which together facilitate passive decay heat removal, and (2) a sodium-bonded metallic fuel pin design with thermal and neutronic properties that provide passive core responses which control and mitigate the consequences of reactor accidents.

  4. Transportation and Stationary Power Integration: Workshop Proceedings...

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

    Integration: Workshop Proceedings Transportation and Stationary Power Integration: Workshop Proceedings Proceedings for the Transportation and Stationary Power Integration Workshop...

  5. Particle transport in plasma reactors

    SciTech Connect (OSTI)

    Rader, D.J.; Geller, A.S.; Choi, Seung J. [Sandia National Labs., Albuquerque, NM (United States); Kushner, M.J. [Illinois Univ., Urbana, IL (United States)

    1995-01-01T23:59:59.000Z

    SEMATECH and the Department of Energy have established a Contamination Free Manufacturing Research Center (CFMRC) located at Sandia National Laboratories. One of the programs underway at the CFMRC is directed towards defect reduction in semiconductor process reactors by the application of computational modeling. The goal is to use fluid, thermal, plasma, and particle transport models to identify process conditions and tool designs that reduce the deposition rate of particles on wafers. The program is directed toward defect reduction in specific manufacturing tools, although some model development is undertaken when needed. The need to produce quantifiable improvements in tool defect performance requires the close cooperation among Sandia, universities, SEMATECH, SEMATECH member companies, and equipment manufacturers. Currently, both plasma (e.g., etch, PECVD) and nonplasma tools (e.g., LPCVD, rinse tanks) are being worked on under this program. In this paper the authors summarize their recent efforts to reduce particle deposition on wafers during plasma-based semiconductor manufacturing.

  6. Integrated Transportation System Design Optimization

    E-Print Network [OSTI]

    Integrated Transportation System Design Optimization by Christine Taylor B.S. Cornell University by . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Professor Jaime Peraire Chairman, Department Graduate Committee #12;2 #12;Integrated Transportation System Abstract Traditionally, the design of a transportation system has focused on either the vehicle design

  7. Transportation and Stationary Power Integration Workshop Attendees...

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

    Attendees List Transportation and Stationary Power Integration Workshop Attendees List List of attendees for the Transportation and Stationary Power Integration Workshop...

  8. Particle Transport in Parallel-Plate Reactors

    SciTech Connect (OSTI)

    Rader, D.J.; Geller, A.S.

    1999-08-01T23:59:59.000Z

    A major cause of semiconductor yield degradation is contaminant particles that deposit on wafers while they reside in processing tools during integrated circuit manufacturing. This report presents numerical models for assessing particle transport and deposition in a parallel-plate geometry characteristic of a wide range of single-wafer processing tools: uniform downward flow exiting a perforated-plate showerhead separated by a gap from a circular wafer resting on a parallel susceptor. Particles are assumed to originate either upstream of the showerhead or from a specified position between the plates. The physical mechanisms controlling particle deposition and transport (inertia, diffusion, fluid drag, and external forces) are reviewed, with an emphasis on conditions encountered in semiconductor process tools (i.e., sub-atmospheric pressures and submicron particles). Isothermal flow is assumed, although small temperature differences are allowed to drive particle thermophoresis. Numerical solutions of the flow field are presented which agree with an analytic, creeping-flow expression for Re < 4. Deposition is quantified by use of a particle collection efficiency, which is defined as the fraction of particles in the reactor that deposit on the wafer. Analytic expressions for collection efficiency are presented for the limiting case where external forces control deposition (i.e., neglecting particle diffusion and inertia). Deposition from simultaneous particle diffusion and external forces is analyzed by an Eulerian formulation; for creeping flow and particles released from a planar trap, the analysis yields an analytic, integral expression for particle deposition based on process and particle properties. Deposition from simultaneous particle inertia and external forces is analyzed by a Lagrangian formulation, which can describe inertia-enhanced deposition resulting from particle acceleration in the showerhead. An approximate analytic expression is derived for particle velocity at the showerhead exit as a function of showerhead geometry, flow rate, and gas and particle properties. The particle showerhead-exit velocity is next used as an initial condition for particle transport between the plates to determine whether the particle deposits on the wafer, as a function of shower-head-exit particle velocity, the plate separation, flow rate, and gas and particle properties. Based on the numerical analysis, recommendations of best practices are presented that should help tool operators and designers reduce particle deposition in real tools. These guidelines are not intended to replace detailed calculations, but to provide the user with a general feel for inherently-clean practices.

  9. Multilevel transport solution of LWR reactor cores

    SciTech Connect (OSTI)

    Jose Ignacio Marquez Damian; Cassiano R.E. de Oliveira; HyeonKae Park

    2008-09-01T23:59:59.000Z

    This work presents a multilevel approach for the solution of the transport equation in typical LWR assemblies and core configurations. It is based on the second-order, even-parity formulation of the transport equation, which is solved within the framework provided by the finite element-spherical harmonics code EVENT. The performance of the new solver has been compared with that of the standard conjugate gradient solver for diffusion and transport problems on structured and unstruc-tured grids. Numerical results demonstrate the potential of the multilevel scheme for realistic reactor calculations.

  10. Integrated reformer and shift reactor

    DOE Patents [OSTI]

    Bentley, Jeffrey M.; Clawson, Lawrence G.; Mitchell, William L.; Dorson, Matthew H.

    2006-06-27T23:59:59.000Z

    A hydrocarbon fuel reformer for producing diatomic hydrogen gas is disclosed. The reformer includes a first reaction vessel, a shift reactor vessel annularly disposed about the first reaction vessel, including a first shift reactor zone, and a first helical tube disposed within the first shift reactor zone having an inlet end communicating with a water supply source. The water supply source is preferably adapted to supply liquid-phase water to the first helical tube at flow conditions sufficient to ensure discharge of liquid-phase and steam-phase water from an outlet end of the first helical tube. The reformer may further include a first catalyst bed disposed in the first shift reactor zone, having a low-temperature shift catalyst in contact with the first helical tube. The catalyst bed includes a plurality of coil sections disposed in coaxial relation to other coil sections and to the central longitudinal axis of the reformer, each coil section extending between the first and second ends, and each coil section being in direct fluid communication with at least one other coil section.

  11. Transportation and Stationary Power Integration Workshop Agenda...

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

    Agenda, October 27, 2008, Phoenix, Arizonia Transportation and Stationary Power Integration Workshop Agenda, October 27, 2008, Phoenix, Arizonia Agenda for the Transportation and...

  12. Transportation and Stationary Power Integration Workshop Session...

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

    Integration Workshop Session II: State and Industry Perspectives Transportation and Stationary Power Integration Workshop Session II: State and Industry Perspectives Opportunities...

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

    SciTech Connect (OSTI)

    Michael Swanson; Daniel Laudal

    2008-03-31T23:59:59.000Z

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

  14. Optimum Reactor Outlet Temperatures for High Temperature Gas-Cooled Reactors Integrated with Industrial Processes

    SciTech Connect (OSTI)

    Lee O. Nelson

    2011-04-01T23:59:59.000Z

    This report summarizes the results of a temperature sensitivity study conducted to identify the optimum reactor operating temperatures for producing the heat and hydrogen required for industrial processes associated with the proposed new high temperature gas-cooled reactor. This study assumed that primary steam outputs of the reactor were delivered at 17 MPa and 540°C and the helium coolant was delivered at 7 MPa at 625–925°C. The secondary outputs of were electricity and hydrogen. For the power generation analysis, it was assumed that the power cycle efficiency was 66% of the maximum theoretical efficiency of the Carnot thermodynamic cycle. Hydrogen was generated via the hightemperature steam electrolysis or the steam methane reforming process. The study indicates that optimum or a range of reactor outlet temperatures could be identified to further refine the process evaluations that were developed for high temperature gas-cooled reactor-integrated production of synthetic transportation fuels, ammonia, and ammonia derivatives, oil from unconventional sources, and substitute natural gas from coal.

  15. Supertruck - Improving Transportation Efficiency through Integrated...

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

    Research Supertruck - Improving Transportation Efficiency through Integrated Vehicle, Engine and Powertrain Research 2012 DOE Hydrogen and Fuel Cells Program and Vehicle...

  16. Transportation and Stationary Power Integration with Hydrogen...

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

    with Hydrogen and Fuel Cell Technology in Connecticut Transportation and Stationary Power Integration with Hydrogen and Fuel Cell Technology in Connecticut Overview of strengths,...

  17. Transportation and Stationary Power Integration Workshop: ""An...

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

    ""An Automaker's Views on the Transition to Hydrogen and Fuel Cell Vehicles Transportation and Stationary Power Integration Workshop: ""An Automaker's Views on the Transition to...

  18. Adaptive Nodal Transport Methods for Reactor Transient Analysis

    SciTech Connect (OSTI)

    Thomas Downar; E. Lewis

    2005-08-31T23:59:59.000Z

    Develop methods for adaptively treating the angular, spatial, and time dependence of the neutron flux in reactor transient analysis. These methods were demonstrated in the DOE transport nodal code VARIANT and the US NRC spatial kinetics code, PARCS.

  19. Integrated transportation system design optimization

    E-Print Network [OSTI]

    Taylor, Christine P. (Christine Pia), 1979-

    2007-01-01T23:59:59.000Z

    Traditionally, the design of a transportation system has focused on either the vehicle design or the network flow, assuming the other as given. However, to define a system level architecture for a transportation system, ...

  20. Integral reactor system and method for fuel cells

    DOE Patents [OSTI]

    Fernandes, Neil Edward; Brown, Michael S; Cheekatamarla, Praveen; Deng, Thomas; Dimitrakopoulos, James; Litka, Anthony F

    2013-11-19T23:59:59.000Z

    A reactor system is integrated internally within an anode-side cavity of a fuel cell. The reactor system is configured to convert hydrocarbons to smaller species while mitigating the lower production of solid carbon. The reactor system may incorporate one or more of a pre-reforming section, an anode exhaust gas recirculation device, and a reforming section.

  1. Integrated autothermal reactor concepts for oxidative coupling and reforming of

    E-Print Network [OSTI]

    Twente, Universiteit

    #12;Integrated autothermal reactor concepts for oxidative coupling and reforming of methane #12-90-365-2985-3, DOI 10.3990/1.9789036529853 #12;INTEGRATED AUTOTHERMAL REACTOR CONCEPTS FOR OXIDATIVE COUPLING . . . . . . . . . . . . . . . . . . . . 31 2.5 Dual function catalyst particle . . . . . . . . . . . . . . . . . . . . . . 34 2.6 Autothermal

  2. NEW OPTIMIZATION-BASED APPROACH TO CHEMICAL REACTOR SYNTHESIS TOWARDS THE FULL INTEGRATION OF REACTOR

    E-Print Network [OSTI]

    Van den Hof, Paul

    NEW OPTIMIZATION-BASED APPROACH TO CHEMICAL REACTOR SYNTHESIS ­ TOWARDS THE FULL INTEGRATION OF REACTOR DESIGN, OPERATION AND CONTROL Nikola Nikacevi1 , Adrie Huesman1 , Paul Van den Hof1 , Andrzej, Leeghwaterstraat 44, 2628 CA Delft, The Netherlands Conceptual chemical reactor design still follows traditional

  3. Evaluation of in-vessel corium retention through external reactor vessel cooling for integral reactor

    SciTech Connect (OSTI)

    Park, R. J.; Lee, J. R.; Kim, S. B.; Jin, Y.; Kim, H. Y. [Korea Atomic Energy Research Inst., 1045 Daedeok-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

    2012-07-01T23:59:59.000Z

    In-vessel corium retention through external reactor vessel cooling (IVR-ERVC) for a small integral reactor has been evaluated to determine the thermal margin for the prevention of a reactor vessel failure. A thermal load analysis from the corium pool to the outer reactor vessel wall in the lower plenum of the reactor vessel has been performed to determine the heat flux distribution. The critical heat flux (CHF) on the outer reactor vessel wall has been determined to fix the maximum heat removal rate through the external coolant between the outer reactor vessel and the insulation of the reactor vessel. Finally, the thermal margin has been evaluated by comparison of the thermal load with the maximum heat removal rate of the CHF on the outer reactor vessel wall. The maximum heat flux from the corium pool to the outer reactor vessel is estimated at approximately 0.25 MW/m{sup 2} in the metallic layer because of the focusing effect. The CHF of the outer reactor vessel is approximately 1.1 MW/m{sup 2} because of a two phase natural circulation mass flow. Since the thermal margin for the IVR-ERVC is sufficient, the reactor vessel integrity is maintained during a severe accident of a small integral reactor. (authors)

  4. A high integrity transportable supercompactor

    SciTech Connect (OSTI)

    Sims, J.; Schmidt, G.

    1993-12-31T23:59:59.000Z

    Current transportable equipment produced to utilize high force compaction to reduce the overall volume of drums containing solid radioactive material prior to disposal, were originally designed to standards which will eventually become obsolete. At the time these machines were produced, they were state-of-the-art, but are now indicating their weaknesses in operational and safety aspects. This paper formulates a concept for a Transportable Supercompactor for handling alpha and beta/gamma bearing wastes, low operator dose uptake, contamination and radiation control systems, liquids collection, the maintenance demands of a contained press, etc., taking into account the latest technical and safety considerations. The possibility of using the concept as a skid mounted fixed Supercompactor is also reviewed in this paper.

  5. Neutron transport analysis for nuclear reactor design

    DOE Patents [OSTI]

    Vujic, Jasmina L. (Lisle, IL)

    1993-01-01T23:59:59.000Z

    Replacing regular mesh-dependent ray tracing modules in a collision/transfer probability (CTP) code with a ray tracing module based upon combinatorial geometry of a modified geometrical module (GMC) provides a general geometry transfer theory code in two dimensions (2D) for analyzing nuclear reactor design and control. The primary modification of the GMC module involves generation of a fixed inner frame and a rotating outer frame, where the inner frame contains all reactor regions of interest, e.g., part of a reactor assembly, an assembly, or several assemblies, and the outer frame, with a set of parallel equidistant rays (lines) attached to it, rotates around the inner frame. The modified GMC module allows for determining for each parallel ray (line), the intersections with zone boundaries, the path length between the intersections, the total number of zones on a track, the zone and medium numbers, and the intersections with the outer surface, which parameters may be used in the CTP code to calculate collision/transfer probability and cross-section values.

  6. Neutron transport analysis for nuclear reactor design

    DOE Patents [OSTI]

    Vujic, J.L.

    1993-11-30T23:59:59.000Z

    Replacing regular mesh-dependent ray tracing modules in a collision/transfer probability (CTP) code with a ray tracing module based upon combinatorial geometry of a modified geometrical module (GMC) provides a general geometry transfer theory code in two dimensions (2D) for analyzing nuclear reactor design and control. The primary modification of the GMC module involves generation of a fixed inner frame and a rotating outer frame, where the inner frame contains all reactor regions of interest, e.g., part of a reactor assembly, an assembly, or several assemblies, and the outer frame, with a set of parallel equidistant rays (lines) attached to it, rotates around the inner frame. The modified GMC module allows for determining for each parallel ray (line), the intersections with zone boundaries, the path length between the intersections, the total number of zones on a track, the zone and medium numbers, and the intersections with the outer surface, which parameters may be used in the CTP code to calculate collision/transfer probability and cross-section values. 28 figures.

  7. Advanced Reactors Thermal Energy Transport for Process Industries

    SciTech Connect (OSTI)

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

    2014-07-01T23:59:59.000Z

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

  8. Natural convection heat transport in a small, HLMC reactor system

    SciTech Connect (OSTI)

    Spencer, B.W.; Sienicki, J.J.; Farmer, M.T. [Argonne National Lab., IL (United States)

    1999-09-01T23:59:59.000Z

    Concepts are being developed and evaluated at Argonne National Laboratory for a small nuclear steam supply system (NSSS) with proliferation-resistant features targeted for export to developing countries. Here the authors are specifically investigating how simple and compact such a system can be. A heavy-liquid-metal coolant (HLMC) is being considered owing to its excellent heat transport characteristics and its relative inertness with the reference thermodynamic working fluid (water/steam). The purpose of the present work is to explore the possibility to take advantage of these HLMC characteristics by eliminating the intermediate loop needed in sodium-cooled systems and additionally eliminating the primary system coolant pumps. The criteria imposed on the system include the following: (1) low power, i.e., 300 MW(thermal); (2) small size for factory fabrication and overland transportation; (3) elimination of fuel access at the site (no refueling, fuel shuffling, nor storage at site); integral fueled module replacement at 15-yr goal interval; and (4) completion of all research and development needed for detailed prototype design within 5 yr. To accomplish the latter requirement, the authors are addressing whether existing coolant and materials technology is capable of supporting the sought-after simplifications. In this regard, they are at present considering technology developed in Russia for Pb-Bi eutectic as a reactor coolant and ferritic-martensitic stainless steel with oxide-layer corrosion protection as cladding. The figure of merit in the investigation is the peak cladding temperature insofar as the cladding technology is considered proven to {approximately}600 C.

  9. A Small Secure Transportable Autonomous Lead-Cooled Fast Reactor for Deployment at Remote Sites

    SciTech Connect (OSTI)

    Sienicki, J .J.; Smith, M.A.; Mosseytsev, A.V.; Yang, W.S.; Wade, D.C.

    2004-10-06T23:59:59.000Z

    This presentation discusses a small secure transportable autonomous lead-cooled fast reactor for deployment at remote sites.

  10. TRANSPORTATION ENERGY FORECASTS FOR THE 2007 INTEGRATED ENERGY

    E-Print Network [OSTI]

    has developed longterm forecasts of transportation energy demand as well as projected ranges of transportation fuel and crude oil import requirements. The transportation energy demand forecasts makeCALIFORNIA ENERGY COMMISSION TRANSPORTATION ENERGY FORECASTS FOR THE 2007 INTEGRATED ENERGY POLICY

  11. The Integral Fast Reactor: A practical approach to waste management

    SciTech Connect (OSTI)

    Laidler, J.J.

    1993-12-31T23:59:59.000Z

    This report discusses development of the method for pyroprocessing of spent fuel from the Integral Fast Reactor (or Advanced Liquid Metal Reactor). The technology demonstration phase, in which recycle will be demonstrated with irradiated fuel from the EBR-II reactor has been reached. Methods for recovering actinides from spent LWR fuel are at an earlier stage of development but appear to be technically feasible at this time, and a large-scale demonstration of this process has begun. The utilization of fully compatible processes for recycling valuable spent fuel materials promises to provide substantial economic incentives for future applications of the pyroprocessing technology.

  12. Integrated intelligent systems in advanced reactor control rooms

    SciTech Connect (OSTI)

    Beckmeyer, R.R.

    1989-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2010-02-01T23:59:59.000Z

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

  14. Numerical simulations of ion transport membrane oxy-fuel reactors for CO? capture applications

    E-Print Network [OSTI]

    Hong, Jongsup

    2013-01-01T23:59:59.000Z

    Numerical simulations were performed to investigate the key features of oxygen permeation and hydrocarbon conversion in ion transport membrane (ITM) reactors. ITM reactors have been suggested as a novel technology to enable ...

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

    SciTech Connect (OSTI)

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

    1994-12-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1994-10-01T23:59:59.000Z

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

  17. Integrated systems analysis of the PIUS reactor

    SciTech Connect (OSTI)

    Fullwood, F.; Kroeger, P.; Higgins, J. [Brookhaven National Lab., Upton, NY (United States)] [and others

    1993-11-01T23:59:59.000Z

    Results are presented of a systems failure analysis of the PIUS plant systems that are used during normal reactor operation and postulated accidents. This study was performed to provide the NRC with an understanding of the behavior of the plant. The study applied two diverse failure identification methods, Failure Modes Effects & Criticality Analysis (FMECA) and Hazards & Operability (HAZOP) to the plant systems, supported by several deterministic analyses. Conventional PRA methods were also used along with a scheme for classifying events by initiator frequency and combinations of failures. Principal results of this study are: (a) an extensive listing of potential event sequences, grouped in categories that can be used by the NRC, (b) identification of support systems that are important to safety, and (c) identification of key operator actions.

  18. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    SciTech Connect (OSTI)

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-04-01T23:59:59.000Z

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.

  19. The design of a reduced diameter Pebble Bed Modular Reactor for reactor pressure vessel transport by railcar

    E-Print Network [OSTI]

    Everson, Matthew S

    2009-01-01T23:59:59.000Z

    Many desirable locations for Pebble Bed Modular Reactors are currently out of consideration as construction sites for current designs due to limitations on the mode of transportation for large RPVs. In particular, the ...

  20. Reduced Order Model Compensator Control of Species Transport in a CVD Reactor

    E-Print Network [OSTI]

    Reduced Order Model Compensator Control of Species Transport in a CVD Reactor G.M. Kepler, H for computation of feedback controls and compensators in a high pressure chemical vapor deposition (HPCVD) reactor University to design and build such a HPCVD reactor with real­time sensing and control as an innovative

  1. Transportation and Stationary Power Integration Workshop: A California...

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

    A California Perspective Transportation and Stationary Power Integration Workshop: A California Perspective Overview of California regulations, latest funded hydrogen stations, and...

  2. Nuclear reactor heat transport system component low friction support system

    DOE Patents [OSTI]

    Wade, Elman E. (Ruffs Dale, PA)

    1980-01-01T23:59:59.000Z

    A support column for a heavy component of a liquid metal fast breeder reactor heat transport system which will deflect when the pipes leading coolant to and from the heavy component expand or contract due to temperature changes includes a vertically disposed pipe, the pipe being connected to the heavy component by two longitudinally spaced cycloidal dovetail joints wherein the distal end of each of the dovetails constitutes a part of the surface of a large diameter cylinder and the centerlines of these large diameter cylinders intersect at right angles and the pipe being supported through two longitudinally spaced cycloidal dovetail joints wherein the distal end of each of the dovetails constitutes a part of the surface of a large diameter cylinder and the centerlines of these large diameter cylinders intersect at right angles, each of the cylindrical surfaces bearing on a flat and horizontal surface.

  3. (COMEDIE program review and fission product transport in MHTGR reactor)

    SciTech Connect (OSTI)

    Stansfield, O.M.

    1990-03-15T23:59:59.000Z

    The subcontract between Martin Marietta Energy Systems, Inc., and the CEA provides for the refurbishment of the high pressure COMEDIE test loop in the SILOE reactor and a series of experiments to characterize fission product lift-off from MHTGR heat exchanger surfaces under several depressurization accident scenarios. The data will contribute to the validation of models and codes used to predict fission product transport in the MHTGR. In the meeting at CEA headquarters in Paris the program schedule and preparation for the DCAA and Quality Assurance audits were discussed. Long-range interest in expanded participation in the gas-cooled reactor technology Umbrella Agreement was also expressed by the CEA. At the CENG, in Grenoble, technical details on the loop design, fabrication components, development of test procedures, and preparation for the DOE quality assurance (QA) audit in May were discussed. After significant delays in CY 1989 it appears that good progress is being made in CY 1990 and the first major test will be initiated by December. An extensive list of agreements and commitments was generated to facilitate the coordination and planning of future work. 2 figs., 2 tabs.

  4. Oxygen transport membrane system and method for transferring heat to catalytic/process reactors

    DOE Patents [OSTI]

    Kelly, Sean M; Kromer, Brian R; Litwin, Michael M; Rosen, Lee J; Christie, Gervase Maxwell; Wilson, Jamie R; Kosowski, Lawrence W; Robinson, Charles

    2014-01-07T23:59:59.000Z

    A method and apparatus for producing heat used in a synthesis gas production is provided. The disclosed method and apparatus include a plurality of tubular oxygen transport membrane elements adapted to separate oxygen from an oxygen containing stream contacting the retentate side of the membrane elements. The permeated oxygen is combusted with a hydrogen containing synthesis gas stream contacting the permeate side of the tubular oxygen transport membrane elements thereby generating a reaction product stream and radiant heat. The present method and apparatus also includes at least one catalytic reactor containing a catalyst to promote the stream reforming reaction wherein the catalytic reactor is surrounded by the plurality of tubular oxygen transport membrane elements. The view factor between the catalytic reactor and the plurality of tubular oxygen transport membrane elements radiating heat to the catalytic reactor is greater than or equal to 0.5.

  5. On the Criticality Safety of Transuranic Sodium Fast Reactor Fuel Transport Casks

    SciTech Connect (OSTI)

    Samuel Bays; Ayodeji Alajo

    2010-05-01T23:59:59.000Z

    This work addresses the neutronic performance and criticality safety issues of transport casks for fuel pertaining to low conversion ratio sodium cooled fast reactors, conventionally known as Advanced Burner Reactors. The criticality of a one, three, seven and 19-assembly cask capacity is presented. Both dry “helium” and flooded “water” filled casks are considered. No credit for fuel burnup or fission products was assumed. As many as possible of the conservatisms used in licensing light water reactor universal transport casks were incorporated into this SFR cask criticality design and analysis. It was found that at 7-assemblies or more, adding moderator to the SFR cask increases criticality margin. Also, removal of MAs from the fuel increases criticality margin of dry casks and takes a slight amount of margin away for wet casks. Assuming credit for borated fuel tube liners, this design analysis suggests that as many as 19 assemblies can be loaded in a cask if limited purely by criticality safety. If no credit for boron is assumed, the cask could possibly hold seven assemblies if low conversion ratio fast reactor grade fuel and not breeder reactor grade fuel is assumed. The analysis showed that there is a need for new cask designs for fast reactors spent fuel transportation. There is a potential of modifying existing transportation cask design as the starting point for fast reactor spent fuel transportation.

  6. KINETICS OF HOT-GAS DESULFURIZATION SORBENTS FOR TRANSPORT REACTORS

    SciTech Connect (OSTI)

    K.C. Kwon

    2002-01-01T23:59:59.000Z

    Hot-gas desulfurization for the integrated gasification combined cycle (IGCC) process has been investigated by many researchers to remove effectively hydrogen sulfide with various metal oxide sorbents at elevated temperatures. Various metal oxide sorbents are formulated with metal oxides such as Fe, Co, Zn, and Ti. Initial reaction kinetics of formulated sorbents with hydrogen sulfide is studied in the presence of various amounts of moisture and hydrogen at various reaction temperatures. The objectives of this research are to study initial reaction kinetics for a sorbent-hydrogen sulfide heterogeneous reaction system, to investigate effects of concentrations of hydrogen sulfide, hydrogen, and moisture on dynamic absorption of H{sub 2}S into sorbents, and to evaluate effects of temperature and sorbent amounts on dynamic absorption of H{sub 2}S into sorbents. Experimental data on initial reaction kinetics of hydrogen sulfide with metal oxide sorbents were obtained with a 0.83-cm{sup 3} differential reactor. The reactivity of MCRH-67 was examined in this report. This sorbent was obtained from the Research Triangle Institute (RTI). The sorbent in the form of 130 mm particles are reacted with 18000-ppm hydrogen sulfide at 350-525 C. The range of space time of reaction gas mixtures is 0.069-0.088 s. The range of reaction duration is 4-180 s.

  7. TRANSPORTATION ENERGY FORECASTS FOR THE 2007 INTEGRATED ENERGY

    E-Print Network [OSTI]

    requirements. The transportation energy demand forecasts make assumptions about fuel price forecastsCALIFORNIA ENERGY COMMISSION TRANSPORTATION ENERGY FORECASTS FOR THE 2007 INTEGRATED ENERGY POLICY ENERGY COMMISSION Gordon Schremp, Jim Page, and Malachi Weng-Gutierrez Principal Authors Jim Page Project

  8. Integration of Groundwater Transport Models with Wireless Sensor Networks

    E-Print Network [OSTI]

    Han, Qi "Chee"

    Integration of Groundwater Transport Models with Wireless Sensor Networks Kevin Barnhart1 , I.jayasumana@colostate.edu, Fort Collins, CO, USA ABSTRACT Groundwater transport modeling is intended to aid in remediation be conceptualized in the WSN context. INTRODUCTION As groundwater contamination is an established problem with many

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

    SciTech Connect (OSTI)

    Pablo Rubiolo, Principal Investigator

    2003-03-21T23:59:59.000Z

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

  10. NREL: Transportation Research - Electric Vehicle Grid Integration

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What's Possible for Renewable Energy: Grid IntegrationReportTransmission PlanningCapabilities

  11. NREL: Transportation Research - Systems Analysis and Integration

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What's Possible for Renewable Energy: Grid IntegrationReportTransmissionResearchNews NRELSuccessSystems

  12. THE DYNAMICS OF CHEMICAL REACTORS WITH HEAT INTEGRATION

    E-Print Network [OSTI]

    Skogestad, Sigurd

    1958 analyzed the stability of the steady states of autothermal reactors. Author to whom correspondence behavior in autothermal reactors was pre- sented by Reilly an Schmitz 1966,1967 and Pareja and Reilly 1969

  13. THE DYNAMICS OF CHEMICAL REACTORS WITH HEAT INTEGRATION

    E-Print Network [OSTI]

    Skogestad, Sigurd

    ) and Aris and Amundson (1958) analyzed the stability of the steady states of autothermal reactors. \\Lambda: 47­7359­4080. Limit cycle behavior in autothermal reactors was pre­ sented by Reilly an Schmitz (1966

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

    SciTech Connect (OSTI)

    Michael L. Swanson

    2005-08-30T23:59:59.000Z

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

  15. Integrated simulation of ELM Triggered by Pellet Through Energy Absorption and Transport Enhancement

    E-Print Network [OSTI]

    Integrated simulation of ELM Triggered by Pellet Through Energy Absorption and Transport Enhancement

  16. Integrated Simulation of ELM Triggered by Pellet through Energy Absorption and Transport Enhancement

    E-Print Network [OSTI]

    Integrated Simulation of ELM Triggered by Pellet through Energy Absorption and Transport Enhancement

  17. Hybrid Plasma Reactor/Filter for Transportable Collective Protection Systems

    SciTech Connect (OSTI)

    Josephson, Gary B.; Tonkyn, Russell G.; Frye, J. G.; Riley, Brian J.; Rappe, Kenneth G.

    2011-04-06T23:59:59.000Z

    Pacific Northwest National Laboratory (PNNL) has performed an assessment of a Hybrid Plasma/Filter system as an alternative to conventional methods for collective protection. The key premise of the hybrid system is to couple a nonthermal plasma (NTP) reactor with reactive adsorption to provide a broader envelope of protection than can be provided through a single-solution approach. The first step uses highly reactive species (e.g. oxygen radicals, hydroxyl radicals, etc.) created in a nonthermal plasma (NTP) reactor to destroy the majority (~75% - 90%) of an incoming threat. Following the NTP reactor an O3 reactor/filter uses the O3 created in the NTP reactor to further destroy the remaining organic materials. This report summarizes the laboratory development of the Hybrid Plasma Reactor/Filter to protect against a ‘worst-case’ simulant, methyl bromide (CH3Br), and presents a preliminary engineering assessment of the technology to Joint Expeditionary Collective Protection performance specifications for chemical vapor air purification technologies.

  18. Pyrometallurgical processing of Integral Fast Reactor metal fuels

    SciTech Connect (OSTI)

    Battles, J.E.; Miller, W.E.; Gay, E.C.

    1991-01-01T23:59:59.000Z

    The pyrometallurgical process for recycling spent metal fuels from the Integral Fast Reactor is now in an advanced state of development. This process involves electrorefining spent fuel with a cadmium anode, solid and liquid cathodes, and a molten salt electrolyte (LiCl-KCl) at 500{degrees}C. The initial process feasibility and flowsheet verification studies have been conducted in a laboratory-scale electrorefiner. Based on these studies, a dual cathode approach has been adopted, where uranium is recovered on a solid cathode mandrel and uranium-plutonium is recovered in a liquid cadmium cathode. Consolidation and purification (salt and cadmium removal) of uranium and uranium-plutonium products from the electrorefiner have been successful. The process is being developed with the aid of an engineering-scale electrorefiner, which has been successfully operated for more than three years. In this electrorefiner, uranium has been electrotransported from the cadmium anode to a solid cathode in 10 kg quantities. Also, anodic dissolution of 10 kg batches of chopped, simulated fuel (U--10% Zr) has been demonstrated. Development of the liquid cadmium cathode for recovering uranium-plutonium is under way.

  19. Integrated System for Retrieval, Transportation and Consolidated Storage of Used Nuclear Fuel in the US - 13312

    SciTech Connect (OSTI)

    Bracey, William; Bondre, Jayant; Shelton, Catherine [Transnuclear, Inc., 7135 Minstrel Way Suite 300, Columbia MD 21045 (United States)] [Transnuclear, Inc., 7135 Minstrel Way Suite 300, Columbia MD 21045 (United States); Edmonds, Robert [AREVA Federal Services, 7207 IBM Drive, Charlotte NC 28262 (United States)] [AREVA Federal Services, 7207 IBM Drive, Charlotte NC 28262 (United States)

    2013-07-01T23:59:59.000Z

    The current inventory of used nuclear fuel assemblies (UNFAs) from commercial reactor operations in the United States totals approximately 65,000 metric tons or approximately 232,000 UNFAs primarily stored at the 104 operational reactors in the US and a small number of decommissioned reactors. This inventory is growing at a rate of roughly 2,000 to 2,400 metric tons each year, (Approx. 7,000 UNFAs) as a result of ongoing commercial reactor operations. Assuming an average of 10 metric tons per storage/transportation casks, this inventory of commercial UNFAs represents about 6,500 casks with an additional of about 220 casks every year. In January 2010, the Blue Ribbon Commission (BRC) [1] was directed to conduct a comprehensive review of policies for managing the back end of the nuclear fuel cycle and recommend a new plan. The BRC issued their final recommendations in January 2012. One of the main recommendations is for the United States to proceed promptly to develop one or more consolidated storage facilities (CSF) as part of an integrated, comprehensive plan for safely managing the back end of the nuclear fuel cycle. Based on its extensive experience in storage and transportation cask design, analysis, licensing, fabrication, and operations including transportation logistics, Transnuclear, Inc. (TN), an AREVA Subsidiary within the Logistics Business Unit, is engineering an integrated system that will address the complete process of commercial UNFA management. The system will deal with UNFAs in their current storage mode in various configurations, the preparation including handling and additional packaging where required and transportation of UNFAs to a CSF site, and subsequent storage, operation and maintenance at the CSF with eventual transportation to a future repository or recycling site. It is essential to proceed by steps to ensure that the system will be the most efficient and serve at best its purpose by defining: the problem to be resolved, the criteria to evaluate the solutions, and the alternative solutions. The complexity of the project is increasing with time (more fuel assemblies, new storage systems, deteriorating logistics infrastructure at some sites, etc.) but with the uncertainty on the final disposal path, flexibility and simplicity will be critical. (authors)

  20. Subtask 7.4 - Power River Basin Subbituminous Coal-Biomass Cogasification Testing in a Transport Reactor

    SciTech Connect (OSTI)

    Michael Swanson; Daniel Laudal

    2009-03-01T23:59:59.000Z

    The U.S. Department of Energy (DOE) National Energy Technology Laboratory Office of Coal and Environmental Systems has as its mission to develop advanced gasification-based technologies for affordable, efficient, zero-emission power generation. These advanced power systems, which are expected to produce near-zero pollutants, are an integral part of DOE's Vision 21 Program. DOE has also been developing advanced gasification systems that lower the capital and operating costs of producing syngas for chemical production. A transport reactor has shown potential to be a low-cost syngas producer compared to other gasification systems since its high-throughput-per-unit cross-sectional area reduces capital costs. This work directly supports the Power Systems Development Facility utilizing the Kellogg Brown and Root transport reactor located at the Southern Company Services Wilsonville, Alabama, site. Over 3600 hours of operation on 17 different coals ranging from bituminous to lignite along with a petroleum coke has been completed to date in the pilot-scale transport reactor development unit (TRDU) at the Energy & Environmental Research Center (EERC). The EERC has established an extensive database on the operation of these various fuels in both air- and oxygen-blown modes utilizing a pilot-scale transport reactor gasifier. This database has been useful in determining the effectiveness of design changes on an advanced transport reactor gasifier and for determining the performance of various feedstocks in a transport reactor. The effects of different fuel types on both gasifier performance and the operation of the hot-gas filter system have been determined. It has been demonstrated that corrected fuel gas heating values ranging from 90 to 130 Btu/scf have been achieved in air-blown mode, while heating values up to 230 Btu/scf on a dry basis have been achieved in oxygen-blown mode. Carbon conversions up to 90% have also been obtained and are highly dependent on the oxygen-coal ratio. Higher-reactivity (low-rank) coals appear to perform better in a transport reactor than the less reactive bituminous coals. Factors that affect TRDU product gas quality appear to be coal type, temperature, and oxygen/fuel ratios. During this series of tests, a previously tested baseline Powder River Basin (PRB) subbituminous coal from the Peabody Energy North Antelope Rochelle Mine near Gillette, Wyoming was mixed with 20 wt% biomass. Two types of biomass were used - wood waste and switchgrass. Gas and particulate sampling at several locations in the riser provided information on coal devolatilization and cracking chemistry as a function of residence time, transport gas, and mode of operation. The goal of these tests was to compare the operating data and sample chemistry of the coal-biomass mixture to the PRB coal, with a focus on Fischer-Tropsch coal-to-liquid production in oxygen-blown mode. Data are to be provided to DOE to determine kinetic rates of devolatilization and tar cracking.

  1. The Ongoing Impact of the U.S. Fast Reactor Integral Experiments Program

    SciTech Connect (OSTI)

    John D. Bess; Michael A. Pope; Harold F. McFarlane

    2012-11-01T23:59:59.000Z

    The creation of a large database of integral fast reactor physics experiments advanced nuclear science and technology in ways that were unachievable by less capital intensive and operationally challenging approaches. They enabled the compilation of integral physics benchmark data, validated (or not) analytical methods, and provided assurance of future rector designs The integral experiments performed at Argonne National Laboratory (ANL) represent decades of research performed to support fast reactor design and our understanding of neutronics behavior and reactor physics measurements. Experiments began in 1955 with the Zero Power Reactor No. 3 (ZPR-3) and terminated with the Zero Power Physics Reactor (ZPPR, originally the Zero Power Plutonium Reactor) in 1990 at the former ANL-West site in Idaho, which is now part of the Idaho National Laboratory (INL). Two additional critical assemblies, ZPR-6 and ZPR-9, operated at the ANL-East site in Illinois. A total of 128 fast reactor assemblies were constructed with these facilities [1]. The infrastructure and measurement capabilities are too expensive to be replicated in the modern era, making the integral database invaluable as the world pushes ahead with development of liquid metal cooled reactors.

  2. 22.39 Integration of Reactor Design, Operations, and Safety, Fall 2005

    E-Print Network [OSTI]

    Todreas, Neil E.

    This course integrates studies of reactor physics and engineering sciences into nuclear power plant design. Topics include materials issues in plant design and operations, aspects of thermal design, fuel depletion and ...

  3. PPPL3224, Preprint: February 1997, UC420, 426 Plasma Transport Control and SelfSustaining Fusion Reactor*

    E-Print Network [OSTI]

    to demonstrate this concept with D­T reactor­grade plasmas. For edge transport control, a method based an economical and environmentally sound fusion reactor prototype is the ultimate goal of controlled fusionPPPL­3224, Preprint: February 1997, UC­420, 426 Plasma Transport Control and Self­Sustaining Fusion

  4. PPPL-3224, Preprint: February 1997, UC-420, 426 Plasma Transport Control and Self-Sustaining Fusion Reactor*

    E-Print Network [OSTI]

    to demonstrate this concept with D-T reactor-grade plasmas. For edge transport control, a method based an economical and environmentally sound fusion reactor prototype is the ultimate goal of controlled fusionPPPL-3224, Preprint: February 1997, UC-420, 426 Plasma Transport Control and Self-Sustaining Fusion

  5. Modeling requirements for full-scope reactor simulators of fission-product transport during severe accidents

    SciTech Connect (OSTI)

    Ellison, P.G.; Monson, P.R. (Westinghouse Savannah River Co., Aiken, SC (United States)); Mitchell, H.A. (Concord Associates, Inc., Knoxville, TN (United States))

    1990-01-01T23:59:59.000Z

    This paper describes in the needs and requirements to properly and efficiently model fission product transport on full scope reactor simulators. Current LWR simulators can be easily adapted to model severe accident phenomena and the transport of radionuclides. Once adapted these simulators can be used as a training tool during operator training exercises for training on severe accident guidelines, for training on containment venting procedures, or as training tool during site wide emergency training exercises.

  6. Modeling requirements for full-scope reactor simulators of fission-product transport during severe accidents

    SciTech Connect (OSTI)

    Ellison, P.G.; Monson, P.R. [Westinghouse Savannah River Co., Aiken, SC (United States); Mitchell, H.A. [Concord Associates, Inc., Knoxville, TN (United States)

    1990-12-31T23:59:59.000Z

    This paper describes in the needs and requirements to properly and efficiently model fission product transport on full scope reactor simulators. Current LWR simulators can be easily adapted to model severe accident phenomena and the transport of radionuclides. Once adapted these simulators can be used as a training tool during operator training exercises for training on severe accident guidelines, for training on containment venting procedures, or as training tool during site wide emergency training exercises.

  7. Advances toward a transportable antineutrino detector system for reactor monitoring and safeguards

    SciTech Connect (OSTI)

    Reyna, D. [Sandia National Laboratories, Livermore, CA 94550 (United States); Bernstein, A. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Lund, J.; Kiff, S.; Cabrera-Palmer, B. [Sandia National Laboratories, Livermore, CA 94550 (United States); Bowden, N. S.; Dazeley, S.; Keefer, G. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States)

    2011-07-01T23:59:59.000Z

    Nuclear reactors have served as the neutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these very weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Our SNL/LLNL collaboration has demonstrated that such antineutrino based monitoring is feasible using a relatively small cubic meter scale liquid scintillator detector at tens of meters standoff from a commercial Pressurized Water Reactor (PWR). With little or no burden on the plant operator we have been able to remotely and automatically monitor the reactor operational status (on/off), power level, and fuel burnup. The initial detector was deployed in an underground gallery that lies directly under the containment dome of an operating PWR. The gallery is 25 meters from the reactor core center, is rarely accessed by plant personnel, and provides a muon-screening effect of some 20-30 meters of water equivalent earth and concrete overburden. Unfortunately, many reactor facilities do not contain an equivalent underground location. We have therefore attempted to construct a complete detector system which would be capable of operating in an aboveground location and could be transported to a reactor facility with relative ease. A standard 6-meter shipping container was used as our transportable laboratory - containing active and passive shielding components, the antineutrino detector and all electronics, as well as climate control systems. This aboveground system was deployed and tested at the San Onofre Nuclear Generating Station (SONGS) in southern California in 2010 and early 2011. We will first present an overview of the initial demonstrations of our below ground detector. Then we will describe the aboveground system and the technological developments of the two antineutrino detectors that were deployed. Finally, some preliminary results of our aboveground test will be shown. (authors)

  8. Integration of High-Temperature Gas-Cooled Reactors into Industrial Process Applications

    SciTech Connect (OSTI)

    Lee Nelson

    2011-09-01T23:59:59.000Z

    This report is a summary of analyses performed by the NGNP project to determine whether it is technically and economically feasible to integrate high temperature gas cooled reactor (HTGR) technology into industrial processes. To avoid an overly optimistic environmental and economic baseline for comparing nuclear integrated and conventional processes, a conservative approach was used for the assumptions and calculations.

  9. An integrated approach for the verification of fresh mixed oxide fuel (MOX) assemblies at light water reactor MOX recycle reactors

    SciTech Connect (OSTI)

    Menlove, Howard O [Los Alamos National Laboratory; Lee, Sang - Yoon [Los Alamos National Laboratory

    2009-01-01T23:59:59.000Z

    This paper presents an integrated approach for the verification of mixed oxide (MOX) fuel assemblies prior to their being loaded into the reactor. There is a coupling of the verification approach that starts at the fuel fabrication plant and stops with the transfer of the assemblies into the thermal reactor. The key measurement points are at the output of the fuel fabrication plant, the receipt at the reactor site, and the storage in the water pool as fresh fuel. The IAEA currently has the capability to measure the MOX fuel assemblies at the output of the fuel fabrication plants using a passive neutron coincidence counting systems of the passive neutron collar (PNCL) type. Also. at the MOX reactor pool, the underwater coincidence counter (UWCC) has been developed to measure the MOX assemblies in the water. The UWCC measurement requires that the fuel assembly be lifted about two meters up in the storage rack to avoid interference from the fuel that is stored in the rack. This paper presents a new method to verify the MOX fuel assemblies that are in the storage rack without the necessity of moving the fuel. The detector system is called the Underwater MOX Verification System (UMVS). The integration and relationship of the three measurements systems is described.

  10. Utilizing a Russian space nuclear reactor for a United States space mission: Systems integration issues

    SciTech Connect (OSTI)

    Reynolds, E.; Schaefer, E. [Johns Hopkins Univ., Laurel, MD (United States). Applied Physics Lab.; Polansky, G.; Lacy, J. [Phillips Lab., Albuquerque, NM (United States); Bocharov, A. [GDBMB, St. Petersburg (Russian Federation)

    1993-09-30T23:59:59.000Z

    The Nuclear Electric Propulsion Space Test Program (NEPSTP) has developed a cooperative relationship with several institutes of the former Soviet Union to evaluate Russian space hardware on a US spacecraft One component is the Topaz II Nuclear Power System; a built and flight qualified nuclear reactor that has yet to be tested in space. The access to the Topaz II reactor provides the NEPSTP with a rare opportunity; to conduct an early flight demonstration of nuclear electric propulsion at a relatively low cost. This opportunity, however, is not without challenges. Topaz II was designed to be compatible with Russian spacecraft and launch vehicles. It was manufactured and flight qualified by Russian techniques and standards and conforms to safety requirements of the former Soviet Union, not the United States. As it is desired to make minimal modifications to the Topaz II, integrating the reactor system with a United States spacecraft and launch vehicle presents an engineering challenge. This paper documents the lessons teamed regarding the integration of reactor based spacecraft and also some insight about integrating Russian hardware. It examines the planned integration flow along with specific reactor requirements that affect the spacecraft integration including American-Russian space system compatibility.

  11. INTEGRATION OF HIGH TEMPERATURE GAS REACTORS WITH IN SITU OIL SHALE RETORTING

    SciTech Connect (OSTI)

    Eric P. Robertson; Michael G. McKellar; Lee O. Nelson

    2011-05-01T23:59:59.000Z

    This paper evaluates the integration of a high-temperature gas-cooled reactor (HTGR) to an in situ oil shale retort operation producing 7950 m3/D (50,000 bbl/day). The large amount of heat required to pyrolyze the oil shale and produce oil would typically be provided by combustion of fossil fuels, but can also be delivered by an HTGR. Two cases were considered: a base case which includes no nuclear integration, and an HTGR-integrated case.

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

    SciTech Connect (OSTI)

    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.

    2003-11-12T23:59:59.000Z

    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.

  13. Packaging, Transportation, and Disposal Logistics for Large Radioactively Contaminated Reactor Decommissioning Components

    SciTech Connect (OSTI)

    Lewis, Mark S. [EnergySolutions: 140 Stoneridge Drive, Columbia, SC 29210 (United States)

    2008-01-15T23:59:59.000Z

    The packaging, transportation and disposal of large, retired reactor components from operating or decommissioning nuclear plants pose unique challenges from a technical as well as regulatory compliance standpoint. In addition to the routine considerations associated with any radioactive waste disposition activity, such as characterization, ALARA, and manifesting, the technical challenges for large radioactively contaminated components, such as access, segmentation, removal, packaging, rigging, lifting, mode of transportation, conveyance compatibility, and load securing require significant planning and execution. In addition, the current regulatory framework, domestically in Titles 49 and 10 and internationally in TS-R-1, does not lend itself to the transport of these large radioactively contaminated components, such as reactor vessels, steam generators, reactor pressure vessel heads, and pressurizers, without application for a special permit or arrangement. This paper addresses the methods of overcoming the technical and regulatory challenges. The challenges and disposition decisions do differ during decommissioning versus component replacement during an outage at an operating plant. During decommissioning, there is less concern about critical path for restart and more concern about volume reduction and waste minimization. Segmentation on-site is an available option during decommissioning, since labor and equipment will be readily available and decontamination activities are routine. The reactor building removal path is also of less concern and there are more rigging/lifting options available. Radionuclide assessment is necessary for transportation and disposal characterization. Characterization will dictate the packaging methodology, transportation mode, need for intermediate processing, and the disposal location or availability. Characterization will also assist in determining if the large component can be transported in full compliance with the transportation and disposal regulations and criteria or if special authorizations must be granted to transport and/or dispose. The U.S. DOT routinely issues special permits for large components where compliance with regulatory or acceptance criteria is impractical or impossible to meet. Transportation and disposal safety must be maintained even under special permits or authorizations. For example, if transported un-packaged, performance analysis must still be performed to assess the ability of the large component's outer steel shell to contain the internal radioactive contamination under normal transportation conditions and possibly incidence normal to transportation. The dimensions and weight of a large component must be considered when determining the possible modes of transportation (rail, water, or highway). At some locations, rail and/or barge access is unavailable. Many locations that once had an active rail spur to deliver new construction materials and components have let the spur deteriorate to the point that repair and upgrade of the spur is no longer economically feasible. Barge slips that have not been used since new plant construction require significant repair and/or dredging. Short on-site haul routes must be assessed for surface and subsurface conditions, as well as longer off-site routes. Off-site routes require clearance approvals from the regulatory authorities or, in the case of rail transport, the rail lines. Significant engineering planning and analysis must be performed during the pre-mobilization. In conclusion, the packaging, transportation, and disposal of large, oversized radioactively contaminated components removed during plant decommissioning is complex. However, over the last 15 years, a 100 or more components have been safely and compliantly packaged and transported for processing and/or disposal.

  14. Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

    SciTech Connect (OSTI)

    Mourao, Rogerio P.; Leite da Silva, Luiz [Centro de Desenvolvimento da Tecnologia Nuclear, Belo Horizonte (Brazil)] [Centro de Desenvolvimento da Tecnologia Nuclear, Belo Horizonte (Brazil); Miranda, Carlos A.; Mattar Neto, Miguel [Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo (Brazil)] [Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo (Brazil); Quintana, Jose F.A.; Saliba, Roberto O. [Comision Nacional de Energia Atomica, Bariloche (Argentina)] [Comision Nacional de Energia Atomica, Bariloche (Argentina); Novara, Oscar E. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)] [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

    2013-07-01T23:59:59.000Z

    Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away from reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)

  15. Investigation and design of a secure, transportable fluoride-salt-cooled high-temperature reactor (TFHR) for isolated locations

    E-Print Network [OSTI]

    Macdonald, Ruaridh (Ruaridh R.)

    2014-01-01T23:59:59.000Z

    In this work we describe a preliminary design for a transportable fluoride salt cooled high temperature reactor (TFHR) intended for use as a variable output heat and electricity source for off-grid locations. The goals of ...

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

    SciTech Connect (OSTI)

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

    2008-08-01T23:59:59.000Z

    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 materials needed to ensure their safe and reliable operation. The focus of this document will be the overall range of DOE's structural materials research activities being conducted to support VHTR development. By far, the largest portion of material's R&D supporting VHTR development is that being performed directly as part of the Next-Generation Nuclear Plant (NGNP) Project. Supplementary VHTR materials R&D being performed in the DOE program, including university and international research programs and that being performed under direct contracts with the American Society for Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, will also be described. Specific areas of high-priority materials research that will be needed to deploy the NGNP and provide a basis for subsequent VHTRs are described, including the following: (1) Graphite: (a) Extensive unirradiated materials characterization and assessment of irradiation effects on properties must be performed to qualify new grades of graphite for nuclear service, including thermo-physical and mechanical properties and their changes, statistical variations from billot-to-billot and lot-to-lot, creep, and especially, irradiation creep. (b) Predictive models, as well as codification of the requirements and design methods for graphite core supports, must be developed to provide a basis for licensing. (2) Ceramics: Both fibrous and load-bearing ceramics must be qualified for environmental and radiation service as insulating materials. (3) Ceramic Composites: Carbon-carbon and SiC-SiC composites must be qualified for specialized usage in selected high-temperature components, such as core stabilizers, control rods, and insulating covers and ducting. This will require development of component-specific designs and fabrication processes, materials characterization, assessment of environmental and irradiation effects, and establishment of codes and standards for materials testing and design requirements. (4) Pressure Vessel Steels: (a) Qualification of short-term, high-temperature properties of light water rea

  17. Light Water Reactor Sustainability Program Integrated Program Plan

    SciTech Connect (OSTI)

    McCarthy, Kathryn A. [INL; Busby, Jeremy [ORNL; Hallbert, Bruce [INL; Bragg-Sitton, Shannon [INL; Smith, Curtis [INL; Barnard, Cathy [INL

    2014-04-01T23:59:59.000Z

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

  18. Light Water Reactor Sustainability Program Integrated Program Plan

    SciTech Connect (OSTI)

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

    2012-01-01T23:59:59.000Z

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

  19. Light Water Reactor Sustainability Program Integrated Program Plan

    SciTech Connect (OSTI)

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

    2013-04-01T23:59:59.000Z

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

  20. NUHOWS - Storage and Transportation of Irradiated Reactor Components in Large Packages - 13439

    SciTech Connect (OSTI)

    Rae, Glen A. [Transnuclear, Inc., 7135 Minstrel Way, Columbia, MD 21045 (United States)] [Transnuclear, Inc., 7135 Minstrel Way, Columbia, MD 21045 (United States)

    2013-07-01T23:59:59.000Z

    Most irradiated reactor components (hardware such as Control Rod Blades, Fuel Channels, Poison Curtains, etc.) generated at reactors previously required significant processing for size reduction due to the available transportation casks not being physically capable of containing unprocessed material. As of July 1, 2008, disposal for this typical waste class (B and C) became inaccessible (for the major part of the nation) due to the Barnwell, SC disposal facility being closed to all but its three compact states (CT, NJ and SC). Currently in the United States, most facilities are storing their irradiated hardware on-site in the spent fuel pools. Until recently with the opening of the Waste Control Specialists' Texas disposal facility, utilities faced the challenges of spent fuel pool space and capacity management. However, even with WCS's disposal availability, the site currently has annual Curie limitations for disposal, which will continue to promote interim on-site storage until such time as disposal is available. In response, Transnuclear Inc., (TN) an AREVA company, proceeded with designing a new large Radioactive Waste Container (RWC) that can be used to package irradiated hardware without the need for significant processing. The design features of the RWC allows for intermittent loadings of the hardware for better packaging efficiency, higher packaging density, space savings and reduced cost. This RWC is also compatible with TN's on-site modular vault storage system. Once completely loaded, the RWC can be transported to an on-site storage facility, an off-site storage facility and/or an available disposal facility. To accommodate the transportation, TN has designed a large transportation cask, the MP197HB. As the original design was for transporting fuel, it contains the necessary shielding to allow for the transport of unprocessed irradiated reactor components, while significantly reducing the amount of irradiated hardware shipments required with the use of the new RWC. This paper provides information on the unique design features of the RWC, storage module vaults, MP197HB Transportation Cask and cost saving benefits of using the large RWC for packaging, storage, transport and disposal. (authors)

  1. THE INTEGRATION OF PROCESS HEAT APPLICATIONS TO HIGH TEMPERATURE GAS REACTORS

    SciTech Connect (OSTI)

    Michael G. McKellar

    2011-11-01T23:59:59.000Z

    A high temperature gas reactor, HTGR, can produce industrial process steam, high-temperature heat-transfer gases, and/or electricity. In conventional industrial processes, these products are generated by the combustion of fossil fuels such as coal and natural gas, resulting in significant emissions of greenhouse gases such as carbon dioxide. Heat or electricity produced in an HTGR could be used to supply process heat or electricity to conventional processes without generating any greenhouse gases. Process heat from a reactor needs to be transported by a gas to the industrial process. Two such gases were considered in this study: helium and steam. For this analysis, it was assumed that steam was delivered at 17 MPa and 540 C and helium was delivered at 7 MPa and at a variety of temperatures. The temperature of the gas returning from the industrial process and going to the HTGR must be within certain temperature ranges to maintain the correct reactor inlet temperature for a particular reactor outlet temperature. The returning gas may be below the reactor inlet temperature, ROT, but not above. The optimal return temperature produces the maximum process heat gas flow rate. For steam, the delivered pressure sets an optimal reactor outlet temperature based on the condensation temperature of the steam. ROTs greater than 769.7 C produce no additional advantage for the production of steam.

  2. Light Water Reactor Sustainability Program: Integrated Program Plan |

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn't YourTransport(FactDepartment ofLetter Report:40PMDepartmentPresentation fromR&D Roadmap

  3. Technology Evaluation and Integration Group: Center for Transportation Technologies and Systems

    SciTech Connect (OSTI)

    Not Available

    2008-08-01T23:59:59.000Z

    Fact sheet describes the specialized work done by NREL's Technology Evaluation and Integration Group in the Center for Transportation Technologies and Systems.

  4. Gain margin and phase margin analysis of a nuclear reactor control system with multiple transport lags

    SciTech Connect (OSTI)

    Chang, C.H. (Institute of Electronics, National Chiao-Tung Univ. (TW)); Han, K.W. (Chung-Shan Institute and National Chiao-Tung Univ., Hsinchu (TW))

    1989-08-01T23:59:59.000Z

    In this paper a method for finding the boundaries of constant gain margin and phase margin of control systems with transport lags and adjustable parameters is presented. The considered systems are first modified by adding a gain-phase margin tester, then the characteristic equations are formulated, and finally the stability equations are used to find the boundaries of constant gain margin and phase margin. The main advantage of the proposed method is to obtain complete information about the effects of adjustable parameters on gain margin and phase margin and their corresponding crossover frequencies. In order to show the usefulness of the proposed method a nuclear reactor control system with multiple transport lags is chosen as one of the examples.

  5. Transportation impact analysis for shipment of irradiated N-reactor fuel and associated materials

    SciTech Connect (OSTI)

    Daling, P.M.; Harris, M.S.

    1994-12-01T23:59:59.000Z

    An analysis of the radiological and nonradiological impacts of highway transportation of N-Reactor irradiated fuel (N-fuel) and associated materials is described in this report. N-fuel is proposed to be transported from its present locations in the 105-KE and 105-KW Basins, and possibly the PUREX Facility, to the 327 Building for characterization and testing. Each of these facilities is located on the Hanford Site, which is near Richland, Washington. The projected annual shipping quantity is 500 kgU/yr for 5 years for a total of 2500 kgU. It was assumed the irradiated fuel would be returned to the K- Basins following characterization, so the total amount of fuel shipped was assumed to be 5000 kgU. The shipping campaign may also include the transport and characterization of liquids, gases, and sludges from the storage basins, including fuel assembly and/or canister parts that may also be present in the basins. The impacts of transporting these other materials are bounded by the impacts of transporting 5000 kgU of N-fuel. This report was prepared to support an environmental assessment of the N-fuel characterization program. The RADTRAN 4 and GENII computer codes were used to evaluate the radiological impacts of the proposed shipping campaign. RADTRAN 4 was used to calculate the routine exposures and accident risks to workers and the general public from the N-fuel shipments. The GENII computer code was used to calculate the consequences of the maximum credible accident. The results indicate that the transportation of N-fuel in support of the characterization program should not cause excess radiological-induced latent cancer fatalities or traffic-related nonradiological accident fatalities. The consequences of the maximum credible accident are projected to be small and result in no excess latent cancer fatalities.

  6. The Integral Fast Reactor concept: Today's hope for tomorrow's electrical energy needs

    SciTech Connect (OSTI)

    Dwight, C.C.; Phipps, R.D.

    1989-01-01T23:59:59.000Z

    Acid rain and the greenhouse effect are getting more attention as their impacts on the environment become evident around the world. Substantial evidence indicates that fossil fuel combustion for electrical energy production activities is a key cause of those problems. A change in electrical energy production policy is essential to a stable, healthy environment. That change is inevitable, it's just a matter of when and at what cost. Vision now, instead of reaction later, both in technological development and public perception, will help to limit the costs of change. The Integral Fast Reactor (IFR) is a visionary concept developed by Argonne National Laboratory that involves electrical energy production through fissioning of heavy metals by fast neutrons in a reactor cooled by liquid sodium. Physical characteristics of the coolant and fuel give the reactor impressive characteristics of inherent and passive safety. Spent fuel is pyrochemically reprocessed and returned to the reactor in the IFR's closed fuel cycle. Advantages in waste management are realized, and the reactor has the potential for breeding, i.e., producing as much or more fuel than it uses. This paper describes the IFR concept and shows how it is today's hope for tomorrow's electrical energy needs. 14 refs., 1 fig., 1 tab.

  7. The Innovations, Technology and Waste Management Approaches to Safely Package and Transport the World's First Radioactive Fusion Research Reactor for Burial

    SciTech Connect (OSTI)

    Keith Rule; Erik Perry; Jim Chrzanowski; Mike Viola; Ron Strykowsky

    2003-09-15T23:59:59.000Z

    Original estimates stated that the amount of radioactive waste that will be generated during the dismantling of the Tokamak Fusion Test Reactor will approach two million kilograms with an associated volume of 2,500 cubic meters. The materials were activated by 14 MeV neutrons and were highly contaminated with tritium, which present unique challenges to maintain integrity during packaging and transportation. In addition, the majority of this material is stainless steel and copper structural metal that were specifically designed and manufactured for this one-of-a-kind fusion research reactor. This provided further complexity in planning and managing the waste. We will discuss the engineering concepts, innovative practices, and technologies that were utilized to size reduce, stabilize, and package the many unique and complex components of this reactor. This waste was packaged and shipped in many different configurations and methods according to the transportation regulations and disposal facility requirements. For this particular project, we were able to utilize two separate disposal facilities for burial. This paper will conclude with a complete summary of the actual results of the waste management costs, volumes, and best practices that were developed from this groundbreaking and successful project.

  8. Application-specific integrated circuit design for a typical pressurized water reactor pressure channel trip

    SciTech Connect (OSTI)

    Battle, R.E.; Manges, W.W.; Emery, M.S.; Vendermolen, R.I. [Oak Ridge National Lab., TN (United States); Bhatt, S. [Electric Power Research Inst., Palo Alto, CA (United States)

    1994-03-01T23:59:59.000Z

    This article discusses the use of application-specific integrated circuits (ASICs) in nuclear plant safety systems. ASICs have certain advantages over software-based systems because they can be simple enough to be thoroughly tested, and they can be tailored to replace existing equipment. An architecture to replace a pressurized water reactor pressure channel trip is presented. Methods of implementing digital algorithms are also discussed.

  9. Improved parallel solution techniques for the integral transport matrix method

    SciTech Connect (OSTI)

    Zerr, Robert J [Los Alamos National Laboratory; Azmy, Yousry Y [NORTH CAROLINA STATE UNIV.

    2010-11-23T23:59:59.000Z

    Alternative solution strategies to the parallel block Jacobi (PBJ) method for the solution of the global problem with the integral transport matrix method operators have been designed and tested. The most straightforward improvement to the Jacobi iterative method is the Gauss-Seidel alternative. The parallel red-black Gauss-Seidel (PGS) algorithm can improve on the number of iterations and reduce work per iteration by applying an alternating red-black color-set to the subdomains and assigning multiple sub-domains per processor. A parallel GMRES(m) method was implemented as an alternative to stationary iterations. Computational results show that the PGS method can improve on the PBJ method execution by up to {approx}50% when eight sub-domains per processor are used. However, compared to traditional source iterations with diffusion synthetic acceleration, it is still approximately an order of magnitude slower. The best-performing case are opticaUy thick because sub-domains decouple, yielding faster convergence. Further tests revealed that 64 sub-domains per processor was the best performing level of sub-domain division. An acceleration technique that improves the convergence rate would greatly improve the ITMM. The GMRES(m) method with a diagonal block preconditioner consumes approximately the same time as the PBJ solver but could be improved by an as yet undeveloped, more efficient preconditioner.

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

    SciTech Connect (OSTI)

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

    2004-07-01T23:59:59.000Z

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

  11. Preliminary neutronics investigation of the delayed neutron nondestructive assay of an integral fast reactor waste canister

    SciTech Connect (OSTI)

    Wu, C.W.; Henderson, D.L. (Univ. of Wisconsin, Madison, WI (United States). Dept. of Nuclear Engineering and Engineering Physics); Bennett, E.F. (Argonne National Lab., IL (United States))

    1994-11-01T23:59:59.000Z

    An innovative liquid-metal reactor, the Integral Fast Reactor (IFR), is being developed at Argonne National Laboratory. One characteristic of the IFR is the fuel cycle closure. Fissile material bred and fissionable material produced in the reactor are recycled back into the reactor. Waste generated during fuel reprocessing will be packaged into special waste canisters and will be shipped to a repository for final disposal. Prior to its removal from the facility, a measurements of the fissile content will be necessary as a part of an overall fissile material inventory accountability system. A particular form of nondestructive assay called delayed neutron nondestructive assay (DNNDA) is being developed to assist in the establishment of an accountability system. A preliminary neutronics investigation for the current DNNDA has been made to assist and verify the characteristics of the design from a neutronic aspect. A 10[sup 11] n/s, 14-MeV neutron source would provide adequate counting statistics for fissile material at the milligram to gram level.

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

    SciTech Connect (OSTI)

    Hissam, D. Andy; Stewart, Eric [National Aeronautics and Space Administration, Marshall Space Flight Center, ER34, Huntsville, AL 35812 (United States)

    2006-07-01T23:59:59.000Z

    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, Brayton turbine-generator. Since these two systems were developed independently of one another, this evaluation sought to determine if they could be operated 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: N{sub 2}, He/Ar, and He/Xe. Other technical issues, such as electrical breakdown in the generator and the operation of the Brayton foil bearings using various gas mixtures, were also investigated. (authors)

  13. Use of principal components analysis and three-dimensional atmospheric-transport models for reactor-consequence evaluation

    SciTech Connect (OSTI)

    Gudiksen, P.H.; Walton, J.J.; Alpert, D.J.; Johnson, J.D.

    1982-01-01T23:59:59.000Z

    This work explores the use of principal components analysis coupled to three-dimensional atmospheric transport and dispersion models for evaluating the environmental consequences of reactor accidents. This permits the inclusion of meteorological data from multiple sites and the effects of topography in the consequence evaluation; features not normally included in such analyses. The technique identifies prevailing regional wind patterns and their frequencies for use in the transport and dispersion calculations. Analysis of a hypothetical accident scenario involving a release of radioactivity from a reactor situated in a river valley indicated the technique is quite useful whenever recurring wind patterns exist, as is often the case in complex terrain situations. Considerable differences were revealed in a comparison with results obtained from a more conventional Gaussian plume model using only the reactor site meteorology and no topographic effects.

  14. A risk characterization of safety research areas for Integral Fast Reactor program planning

    SciTech Connect (OSTI)

    Mueller, C.J.; Cahalan, J.E.; Hill, D.J.; Kramer, J.M.; Marchaterre, J.F.; Pedersen, D.R.; Sevy, R.H.; Tilbrook, R.W.; Wei, T.Y.; Wright, A.E.

    1988-01-01T23:59:59.000Z

    This paper characterizes the areas of Integral Fast Reactor (IFR) safety research in terms of their importance in addressing the risk of core disruption sequences for innovative designs. Such sequences have traditionally been determined to constitute the primary risk to public health and safety. All core disruption sequences are folded into four fault categories: classic unprotected (unscrammed) events; loss of decay heat; local fault propagation; and failure of critical reactor structures. Event trees are used to describe these sequences and the areas in the IFR Safety and related Base Technology research programs are discussed with respect to their relevance in addressing the key issues in preventing or delimiting core disruptive sequences. Thus a measure of potential for risk reduction is obtained for guidance in establishing research priorites.

  15. The Integration of a Structural Water Gas Shift Catalyst with a Vanadium Alloy Hydrogen Transport Device

    SciTech Connect (OSTI)

    Barton, Thomas; Argyle, Morris; Popa, Tiberiu

    2009-06-30T23:59:59.000Z

    This project is in response to a requirement for a system that combines water gas shift technology with separation technology for coal derived synthesis gas. The justification of such a system would be improved efficiency for the overall hydrogen production. By removing hydrogen from the synthesis gas stream, the water gas shift equilibrium would force more carbon monoxide to carbon dioxide and maximize the total hydrogen produced. Additional benefit would derive from the reduction in capital cost of plant by the removal of one step in the process by integrating water gas shift with the membrane separation device. The answer turns out to be that the integration of hydrogen separation and water gas shift catalysis is possible and desirable. There are no significant roadblocks to that combination of technologies. The problem becomes one of design and selection of materials to optimize, or at least maximize performance of the two integrated steps. A goal of the project was to investigate the effects of alloying elements on the performance of vanadium membranes with respect to hydrogen flux and fabricability. Vanadium was chosen as a compromise between performance and cost. It is clear that the vanadium alloys for this application can be produced, but the approach is not simple and the results inconsistent. For any future contracts, large single batches of alloy would be obtained and rolled with larger facilities to produce the most consistent thin foils possible. Brazing was identified as a very likely choice for sealing the membranes to structural components. As alloying was beneficial to hydrogen transport, it became important to identify where those alloying elements might be detrimental to brazing. Cataloging positive and negative alloying effects was a significant portion of the initial project work on vanadium alloying. A water gas shift catalyst with ceramic like structural characteristics was the second large goal of the project. Alumina was added as a component of conventional high temperature water gas shift iron oxide based catalysts. The catalysts contained Fe-Al-Cr-Cu-O and were synthesized by co-precipitation. A series of catalysts were prepared with 5 to 50 wt% Al2O3, with 8 wt% Cr2O3, 4 wt% CuO, and the balance Fe2O3. All of the catalysts were compared to a reference WGS catalyst (88 wt% FeOx, 8 wt% Cr2O3, and 4 wt% CuO) with no alumina. Alumina addition to conventional high temperature water gas shift catalysts at concentrations of approximately 15 wt% increased CO conversion rates and increase thermal stability. A series of high temperature water gas shift catalysts containing iron, chromia, and copper oxides were prepared with small amounts of added ceria in the system Fe-Cr-Cu-Ce-O. The catalysts were also tested kinetically under WGS conditions. 2-4 wt% ceria addition (at the expense of the iron oxide content) resulted in increased reaction rates (from 22-32% higher) compared to the reference catalyst. The project goal of a 10,000 liter per day WGS-membrane reactor was achieved by a device operating on coal derived syngas containing significant amounts of carbon monoxide and hydrogen sulfide. The membrane flux was equivalent to 52 scfh/ft2 based on a 600 psi syngas inlet pressure and corresponded to membranes costing $191 per square foot. Over 40 hours of iv exposure time to syngas has been achieved for a double membrane reactor. Two modules of the Chart reactor were tested under coal syngas for over 75 hours with a single module tested for 50 hours. The permeance values for the Chart membranes were similar to the REB reactor though total flux was reduced due to significantly thicker membranes. Overall testing of membrane reactors on coal derived syngas was over 115 hours for all reactors tested. Testing of the REB double membrane device exceeded 40 hours. Performance of the double membrane reactor has been similar to the results for the single reactor with good maintenance of flux even after these long exposures to hydrogen sulfide. Of special interest is that the flux is highest at the start of each e

  16. Reactor pressure vessel integrity research at the Oak Ridge National Laboratory

    SciTech Connect (OSTI)

    Corwin, W.R.; Pennell, W.E.; Pace, J.V.

    1995-12-31T23:59:59.000Z

    Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. The RPV is the only key safety-related component of the plant for which a duplicate or redundant backup system does not exist. It is therefore imperative to understand and be able to predict the integrity inherent in the RPV. For this reason, the U.S. Nuclear Regulatory Commission has established the related research programs at ORNL described herein to provide for the development and confirmation of the methods used for: (1) establishing the irradiation exposure conditions within the RPV in the Embrittlement Data Base and Dosimetry Evaluation Program, (2) assessing the effects of irradiation on the RPV materials in the Heavy-Section Steel Irradiation Program, and (3) developing overall structural and fracture analyses of RPVs in the Heavy-Section Steel Technology Program.

  17. Integration of renewable energy into the transport and electricity sectors through V2G

    E-Print Network [OSTI]

    Firestone, Jeremy

    Integration of renewable energy into the transport and electricity sectors through V2G Henrik Lund Renewable energy Wind powerQ1 a b s t r a c t Large-scale sustainable energy systems will be necessary replace oil in the transportation sector, and (2) since today's inexpensive and abundant renewable energy

  18. Instrumentation Needs for Integral Primary System Reactors (IPSRs) - Task 1 Final Report

    SciTech Connect (OSTI)

    Gary D. Storrick; Bojan Petrovic; Luca Oriani; Lawrence E. Conway; Diego Conti

    2005-09-30T23:59:59.000Z

    This report presents the results of the Westinghouse work performed under Task 1 of this Financial Assistance Award and satisfies a Level 2 Milestone for the project. While most of the signals required for control of IPSRs are typical of other PWRs, the integral configuration poses some new challenges in the design or deployment of the sensors/instrumentation and, in some cases, requires completely new approaches. In response to this consideration, the overall objective of Task 1 was to establish the instrumentation needs for integral reactors, provide a review of the existing solutions where available, and, identify research and development needs to be addressed to enable successful deployment of IPSRs. The starting point for this study was to review and synthesize general characteristics of integral reactors, and then to focus on a specific design. Due to the maturity of its design and availability of design information to Westinghouse, IRIS (International Reactor Innovative and Secure) was selected for this purpose. The report is organized as follows. Section 1 is an overview. Section 2 provides background information on several representative IPSRs, including IRIS. A review of the IRIS safety features and its protection and control systems is used as a mechanism to ensure that all critical safety-related instrumentation needs are addressed in this study. Additionally, IRIS systems are compared against those of current advanced PWRs. The scope of this study is then limited to those systems where differences exist, since, otherwise, the current technology already provides an acceptable solution. Section 3 provides a detailed discussion on instrumentation needs for the representative IPSR (IRIS) with detailed qualitative and quantitative requirements summarized in the exhaustive table included as Appendix A. Section 3 also provides an evaluation of the current technology and the instrumentation used for measurement of required parameters in current PWRs. Section 4 examines those instrumentation/measurement needs where differences between IRIS and present PWRs exist and the current PWR implementation cannot be directly employed, and identifies two subcategories. In the first group, resolution can be readily identified, and is essentially an engineering solution (for example, modification of an existing approach, adaptation of existing instrument etc.). The second group presents true technological challenges as it may require new technology development. I n these cases, high level functional requirements have been identified together with relevant technical considerations to guide future development activities.

  19. Integrated transport and renewable energy systems B. V. Mathiesen*

    E-Print Network [OSTI]

    on second generation biofuel production for the substitution of oil products [4]. Worldwide, the focus it. Biomass is a limited resource and it is important to avoid effecting the production, transport gains increasing international attention due to its large oil dependency of app. 95 per cent

  20. Status report on the Small Secure Transportable Autonomous Reactor (SSTAR) /Lead-cooled Fast Reactor (LFR) and supporting research and development.

    SciTech Connect (OSTI)

    Sienicki, J. J.; Moisseytsev, A.; Yang, W. S.; Wade, D. C.; Nikiforova, A.; Hanania, P.; Ryu, H. J.; Kulesza, K. P.; Kim, S. J.; Halsey, W. G.; Smith, C. F.; Brown, N. W.; Greenspan, E.; de Caro, M.; Li, N.; Hosemann, P.; Zhang, J.; Yu, H.; Nuclear Engineering Division; LLNL; LANL; Massachusetts Inst. of Tech.; Ecole des Mines de Paris; Oregon State Univ.; Univ.of California at Berkley

    2008-06-23T23:59:59.000Z

    This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, fast reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been made at ANL to the pre-conceptual design of both the reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the reactor core power adjusts itself to equal the heat removal from the reactor system to the power converter through the large reactivity feedback of the fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled fast reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same neutronics computer codes and methodologies applied to SSTAR. Another section of the report details the SSTAR safety design approach which is based upon defense-in-depth providing multiple levels of protection against the release of radioactive materials and how the inherent safety features of the lead coolant, nitride fuel, fast neutron spectrum core, pool vessel configuration, natural circulation, and containment meet or exceed the requirements for each level of protection. The report also includes recent results of a systematic analysis by LANL of data on corrosion of candidate cladding and structural material alloys of interest to SSTAR by LBE and Pb coolants; the data were taken from a new database on corrosion by liquid metal coolants created at LANL. The analysis methodology that considers penetration of an oxidation front into the alloy and dissolution of the trailing edge of the oxide into the coolant enables the long-term corrosion rate to be extracted from shorter-term corrosion data thereby enabling an evaluation of alloy performance over long core lifetimes (e.g., 30 years) that has heretofore not been possible. A number of candidate alloy specimens with special treatments or coatings which might enhance corrosion resistance at the temperatures at which SSTAR would operate were analyzed following testing in the DELTA loop at LANL including steels that were treated by laser peening at LLNL; laser peening is an approach that alters the oxide-metal bonds which could potentially improve corrosion resistance. LLNL is also carrying out Multi-Scale Modeling of the Fe-Cr system with the goal of assisting in the development of cladding and structural materials having greater resistance to irradiation.

  1. Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

    DOE Patents [OSTI]

    Schreiber, R.B.; Fero, A.H.; Sejvar, J.

    1997-12-16T23:59:59.000Z

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor. 8 figs.

  2. Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

    DOE Patents [OSTI]

    Schreiber, Roger B. (Penn Twp., PA); Fero, Arnold H. (New Kensington, PA); Sejvar, James (Murrysville, PA)

    1997-01-01T23:59:59.000Z

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor.

  3. Transportation and Stationary Power Integration Workshop | Department of

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page onYouTube YouTube Note: Since the.pdfBreakingMayDepartmentTest forTechnologies » Transportation TechnologiesEnergy

  4. Integration of Nontraditional Isotopic Systems Into Reaction-Transport

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn't YourTransport(FactDepartment of EnergyIndustry ResearchPerspective | DepartmentModels of EGS

  5. Comparison of structural integrity of casks for spent-fuel transportation

    SciTech Connect (OSTI)

    Eifert, E.J.; Burger, C.S.; Herger, A.S. [Univ. of New Mexico, Albuquerque, NM (United States); Ammerman, D.J.; McConnell, P. [Sandia National Laboratories, Albuquerque, NM (United States)

    1995-12-31T23:59:59.000Z

    This paper presents the results of a series of finite element analyses comparing the structural integrity of two shipping casks for transportation of high-level waste (HLW). The objective of this project is to assess the advisability of utilizing ductile iron (DI) for type-B transport cask construction by investigating its structural response under severe loading conditions. This response is compared to that of a stainless steel (SS) cask under comparable loading conditions.

  6. International Atomic Energy Agency (IAEA) Coordinated Research Projects on Structural Integrity of Reactor Pressure Vessels

    SciTech Connect (OSTI)

    Server, W. L. [ATI Consulting, Pinehurst, NC; Nanstad, Randy K [ORNL

    2009-01-01T23:59:59.000Z

    The International Atomic Energy Agency (IAEA) has conducted a series of Coordinated Research Projects (CRPs) that have focused on irradiated reactor pressure vessel (RPV) steel fracture toughness properties and approaches for assuring structural integrity of RPVs throughout operating life. A series of nine CRPs have been sponsored by the IAEA, starting in the early 1970s, focused on neutron radiation effects on RPV steels. The purpose of the CRPs was to develop comparisons and correlations to test the uniformity of irradiated results through coordinated international research studies and data sharing. Consideration of dose rate effects, effects of alloying (nickel, manganese, silicon, etc.) and residual elements (eg., copper and phosphorus), and drop in upper shelf toughness are also important for assessing neutron embrittlement effects. The ultimate use of embrittlement understanding is assuring structural integrity of the RPV under current and future operation and accident conditions. Material fracture toughness is the key ingredient needed for this assessment, and many of the CRPs have focused on measurement and application of irradiated fracture toughness. This paper presents an overview of the progress made since the inception of the CRPs in the early 1970s. The chronology and importance of each CRP have been reviewed and put into context for continued and long-term safe operation of RPVs.

  7. The influence of selected containment structures on debris dispersal and transport following high pressure melt ejection from the reactor vessel

    SciTech Connect (OSTI)

    Pilch, M.; Tarbell, W.W.; Brockmann, J.E.

    1988-09-01T23:59:59.000Z

    High pressure expulsion of molten core debris from the reactor pressure vessel may result in dispersal of the debris from the reactor cavity. In most plants, the cavity exits into the containment such that the debris impinges on structures. Retention of the debris on the structures may affect the further transport of the debris throughout the containment. Two tests were done with scaled structural shapes placed at the exit of 1:10 linear scale models of the Zion cavity. The results show that the debris does not adhere significantly to structures. The lack of retention is attributed to splashing from the surface and reentrainment in the gas flowing over the surface. These processes are shown to be applicable to reactor scale. A third experiment was done to simulate the annular gap between the reactor vessel and cavity wall. Debris collection showed that the fraction of debris exiting through the gap was greater than the gap-to-total flow area ratio. Film records indicate that dispersal was primarily by entrainment of the molten debris in the cavity. 29 refs., 36 figs., 11 tabs.

  8. Integrating Safety, Operations, Security, and Safeguards (ISOSS) into the design of small modular reactors : a handbook.

    SciTech Connect (OSTI)

    Middleton, Bobby D.; Mendez, Carmen Margarita [Sociotecnia Solutions] [Sociotecnia Solutions

    2013-10-01T23:59:59.000Z

    The existing regulatory environment for nuclear reactors impacts both the facility design and the cost of operations once the facility is built. Delaying the consideration of regulatory requirements until late in the facility design - or worse, until after construction has begun - can result in costly retrofitting as well as increased operational costs to fulfill safety, security, safeguards, and emergency readiness requirements. Considering the scale and scope, as well as the latest design trends in the next generation of nuclear facilities, there is an opportunity to evaluate the regulatory requirements and optimize the design process for Small Modular Reactors (SMRs), as compared to current Light Water Reactors (LWRs). To this end, Sandia has embarked on an initiative to evaluate the interactions of regulations and operations as an approach to optimizing the design of SMR facilities, supporting operational efficiencies, as well as regulatory requirements. The early stages of this initiative consider two focus areas. The first focus area, reported by LaChance, et al. (2007), identifies the regulatory requirements established for the current fleet of LWR facilities regarding Safety, Security, Operations, Safeguards, and Emergency Planning, and evaluates the technical bases for these requirements. The second focus area, developed in this report, documents the foundations for an innovative approach that supports a design framework for SMR facilities that incorporates the regulatory environment, as well as the continued operation of the facility, into the early design stages, eliminating the need for costly retrofitting and additional operating personnel to fulfill regulatory requirements. The work considers a technique known as Integrated Safety, Operations, Security and Safeguards (ISOSS) (Darby, et al., 2007). In coordination with the best practices of industrial operations, the goal of this effort is to develop a design framework that outlines how ISOSS requirements can be incorporated into the pre-conceptual through early facility design stages, seeking a cost-effective design that meets both operational efficiencies and the regulatory environment. The larger scope of the project, i.e., in future stages, includes the identification of potentially conflicting requirements identified by the ISOSS framework, including an analysis of how regulatory requirements may be changed to account for the intrinsic features of SMRs.

  9. Two-dimensional neutronic analysis of the TAMU Nuclear Science Center Reactor using transport and diffusion theory based codes 

    E-Print Network [OSTI]

    Davis, John Wesley

    1988-01-01T23:59:59.000Z

    is not normally treated explicitly so that the critical state of' a reactor system can be analyzed. The time-independent homogeneous form of the Boltzmann transport equation for a multiplying system is written [fl . V + Zt(r", E)]g(r, E, 0) = dE' dfI'Z, (F, E...' ~ E, Q' fI)tf (r. E', Q') / i I Equation (I) states that the losses due to leakage and collisions equals the source of neutrons from scattering and fissions, at a point r", energy E, and in direction fl, per unit volume, energy, and solid angle...

  10. ENGINEERING DEVELOPMENT OF CERAMIC MEMBRANE REACTOR SYSTEM FOR CONVERTING NATURAL GAS TO HYDROGEN AND SYNTHESIS GAS FOR LIQUID TRANSPORTATION FUELS

    SciTech Connect (OSTI)

    NONE

    1998-08-01T23:59:59.000Z

    The objective of this contract is to research, develop and demonstrate a novel ceramic membrane reactor system for the low-cost conversion of natural gas to synthesis gas and hydrogen for liquid transportation fuels: the ITM Syngas process. Through an eight-year, three-phase program, the technology will be developed and scaled up to obtain the technical, engineering, operating and economic data necessary for the final step to full commercialization of the Gas-to-Liquids (GTL) conversion technology. This report is a summary of activities through July 1999.

  11. Engineering development of ceramic membrane reactor system for converting natural gas to hydrogen and synthesis gas for liquid transportation fuels

    SciTech Connect (OSTI)

    NONE

    1998-07-01T23:59:59.000Z

    The objective of this contract is to research, develop and demonstrate a novel ceramic membrane reactor system for the low-cost conversion of natural gas to synthesis gas and hydrogen for liquid transportation fuels: the ITM Syngas process. Through an eight-year, three-phase program, the technology will be developed and scaled up to obtain the technical, engineering, operating and economic data necessary for the final step to full commercialization of the Gas-to-Liquids (GTL) conversion technology. This report is a summary of activities through June 1998.

  12. ENGINEERING DEVELOPMENT OF CERAMIC MEMBRANE REACTOR SYSTEM FOR CONVERTING NATURAL GAS TO HYDROGEN AND SYNTHESIS GAS FOR LIQUID TRANSPORTATION FUELS

    SciTech Connect (OSTI)

    NONE

    1999-12-01T23:59:59.000Z

    The objective of this contract is to research, develop and demonstrate a novel ceramic membrane reactor system for the low-cost conversion of natural gas to synthesis gas and hydrogen for liquid transportation fuels: the ITM Syngas process. Through an eight-year, three-phase program, the technology will be developed and scaled up to obtain the technical, engineering, operating and economic data necessary for the final step to full commercialization of the Gas-to-Liquids (GTL) conversion technology. This report is a summary of activities through November 1999.

  13. ENGINEERING DEVELOPMENT OF CERAMIC MEMBRANE REACTOR SYSTEM FOR CONVERTING NATURAL GAS TO HYDROGEN AND SYNTHESIS GAS FOR LIQUID TRANSPORTATION FUELS

    SciTech Connect (OSTI)

    NONE

    1999-03-01T23:59:59.000Z

    The objective of this contract is to research, develop and demonstrate a novel ceramic membrane reactor system for the low-cost conversion of natural gas to synthesis gas and hydrogen for liquid transportation fuels: the ITM Syngas process. Through an eight-year, three-phase program, the technology will be developed and scaled up to obtain the technical, engineering, operating and economic data necessary for the final step to full commercialization of the Gas-to-Liquids (GTL) conversion technology. This report is a summary of activities through February 1999.

  14. Engineering development of ceramic membrane reactor system for converting natural gas to hydrogen and synthesis gas for liquid transportation fuels

    SciTech Connect (OSTI)

    NONE

    1998-05-01T23:59:59.000Z

    The objective of this contract is to research, develop and demonstrate a novel ceramic membrane reactor system for the low-cost conversion of natural gas to synthesis gas and hydrogen for liquid transportation fuels: the ITM Syngas process. Through an eight-year, three-phase program, the technology will be developed and scaled up to obtain the technical, engineering, operating and economic data necessary for the final step to full commercialization of the Gas-to-Liquids (GTL) conversion technology. This report is a summary of activities through April 1998.

  15. ENGINEERING DEVELOPMENT OF CERAMIC MEMBRANE REACTOR SYSTEM FOR CONVERTING NATURAL GAS TO HYDROGEN AND SYNTHESIS GAS FOR LIQUID TRANSPORTATION FUELS

    SciTech Connect (OSTI)

    NONE

    1999-10-01T23:59:59.000Z

    The objective of this contract is to research, develop and demonstrate a novel ceramic membrane reactor system for the low-cost conversion of natural gas to synthesis gas and hydrogen for liquid transportation fuels: the ITM Syngas process. Through an eight-year, three-phase program, the technology will be developed and scaled up to obtain the technical, engineering, operating and economic data necessary for the final step to full commercialization of the Gas-to-Liquids (GTL) conversion technology. This report is a summary of activities through September 1999.

  16. ENGINEERING DEVELOPMENT OF CERAMIC MEMBRANE REACTOR SYSTEM FOR CONVERTING NATURAL GAS TO HYDROGEN AND SYNTHESIS GAS FOR LIQUID TRANSPORTATION FUELS

    SciTech Connect (OSTI)

    NONE

    2000-02-01T23:59:59.000Z

    The objective of this contract is to research, develop and demonstrate a novel ceramic membrane reactor system for the low-cost conversion of natural gas to synthesis gas and hydrogen for liquid transportation fuels: the ITM Syngas process. Through an eight-year, three-phase program, the technology will be developed and scaled up to obtain the technical, engineering, operating and economic data necessary for the final step to full commercialization of the Gas-to-Liquids (GTL) conversion technology. This report is a summary of activities through January 2000.

  17. ENGINEERING DEVELOPMENT OF CERAMIC MEMBRANE REACTOR SYSTEM FOR CONVERTING NATURAL GAS TO HYDROGEN AND SYNTHESIS GAS FOR LIQUID TRANSPORTATION FUELS

    SciTech Connect (OSTI)

    NONE

    2000-01-01T23:59:59.000Z

    The objective of this contract is to research, develop and demonstrate a novel ceramic membrane reactor system for the low-cost conversion of natural gas to synthesis gas and hydrogen for liquid transportation fuels: the ITM Syngas process. Through an eight-year, three-phase program, the technology will be developed and scaled up to obtain the technical, engineering, operating and economic data necessary for the final step to full commercialization of the Gas-to-Liquids (GTL) conversion technology. This report is a summary of activities through December 1999.

  18. ENGINEERING DEVELOPMENT OF CERAMIC MEMBRANE REACTOR SYSTEM FOR CONVERTING NATURAL GAS TO HYDROGEN AND SYNTHESIS GAS FOR LIQUID TRANSPORTATION FUELS

    SciTech Connect (OSTI)

    NONE

    1999-11-01T23:59:59.000Z

    The objective of this contract is to research, develop and demonstrate a novel ceramic membrane reactor system for the low-cost conversion of natural gas to synthesis gas and hydrogen for liquid transportation fuels: the ITM Syngas process. Through an eight-year, three-phase program, the technology will be developed and scaled up to obtain the technical, engineering, operating and economic data necessary for the final step to full commercialization of the Gas-to-Liquids (GTL) conversion technology. This report is a summary of activities through October 1999.

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

    SciTech Connect (OSTI)

    Not Available

    1992-04-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Ehud Greenspan

    2003-10-31T23:59:59.000Z

    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.

  1. Experimental characterization of an Ion Transport Membrane (ITM) reactor for methane oxyfuel combustion

    E-Print Network [OSTI]

    Apo, Daniel Jolomi

    2012-01-01T23:59:59.000Z

    Ion Transport Membranes (ITM) which conduct both electrons and oxygen ions have been investigated experimentally for oxygen separation and fuel (mostly methane) conversion purposes over the last three decades. The fuel ...

  2. Thermal hydraulic performance analysis of a small integral pressurized water reactor core

    E-Print Network [OSTI]

    Blair, Stuart R. (Stuart Ryan), 1972-

    2003-01-01T23:59:59.000Z

    A thermal hydraulic analysis of the International Reactor Innovative and Secure (IRIS) core has been performed. Thermal margins for steady state and a selection of Loss Of Flow Accidents have been assessed using three ...

  3. Economic Rationale for Safety Investment in Integrated Gasification Combined-Cycle Gas Turbine Membrane Reactor Modules

    E-Print Network [OSTI]

    Koc, Reyyan; Kazantzis, Nikolaos K.; Nuttall, William J.; Ma, Yi Hua

    2012-05-09T23:59:59.000Z

    . Please notice that after the condensation of steam and given the fact that CO2 is at a high pressure (~25 atm), a significant reduction in the compression costs associated with the operation of the sequestration units downstream... : Membrane Reactor HRSG: Heat Recovery Steam Generator EPRG WP 1211   13    The feed specifications, reaction conditions and permeation properties used in the isothermal membrane reactor model are listed in Table 1. In particular, a...

  4. Current status of the Run-Beyond-Cladding Breach (RBCB) tests for the Integral Fast Reactor (IFR). Metallic Fuels Program

    SciTech Connect (OSTI)

    Batte, G.L.; Pahl, R.G. [Argonne National Lab., Idaho Falls, ID (United States); Hofman, G.L. [Argonne National Lab., IL (United States)

    1993-09-01T23:59:59.000Z

    This paper describes the results from the Integral Fast Reactor (IFR) metallic fuel Run-Beyond-Cladding-Breach (RBCB) experiments conducted in the Experimental Breeder Reactor II (EBR-II). Included in the report are scoping test results and the data collected from the prototypical tests as well as the exam results and discussion from a naturally occurring breach of one of the lead IFR fuel tests. All results showed a characteristic delayed neutron and fission gas release pattern that readily allows for identification and evaluation of cladding breach events. Also, cladding breaches are very small and do not propagate during extensive post breach operation. Loss of fuel from breached cladding was found to be insignificant. The paper will conclude with a brief description of future RBCB experiments planned for irradiation in EBR-II.

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

    SciTech Connect (OSTI)

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

    2005-12-01T23:59:59.000Z

    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.

  6. Coupled Reactor Kinetics and Heat Transfer Model for Heat Pipe Cooled Reactors

    SciTech Connect (OSTI)

    WRIGHT,STEVEN A.; HOUTS,MICHAEL

    2000-11-22T23:59:59.000Z

    Heat pipes are often proposed as cooling system components for small fission reactors. SAFE-300 and STAR-C are two reactor concepts that use heat pipes as an integral part of the cooling system. Heat pipes have been used in reactors to cool components within radiation tests (Deverall, 1973); however, no reactor has been built or tested that uses heat pipes solely as the primary cooling system. Heat pipe cooled reactors will likely require the development of a test reactor to determine the main differences in operational behavior from forced cooled reactors. The purpose of this paper is to describe the results of a systems code capable of modeling the coupling between the reactor kinetics and heat pipe controlled heat transport. Heat transport in heat pipe reactors is complex and highly system dependent. Nevertheless, in general terms it relies on heat flowing from the fuel pins through the heat pipe, to the heat exchanger, and then ultimately into the power conversion system and heat sink. A system model is described that is capable of modeling coupled reactor kinetics phenomena, heat transfer dynamics within the fuel pins, and the transient behavior of heat pipes (including the melting of the working fluid). The paper focuses primarily on the coupling effects caused by reactor feedback and compares the observations with forced cooled reactors. A number of reactor startup transients have been modeled, and issues such as power peaking, and power-to-flow mismatches, and loading transients were examined, including the possibility of heat flow from the heat exchanger back into the reactor. This system model is envisioned as a tool to be used for screening various heat pipe cooled reactor concepts, for designing and developing test facility requirements, for use in safety evaluations, and for developing test criteria for in-pile and out-of-pile test facilities.

  7. Three important parts of an integrated plant are reactors, separators and a heat exchanger network (HEN) for heat recovery. Within the process engineering community, much

    E-Print Network [OSTI]

    Skogestad, Sigurd

    exchanger network (HEN) for heat recovery. Within the process engineering community, much attention has beeni ABSTRACT Three important parts of an integrated plant are reactors, separators and a heat and in particular to optimal operation of HENs. The purpose of heat integration is to save energy, but the HEN also

  8. Innovative and Advanced Coupled Neutron Transport and Thermal Hydraulic Method (Tool) for the Design, Analysis and Optimization of VHTR/NGNP Prismatic Reactors

    SciTech Connect (OSTI)

    Rahnema, Farzad; Garimeela, Srinivas; Ougouag, Abderrafi; Zhang, Dingkang

    2013-11-29T23:59:59.000Z

    This project will develop a 3D, advanced coarse mesh transport method (COMET-Hex) for steady- state and transient analyses in advanced very high-temperature reactors (VHTRs). The project will lead to a coupled neutronics and thermal hydraulic (T/H) core simulation tool with fuel depletion capability. The computational tool will be developed in hexagonal geometry, based solely on transport theory without (spatial) homogenization in complicated 3D geometries. In addition to the hexagonal geometry extension, collaborators will concurrently develop three additional capabilities to increase the code’s versatility as an advanced and robust core simulator for VHTRs. First, the project team will develop and implement a depletion method within the core simulator. Second, the team will develop an elementary (proof-of-concept) 1D time-dependent transport method for efficient transient analyses. The third capability will be a thermal hydraulic method coupled to the neutronics transport module for VHTRs. Current advancements in reactor core design are pushing VHTRs toward greater core and fuel heterogeneity to pursue higher burn-ups, efficiently transmute used fuel, maximize energy production, and improve plant economics and safety. As a result, an accurate and efficient neutron transport, with capabilities to treat heterogeneous burnable poison effects, is highly desirable for predicting VHTR neutronics performance. This research project’s primary objective is to advance the state of the art for reactor analysis.

  9. Fate and transport of phenol in a packed bed reactor containing simulated solid waste

    SciTech Connect (OSTI)

    Saquing, Jovita M., E-mail: jmsaquing@gmail.com [Department of Civil, Construction, and Environmental Engineering, Campus Box 7908, North Carolina State University, Raleigh, NC 27695-7908 (United States); Knappe, Detlef R.U., E-mail: knappe@ncsu.edu [Department of Civil, Construction, and Environmental Engineering, Campus Box 7908, North Carolina State University, Raleigh, NC 27695-7908 (United States); Barlaz, Morton A., E-mail: barlaz@ncsu.edu [Department of Civil, Construction, and Environmental Engineering, Campus Box 7908, North Carolina State University, Raleigh, NC 27695-7908 (United States)

    2012-02-15T23:59:59.000Z

    Highlights: Black-Right-Pointing-Pointer Anaerobic column experiments were conducted at 37 Degree-Sign C using a simulated waste mixture. Black-Right-Pointing-Pointer Sorption and biodegradation model parameters were determined from batch tests. Black-Right-Pointing-Pointer HYDRUS simulated well the fate and transport of phenol in a fully saturated waste column. Black-Right-Pointing-Pointer The batch biodegradation rate and the rate obtained by inverse modeling differed by a factor of {approx}2. Black-Right-Pointing-Pointer Tracer tests showed the importance of hydrodynamic parameters to improve model estimates. - Abstract: An assessment of the risk to human health and the environment associated with the presence of organic contaminants (OCs) in landfills necessitates reliable predictive models. The overall objectives of this study were to (1) conduct column experiments to measure the fate and transport of an OC in a simulated solid waste mixture, (2) compare the results of column experiments to model predictions using HYDRUS-1D (version 4.13), a contaminant fate and transport model that can be parameterized to simulate the laboratory experimental system, and (3) determine model input parameters from independently conducted batch experiments. Experiments were conducted in which sorption only and sorption plus biodegradation influenced OC transport. HYDRUS-1D can reasonably simulate the fate and transport of phenol in an anaerobic and fully saturated waste column in which biodegradation and sorption are the prevailing fate processes. The agreement between model predictions and column data was imperfect (i.e., within a factor of two) for the sorption plus biodegradation test and the error almost certainly lies in the difficulty of measuring a biodegradation rate that is applicable to the column conditions. Nevertheless, a biodegradation rate estimate that is within a factor of two or even five may be adequate in the context of a landfill, given the extended retention time and the fact that leachate release will be controlled by the infiltration rate which can be minimized by engineering controls.

  10. Integral cross section measurement of the U 235 ( n , n ' ) U 235 m reaction in a pulsed reactor

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

    Bélier, G.; Bond, E. M.; Vieira, D. J.; Authier, N.; Becker, J. A.; Hyneck, D.; Jacquet, X.; Jansen, Y.; Legendre, J.; Macri, R.; et al

    2015-04-01T23:59:59.000Z

    The integral measurement of the neutron inelastic cross section leading to the 26-minute half-life 235mU isomer in a fission-like neutron spectrum is presented. The experiment has been performed at a pulsed reactor, where the internal conversion decay of the isomer was measured using a dedicated electron detector after activation. The sample preparation, efficiency measurement, irradiation, radiochemistry purification, and isomer decay measurement will be presented. We determined the integral cross section for the ²³?U(n,n')235mU reaction to be 1.00±0.13b. This result supports an evaluation performed with TALYS-1.4 code with respect to the isomer excitation as well as the total neutron inelastic scatteringmore »cross section.« less

  11. Integral cross section measurement of the U 235 ( n , n ' ) U 235 m reaction in a pulsed reactor

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

    Bélier, G.; Bond, E. M.; Vieira, D. J.; Authier, N.; Becker, J. A.; Hyneck, D.; Jacquet, X.; Jansen, Y.; Legendre, J.; Macri, R.; Méot, V.; Romain, P.

    2015-04-01T23:59:59.000Z

    The integral measurement of the neutron inelastic cross section leading to the 26-minute half-life 235mU isomer in a fission-like neutron spectrum is presented. The experiment has been performed at a pulsed reactor, where the internal conversion decay of the isomer was measured using a dedicated electron detector after activation. The sample preparation, efficiency measurement, irradiation, radiochemistry purification, and isomer decay measurement will be presented. We determined the integral cross section for the ²³?U(n,n')235mU reaction to be 1.00±0.13b. This result supports an evaluation performed with TALYS-1.4 code with respect to the isomer excitation as well as the total neutron inelastic scattering cross section.

  12. Transportation

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

    Transportation Home Agenda Awards Exhibitors Lodging Posters Registration Transportation Workshops Contact Us User Meeting Archives Users' Executive Committee Getting to Berkeley...

  13. Transportation

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

    Transportation Print Home Agenda Awards Exhibitors Lodging Posters Registration Transportation Workshops Contact Us User Meeting Archives Users' Executive Committee Getting to...

  14. A scoping study of fission product transport from failed fuel during N Reactor postulated accidents

    SciTech Connect (OSTI)

    Hagrman, D.L.

    1988-01-01T23:59:59.000Z

    This report presents a scoping study of cesium, iodine, and tellurium behavior during a cold leg manifold break in the N Reactor. More detail about fission product behavior than has previously been available is provided and key parameters that control this behavior are identified. The LACE LA1 test and evidence from the Power Burst Facility Severe Fuel Damage tests are used to test the key model applied to determine aerosol behavior. Recommendations for future analysis are also provided. The primary result is that most of the cesium, iodine, and tellurium remains in the molten uranium fuel. Only 0.0035 of the total inventory is calculated to be released. Condensation of the most of the species of cesium and iodine that are released is calculated, with 0.998 of the released cesium and iodine condensing in the spacers and upstream end of the connector tubes. Most of the tellurium that is released condenses, but the chemical reaction of tellurium vapor with surfaces is also a major factor in the behavior of this element.

  15. Coupled full core neutron transport/CFD simulations of pressurized water reactors

    SciTech Connect (OSTI)

    Kochunas, B.; Stimpson, S.; Collins, B.; Downar, T. [Dept. of Nuclear Engineering and Radiological Sciences, Univ. of Michigan, 2200 Bonisteel Blvd, Ann Arbor, MI 48104 (United States); Brewster, R.; Baglietto, E. [CD-adapco, 60 Broadhollow Road, Melville, NY 11747 (United States); Yan, J. [Westinghouse Electric Company LLC, Columbia, SC (United States)

    2012-07-01T23:59:59.000Z

    Recently as part of the CASL project, a capability to perform 3D whole-core coupled neutron transport and computational fluid dynamics (CFD) calculations was demonstrated. This work uses the 2D/1D transport code DeCART and the commercial CFD code STAR-CCM+. It builds on previous CASL work demonstrating coupling for smaller spatial domains. The coupling methodology is described along with the problem simulated and results are presented for fresh hot full power conditions. An additional comparison is made to an equivalent model that uses lower order T/H feedback to assess the importance and cost of high fidelity feedback to the neutronics problem. A simulation of a quarter core Combustion Engineering (CE) PWR core was performed with the coupled codes using a Fixed Point Gauss-Seidel iteration technique. The total approximate calculation requirements are nearly 10,000 CPU hours and 1 TB of memory. The problem took 6 coupled iterations to converge. The CFD coupled model and low order T/H feedback model compared well for global solution parameters, with a difference in the critical boron concentration and average outlet temperature of 14 ppm B and 0.94 deg. C, respectively. Differences in the power distribution were more significant with maximum relative differences in the core-wide pin peaking factor (Fq) of 5.37% and average relative differences in flat flux region power of 11.54%. Future work will focus on analyzing problems more relevant to CASL using models with less approximations. (authors)

  16. Integral Validation of Minor Actinide Nuclear Data by using Samples Irradiated at Dounreay Prototype Fast Reactor

    SciTech Connect (OSTI)

    Tsujimoto, Kazufumi; Oigawa, Hiroyuki; Shinohara, Nobuo [Japan Atomic Energy Research Institute, Shirakata Shirane 2-4, Tokai, Ibaraki 319-1195 (Japan)

    2005-05-24T23:59:59.000Z

    The reliability of nuclear data for minor actinides was evaluated by using the results of the post-irradiation experiment for actinide samples irradiated at the Dounreay Prototype Fast Reactor. The burnup calculations with JENDL-3.3, ENDF/B-VI.8, and JEFF-3.0 were performed. From the comparison between the experimental data and the calculational results, in general, the reliability of nuclear data for the minor actinides are at an adequate level for the conceptual design study of transmutation systems. It is, however, found that improvement of the accuracy is necessary for some nuclides, such as 238Pu, 242Pu, and 241Am.

  17. Transportation

    E-Print Network [OSTI]

    Vinson, Steve

    2013-01-01T23:59:59.000Z

    Transportation in ancient Egypt entailed the use of boats2007 Land transport in Roman Egypt: A study of economics andDieter 1991 Building in Egypt: Pharaonic stone masonry. New

  18. Used Nuclear Fuel Loading and Structural Performance Under Normal Conditions of Transport - Modeling, Simulation and Experimental Integration RD&D Plan

    SciTech Connect (OSTI)

    Adkins, Harold E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-04-01T23:59:59.000Z

    Under current U.S. Nuclear Regulatory Commission regulation, it is not sufficient for used nuclear fuel (UNF) to simply maintain its integrity during the storage period, it must maintain its integrity in such a way that it can withstand the physical forces of handling and transportation associated with restaging the fuel and moving it to treatment or recycling facilities, or a geologic repository. Hence it is necessary to understand the performance characteristics of aged UNF cladding and ancillary components under loadings stemming from transport initiatives. Researchers would like to demonstrate that enough information, including experimental support and modeling and simulation capabilities, exists to establish a preliminary determination of UNF structural performance under normal conditions of transport (NCT). This research, development and demonstration (RD&D) plan describes a methodology, including development and use of analytical models, to evaluate loading and associated mechanical responses of UNF rods and key structural components. This methodology will be used to provide a preliminary assessment of the performance characteristics of UNF cladding and ancillary components under rail-related NCT loading. The methodology couples modeling and simulation and experimental efforts currently under way within the Used Fuel Disposition Campaign (UFDC). The methodology will involve limited uncertainty quantification in the form of sensitivity evaluations focused around available fuel and ancillary fuel structure properties exclusively. The work includes collecting information via literature review, soliciting input/guidance from subject matter experts, performing computational analyses, planning experimental measurement and possible execution (depending on timing), and preparing a variety of supporting documents that will feed into and provide the basis for future initiatives. The methodology demonstration will focus on structural performance evaluation of Westinghouse WE 17×17 pressurized water reactor fuel assemblies with a discharge burnup range of 30-58 GWd/MTU (assembly average), loaded in a representative high-capacity (?32 fuel rod assemblies) transportation package. Evaluations will be performed for representative normal conditions of rail transport involving a rail conveyance capable of meeting the Association of American Railroads (AAR) S-2043 specification. UNF modeling is anticipated to be defined to the pellet-cladding level and take in to account influences associated with spacer grids, intermediate fluid mixers, and control components. The influence of common degradation issues such as ductile-to-brittle-transition will also be accounted for. All model development and analysis will be performed with commercially available software packages exclusively. Inputs and analyses will be completely documented, all supporting information will be traceable, and bases will be defendable so as to be most useful to the U.S. Department of Energy community and mission. The expected completion date is the end of fiscal year (FY) 2013.

  19. Integration of High-Temperature Gas-Cooled Reactors into Industrial Process Applications

    SciTech Connect (OSTI)

    Lee Nelson

    2009-10-01T23:59:59.000Z

    This report is a preliminary comparison of conventional and potential HTGR-integrated processesa in several common industrial areas: ? Producing electricity via a traditional power cycle ? Producing hydrogen ? Producing ammonia and ammonia-derived products, such as fertilizer ? Producing gasoline and diesel from natural gas or coal ? Producing substitute natural gas from coal, and ? Steam-assisted gravity drainage (extracting oil from tar sands).

  20. Integral and Separate Effects Tests for Thermal Hydraulics Code Validation for Liquid-Salt Cooled Nuclear Reactors

    SciTech Connect (OSTI)

    Peterson, Per

    2012-10-30T23:59:59.000Z

    The objective of the 3-year project was to collect integral effects test (IET) data to validate the RELAP5-3D code and other thermal hydraulics codes for use in predicting the transient thermal hydraulics response of liquid salt cooled reactor systems, including integral transient response for forced and natural circulation operation. The reference system for the project is a modular, 900-MWth Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a specific type of Fluoride salt-cooled High temperature Reactor (FHR). Two experimental facilities were developed for thermal-hydraulic integral effects tests (IETs) and separate effects tests (SETs). The facilities use simulant fluids for the liquid fluoride salts, with very little distortion to the heat transfer and fluid dynamics behavior. The CIET Test Bay facility was designed, built, and operated. IET data for steady state and transient natural circulation was collected. SET data for convective heat transfer in pebble beds and straight channel geometries was collected. The facility continues to be operational and will be used for future experiments, and for component development. The CIET 2 facility is larger in scope, and its construction and operation has a longer timeline than the duration of this grant. The design for the CIET 2 facility has drawn heavily on the experience and data collected on the CIET Test Bay, and it was completed in parallel with operation of the CIET Test Bay. CIET 2 will demonstrate start-up and shut-down transients and control logic, in addition to LOFC and LOHS transients, and buoyant shut down rod operation during transients. Design of the CIET 2 Facility is complete, and engineering drawings have been submitted to an external vendor for outsourced quality controlled construction. CIET 2 construction and operation continue under another NEUP grant. IET data from both CIET facilities is to be used for validation of system codes used for FHR modeling, such as RELAP5-3D. A set of numerical models were developed in parallel to the experimental work. RELAP5-3D models were developed for the salt-cooled PB-AHTR, and for the simulat fluid CIET natural circulation experimental loop. These models are to be validated by the data collected from CIET. COMSOL finite element models were used to predict the temperature and fluid flow distribution in the annular pebble bed core; they were instrumental for design of SETs, and they can be used for code-to-code comparisons with RELAP5-3D. A number of other small SETs, and numerical models were constructed, as needed, in support of this work. The experiments were designed, constructed and performed to meet CAES quality assurance requirements for test planning, implementation, and documentation; equipment calibration and documentation, procurement document control; training and personnel qualification; analysis/modeling software verification and validation; data acquisition/collection and analysis; and peer review.

  1. Threat of Hydride Re-orientation to Spent Fuel Integrity During Transportation Accidents: Myth or Reality?

    SciTech Connect (OSTI)

    Rashid, Joe [ANATECH, 5435 Oberlin Drive, San Diego, CA 92121 (United States); Machiels, Albert [EPRI, 3420 Hillview Avenue, Palo Alto, CA 94304 (United States)

    2007-07-01T23:59:59.000Z

    The source-term study conducted by Sandia National Laboratories nearly two decades ago for the spent fuel inventory known at the time, which was in the low-to-medium burnup range ({approx}35 GWd/MTU), showed that the effects of transportation accidents on spent fuel failures, and consequential radioactivity release to the environment, were relatively benign. However, with today's discharged fuel burnups routinely greater than 45 GWd/MTU, potential hydride reorientation during interim dry storage, and its effects on cladding properties, has become one of the primary concerns for spent fuel transportation. Laboratory tests of un-irradiated cladding specimens subjected to heat treatments promoting hydride dissolution followed by re-precipitation in the radial direction have shown that relatively moderate concentrations ({approx}70 ppm) of radial hydrides can significantly degrade cladding ductility, at least at room temperature. The absence of specific data that are relevant to high-burnup spent fuel under dry storage conditions have led to the conjecture, deduced from those tests, that massive cladding failures, possibly resulting in fuel reconfiguration, can be expected during cask drop events. Such conclusions are not borne out by the findings in this paper. The analysis results indicate that cladding failure is bi-modal: a state of failure initiation at the cladding ID remaining as part-wall damage with less than 2% probability of occurrence, and a through-wall failure at a probability of 1 E-5. These results indicate that spent fuel conditions that could promote the formation of radial hydrides during dry storage are not sufficient to produce radial hydrides concentrations of significant levels to cause major threat to spent fuel integrity. It is important to note in this regard that the through-wall cladding failure probability of 1 E-5 is of the same order of magnitude as calculated in the cited Sandia study for low burnup fuel. (authors)

  2. Integrated Water Gas Shift Membrane Reactors Utilizing Novel, Non Precious Metal Mixed Matrix Membrane

    SciTech Connect (OSTI)

    Ferraris, John

    2013-09-30T23:59:59.000Z

    Nanoparticles of zeolitic imidazolate frameworks and other related hybrid materials were prepared by modifying published synthesis procedures by introducing bases, changing stoichiometric ratios, or adjusting reaction conditions. These materials were stable at temperatures >300 °C and were compatible with the polymer matrices used to prepare mixed- matrix membranes (MMMs). MMMs tested at 300 °C exhibited a >30 fold increase in permeability, compared to those measured at 35 °C, while maintaining H{sub 2}/CO{sub 2} selectivity. Measurements at high pressure (up to 30 atm) and high temperature (up to 300 °C) resulted in an increase in gas flux across the membrane with retention of selectivity. No variations in permeability were observed at high pressures at either 35 or 300 °C. CO{sub 2}-induced plasticization was not observed for Matrimid®, VTEC, and PBI polymers or their MMMs at 30 atm and 300 °C. Membrane surface modification by cross-linking with ethanol diamine resulted in an increase in H{sub 2}/CO{sub 2} selectivity at 35 °C. Spectrometric analysis showed that the cross-linking was effective to temperatures <150 °C. At higher temperatures, the cross-linked membranes exhibit a H{sub 2}/CO{sub 2} selectivity similar to the uncross-linked polymer. Performance of the polybenzimidazole (PBI) hollow fibers prepared at Santa Fe Science and Technology (SFST, Inc.) showed increased flux o to a flat PBI membrane. A water-gas shift reactor has been built and currently being optimized for testing under DOE conditions.

  3. Integration

    E-Print Network [OSTI]

    Koschorke, Albrecht; Musanovic, Emina

    2013-01-01T23:59:59.000Z

    Integration By Albrecht Koschorkeby Emina Musanovic [Integration (from Lat. integrare, “toa social unity. Social integration is distinct from systemic

  4. TRITIUM PERMEATION AND TRANSPORT IN THE GASOLINE PRODUCTION SYSTEM COUPLED WITH HIGH TEMPERATURE GAS-COOLED REACTORS (HTGRS)

    SciTech Connect (OSTI)

    Chang H. Oh; Eung S. Kim; Mike Patterson

    2011-05-01T23:59:59.000Z

    This paper describes scoping analyses on tritium behaviors in the HTGR-integrated gasoline production system, which is based on a methanol-to-gasoline (MTG) plant. In this system, the HTGR transfers heat and electricity to the MTG system. This system was analyzed using the TPAC code, which was recently developed by Idaho National Laboratory. The global sensitivity analyses were performed to understand and characterize tritium behaviors in the coupled HTGR/MTG system. This Monte Carlo based random sampling method was used to evaluate maximum 17,408 numbers of samples with different input values. According to the analyses, the average tritium concentration in the product gasoline is about 3.05×10-3 Bq/cm3, and 62 % cases are within the tritium effluent limit (= 3.7x10-3 Bq/cm3[STP]). About 0.19% of released tritium is finally transported from the core to the gasoline product through permeations. This study also identified that the following four parameters are important concerning tritium behaviors in the HTGR/MTG system: (1) tritium source, (2) wall thickness of process heat exchanger, (3) operating temperature, and (4) tritium permeation coefficient of process heat exchanger. These four parameters contribute about 95 % of the total output uncertainties. This study strongly recommends focusing our future research on these four parameters to improve modeling accuracy and to mitigate tritium permeation into the gasol ine product. If the permeation barrier is included in the future study, the tritium concentration will be significantly reduced.

  5. Air quality and Intelligent Transportation Systems : understanding Integrated Innovation, Deployment and Adaptation of Public Technologies

    E-Print Network [OSTI]

    Dodder, Rebecca Susanne

    2006-01-01T23:59:59.000Z

    During the past two decades, Intelligent Transportation Systems (ITS) have provided transportation organizations with increasingly advanced tools both to operate and manage systems in real-time. At the same time, federal ...

  6. An Integrated Assessment of the Impacts of Hydrogen Economy on Transportation, Energy Use, and Air Emissions

    E-Print Network [OSTI]

    Yeh, Sonia; Loughlin, Daniel H.; Shay, Carol; Gage, Cynthia

    2007-01-01T23:59:59.000Z

    Economy on Transportation, Energy Use, and Air Emissions fossil fuel imports such as natural gas.Economy on Transportation, Energy Use, and Air Emissions penetration of H 2 -FCVs could increase the use of natural gasEconomy on Transportation, Energy Use, and Air Emissions With the most cost-effective sources of hydrogen likely to be natural gas

  7. Maglev vehicles and superconductor technology: Integration of high-speed ground transportation into the air travel system

    SciTech Connect (OSTI)

    Johnson, L.R.; Rote, D.M.; Hull, J.R.; Coffey, H.T.; Daley, J.G.; Giese, R.F.

    1989-04-01T23:59:59.000Z

    This study was undertaken to (1) evaluate the potential contribution of high-temperature superconductors (HTSCs) to the technical and economic feasibility of magnetically levitated (maglev) vehicles, (2) determine the status of maglev transportation research in the United States and abroad, (3) identify the likelihood of a significant transportation market for high-speed maglev vehicles, and (4) provide a preliminary assessment of the potential energy and economic benefits of maglev systems. HTSCs should be considered as an enhancing, rather than an enabling, development for maglev transportation because they should improve reliability and reduce energy and maintenance costs. Superconducting maglev transportation technologies were developed in the United States in the late 1960s and early 1970s. Federal support was withdrawn in 1975, but major maglev transportation programs were continued in Japan and West Germany, where full-scale prototypes now carry passengers at speeds of 250 mi/h in demonstration runs. Maglev systems are generally viewed as very-high-speed train systems, but this study shows that the potential market for maglev technology as a train system, e.g., from one downtown to another, is limited. Rather, aircraft and maglev vehicles should be seen as complementing rather than competing transportation systems. If maglev systems were integrated into major hub airport operations, they could become economical in many relatively high-density US corridors. Air traffic congestion and associated noise and pollutant emissions around airports would also be reduced. 68 refs., 26 figs., 16 tabs.

  8. An Analysis of the Impacts of British Transport Reforms on Transit Integration in the Metropolitan Areas

    E-Print Network [OSTI]

    Rivasplata, Charles Richard

    2006-01-01T23:59:59.000Z

    and C. Nash, 1995. Bus deregulation: ten years on, TransportFN Spon. ________, 1996. Deregulation and privatization: theand N. Topham (Eds. ), Bus Deregulation and Privatisation,

  9. Transportation Perspectives on Automotive Cyber Physical System: Integrating Hardware-in-the-Loop, Software-in-the-Loop and Human-in-the-Loop Simulations

    E-Print Network [OSTI]

    Rajkumar, Ragunathan "Raj"

    1 Transportation Perspectives on Automotive Cyber Physical System: Integrating Hardware and development environment to evaluate automotive cyber physical system (CPS) as well as its components foundation of the automotive CPS for developing and testing vehicular networking and sensing technologies

  10. Integrated fuel performance and thermal-hydraulic sub-channel models for analysis of sodium fast reactors

    E-Print Network [OSTI]

    Fricano, Joseph William

    2012-01-01T23:59:59.000Z

    Sodium Fast Reactors (SFR) show promise as an effective way to produce clean safe nuclear power while properly managing the fuel cycle. Accurate computer modeling is an important step in the design and eventual licensing ...

  11. Integration Of Locational Decisions with the Household Activity Pattern Problem and Its Applications in Transportation Sustainability

    E-Print Network [OSTI]

    Kang, Jee E

    2013-01-01T23:59:59.000Z

    hazardous materials transportation: risk analysis, routing/scheduling and facility location,locations, several papers present applications, such as medicine (Or and Pierskalla, 1979; Chan et al, 2001), waste/hazardous

  12. Integration of Locational Decisions with the Household Activity Pattern Problem and Its Applications in Transportation Sustainability

    E-Print Network [OSTI]

    Kang, Jee Eun

    2013-01-01T23:59:59.000Z

    hazardous materials transportation: risk analysis, routing/scheduling and facility location,locations, several papers present applications, such as medicine (Or and Pierskalla, 1979; Chan et al, 2001), waste/hazardous

  13. Integrating regional strategic transportation planning and supply chain management : along the path to sustainability

    E-Print Network [OSTI]

    Sgouridis, Sgouris P

    2005-01-01T23:59:59.000Z

    A systems perspective for regional strategic transportation planning (RSTP) for freight movements involves an understanding of Supply Chain Management (SCM). This thesis argues that private sector freight shippers and ...

  14. Evaluating the Interstate Highway Transportation System in West Africa: Recommendations for an Integrated Highway Network

    E-Print Network [OSTI]

    Nyang, Lamin Bumi

    2010-12-17T23:59:59.000Z

    facilitate development and spur economic growth in the West African region by attracting foreign investors, residential, commercial, and industrial development along its corridor. With an integrated interstate highway system in West Africa, it is expected... infrastructure so that it becomes the catalyst for Africa’s growth. The regionally integrated corridor approach offers prospects for speedier integration of infrastructure systems in Africa. The vision and ultimate objective for Africa should be to create a...

  15. Engineering Development of Ceramic Membrane Reactor System for Converting Natural Gas to Hydrogen and Synthesis Gas for Liquid Transportation Fuels

    SciTech Connect (OSTI)

    Air Products and Chemicals

    2008-09-30T23:59:59.000Z

    An Air Products-led team successfully developed ITM Syngas technology from the concept stage to a stage where a small-scale engineering prototype was about to be built. This technology produces syngas, a gas containing carbon monoxide and hydrogen, by reacting feed gas, primarily methane and steam, with oxygen that is supplied through an ion transport membrane. An ion transport membrane operates at high temperature and oxygen ions are transported through the dense membrane's crystal lattice when an oxygen partial pressure driving force is applied. This development effort solved many significant technical challenges and successfully scaled-up key aspects of the technology to prototype scale. Throughout the project life, the technology showed significant economic benefits over conventional technologies. While there are still on-going technical challenges to overcome, the progress made under the DOE-funded development project proved that the technology was viable and continued development post the DOE agreement would be warranted.

  16. Annual Report for Hybrid Plasma Reactor/Filter for Transportable Collective Protection Systems—Phase 1B

    SciTech Connect (OSTI)

    Josephson, Gary B.; Tonkyn, Russell G.; Rappe, Kenneth G.; Frye, John G.

    2009-06-01T23:59:59.000Z

    Annual report covering the development of a hybrid nonthermal plasma single-pass filtration system for collective protection. This report covers NTP destruction testing on a high priority Toxic Industrial Material and an surrogate for a sulfur containing chemical agent (e.g. mustard), Effects of catalysts in the nonthermal plasma and catalyst poisoning by the sulfur are presented. Also presented are proof-of-principle data for utilizing ozone created in the NTP as a beneficial reactant to destroy adsorbed contaminants in-situ. Catalysts to decompose the ozone within the adsorbent bed are necessary to convert the adsorber into an ozone reactor.

  17. Computational Transportation

    E-Print Network [OSTI]

    Illinois at Chicago, University of

    ), in-vehicle computers, and computers in the transportation infrastructure are integrated ride- sharing, real-time multi-modal routing and navigation, to autonomous/assisted driving

  18. Analysis of the energy transport and deposition within the reaction chamber of the prometheus inertial fusion energy reactor

    SciTech Connect (OSTI)

    Eggleston, J.E.; Abdou, M.A.; Tillack, M.S. [Univ. of California, Los Angeles, CA (United States)

    1994-12-31T23:59:59.000Z

    One of the parameters affecting the feasibility of Inertial Fusion Energy (IFE) devices is the number of shots per unit time, i.e. the repetition rate. The repetition rate limits the achievable power that can be obtained from the reactor. To obtain an estimate of the allowable time between shots, a code named RECON was developed to model the response of the reaction chamber to the pellet explosion. This paper discusses how the code treats the thermodynamic response of the cavity gas and models the condensation/evaporation of this vapor to and from the first wall. A large amount of energy from the pellet microexplosion is carried by the pellet debris and the x-rays generated in the fusion reaction. Models of x-ray attenuation and ion slowing down are used to estimate the fraction of the pellet energy that is absorbed in the vapor. A large amount of energy is absorbed into the cavity gas, which causes it to become partially ionized. The ionization complicates the calculation of the temperature, pressure, and the radiative heat transfer from the gas to the first wall. To treat this problem, methods developed by Zel`dovich and Raizer are used in modeling the internal energy and the radiative heat flux. RECON was developed to run with a relatively short computational time, yet accurate enough for conceptual reactor design calculations.

  19. An Analysis of the Impacts of British Transport Reforms on Transit Integration in the Metropolitan Areas

    E-Print Network [OSTI]

    Rivasplata, Charles Richard

    2006-01-01T23:59:59.000Z

    contrast to many other cities with the smart card system ofthe three case cities. Fare integration: transit smart cardsonly city in the country to have introduced a transit smart

  20. An advanced algorithm for construction of Integral Transport Matrix Method operators using accumulation of single cell coupling factors

    SciTech Connect (OSTI)

    Powell, B. P.; Azmy, Y. Y. [North Carolina State University, Department of Nuclear Engineering, Burlington Engineering Labs, 2500 Stinston Drive, Raleigh, NC 27695 (United States)

    2013-07-01T23:59:59.000Z

    The Integral Transport Matrix Method (ITMM) has been shown to be an effective method for solving the neutron transport equation in large domains on massively parallel architectures. In the limit of very large number of processors, the speed of the algorithm, and its suitability for unstructured meshes, i.e. other than an ordered Cartesian grid, is limited by the construction of four matrix operators required for obtaining the solution in each sub-domain. The existing algorithm used for construction of these matrix operators, termed the differential mesh sweep, is computationally expensive and was developed for a structured grid. This work proposes the use of a new algorithm for construction of these operators based on the construction of a single, fundamental matrix representing the transport of a particle along every possible path throughout the sub-domain mesh. Each of the operators is constructed by multiplying an element of this fundamental matrix by two factors dependent only upon the operator being constructed and on properties of the emitting and incident cells. The ITMM matrix operator construction time for the new algorithm is demonstrated to be shorter than the existing algorithm in all tested cases with both isotropic and anisotropic scattering considered. While also being a more efficient algorithm on a structured Cartesian grid, the new algorithm is promising in its geometric robustness and potential for being applied to an unstructured mesh, with the ultimate goal of application to an unstructured tetrahedral mesh on a massively parallel architecture. (authors)

  1. Recovery Act: Beneficial CO{sub 2} Capture in an Integrated Algal Biorefinery for Renewable Generation and Transportation Fuels

    SciTech Connect (OSTI)

    Lane, Christopher; Hampel, Kristin; Rismani-Yazdi, Hamid; Kessler, Ben; Moats, Kenneth; Park, Jonathan; Schwenk, Jacob; White, Nicholas; Bakhit, Anis; Bargiel, Jeff; Allnutt, F.C.

    2014-03-31T23:59:59.000Z

    DOE DE-FE0001888 Award, Phase 2, funded research, development, and deployment (RD&D) of Phycal’s pilot-scale, algae to biofuels, bioproducts, and processing facility in Hawai’i. Phycal’s algal-biofuel and bioproducts production system integrates several novel and mature technologies into a system that captures and reuses industrially produced carbon dioxide emissions, which would otherwise go directly to the atmosphere, for the manufacture of renewable energy products and bioproducts from algae (note that these algae are not genetically engineered). At the end of Phase 2, the project as proposed was to encompass 34 acres in Central Oahu and provide large open ponds for algal mass culturing, heterotrophic reactors for the Heteroboost™ process, processing facilities, water recycling facilities, anaerobic digestion facilities, and other integrated processes. The Phase 2 award was divided into two modules, Modules 1 & 2, where the Module 1 effort addressed critical scaling issues, tested highest risk technologies, and set the overall infrastructure needed for a Module 2. Phycal terminated the project prior to executing construction of the first Module. This Final Report covers the development research, detailed design, and the proposed operating strategy for Module 1 of Phase 2.

  2. Integrated technical and economic assessments of transport and storage of hydrogen

    SciTech Connect (OSTI)

    Berry, G.D. [Lawrence Livermore National Lab., CA (United States)]|[Illinois Univ., Urbana, IL (United States); Smith, J.R. [Lawrence Livermore National Lab., CA (United States)

    1994-04-01T23:59:59.000Z

    Transportation will be a major market for hydrogen because of its great size and the value of energy at the wheels of a vehicle in comparison to its heating value. Hydrogen also offers important potential efficiency gains over hydrocarbon fuels. However, hydrogen end-use technologies will not develop without a reliable hydrogen supply infrastructure. By the same token, reliable infrastructures will not develop without end-use demand. Our task is to analyze the costs of various infrastructure options for providing hydrogen, as the number of vehicles serviced increased from very small numbers initially, to moderate numbers in the mid-term and to determine if a smooth transition may be possible. We will determine viable market sizes for transport and storage options by examining the technologies and the capital and operating costs of these systems, as well as related issues such as safety, construction time, etc. The product of our work will be data based scenarios of the likely transitions to hydrogen fuel, beginning with small and progressing to larger numbers of vehicles. We are working closely with the suppliers of relevant technologies to (1) determine realistic component costs, and (2) to assure availability of our analyses to business. Preliminary analyses indicate that the cost of transport and storage is as important as production cost in determining the cost of hydrogen fuel to the consumer, and that home electrolysis and centrally processed liquid hydrogen may provide hydrogen in the initial stages.

  3. ITS Version 6 : the integrated TIGER series of coupled electron/photon Monte Carlo transport codes.

    SciTech Connect (OSTI)

    Franke, Brian Claude; Kensek, Ronald Patrick; Laub, Thomas William

    2008-04-01T23:59:59.000Z

    ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of lineartime-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 6, the latest version of ITS, contains (1) improvements to the ITS 5.0 codes, and (2) conversion to Fortran 90. The general user friendliness of the software has been enhanced through memory allocation to reduce the need for users to modify and recompile the code.

  4. Determination of Optimal Process Flowrates and Reactor Design for Autothermal Hydrogen Production in a Heat-Integrated Ceramic Microchannel Network 

    E-Print Network [OSTI]

    Damodharan, Shalini

    2012-07-16T23:59:59.000Z

    emissions [19]. Hence, hydrogen can be produced on large scale from biomass feedstocks in centralized facilities and subsequently distributed at fueling stations and/or community locations as a universal clean fuel for transportation and power...

  5. Determination of Optimal Process Flowrates and Reactor Design for Autothermal Hydrogen Production in a Heat-Integrated Ceramic Microchannel Network

    E-Print Network [OSTI]

    Damodharan, Shalini

    2012-07-16T23:59:59.000Z

    emissions [19]. Hence, hydrogen can be produced on large scale from biomass feedstocks in centralized facilities and subsequently distributed at fueling stations and/or community locations as a universal clean fuel for transportation and power...

  6. Chamber transport

    SciTech Connect (OSTI)

    OLSON,CRAIG L.

    2000-05-17T23:59:59.000Z

    Heavy ion beam transport through the containment chamber plays a crucial role in all heavy ion fusion (HIF) scenarios. Here, several parameters are used to characterize the operating space for HIF beams; transport modes are assessed in relation to evolving target/accelerator requirements; results of recent relevant experiments and simulations of HIF transport are summarized; and relevant instabilities are reviewed. All transport options still exist, including (1) vacuum ballistic transport, (2) neutralized ballistic transport, and (3) channel-like transport. Presently, the European HIF program favors vacuum ballistic transport, while the US HIF program favors neutralized ballistic transport with channel-like transport as an alternate approach. Further transport research is needed to clearly guide selection of the most attractive, integrated HIF system.

  7. Strategy for the Integration of Hydrogen as a Vehicle Fuel into the Existing Natural Gas Vehicle Fueling Infrastructure of the Interstate Clean Transportation Corridor Project: 22 April 2004--31 August 2005

    SciTech Connect (OSTI)

    Gladstein, Neandross and Associates

    2005-09-01T23:59:59.000Z

    Evaluates opportunities to integrate hydrogen into the fueling stations of the Interstate Clean Transportation Corridor--an existing network of LNG fueling stations in California and Nevada.

  8. Sustainability and Transport

    E-Print Network [OSTI]

    Gilbert, Richard

    2006-01-01T23:59:59.000Z

    2005. Integrating Sustainability into the Trans- portationTHOUGHT PIECE Sustainability and Transport by Richardof the concept of sustainability to transport planning. In

  9. Source Term Estimation of Radioxenon Released from the Fukushima Dai-ichi Nuclear Reactors Using Measured Air Concentrations and Atmospheric Transport Modeling

    SciTech Connect (OSTI)

    Eslinger, Paul W. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Biegalski, S. [Univ. of Texas at Austin, TX (United States); Bowyer, Ted W. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Cooper, Matthew W. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Haas, Derek A. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Hayes, James C. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Hoffman, Ian [Radiation Protection Bureau, Health Canada, Ottawa, ON (Canada); Korpach, E. [Radiation Protection Bureau, Health Canada, Ottawa, ON (Canada); Yi, Jing [Radiation Protection Bureau, Health Canada, Ottawa, ON (Canada); Miley, Harry S. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Rishel, Jeremy P. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Ungar, R. Kurt [Radiation Protection Bureau, Health Canada, Ottawa, ON (Canada); White, Brian [Radiation Protection Bureau, Health Canada, Ottawa, ON (Canada); Woods, Vincent T. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States)

    2014-01-01T23:59:59.000Z

    Systems designed to monitor airborne radionuclides released from underground nuclear explosions detected radioactive fallout from the Fukushima Daiichi nuclear accident in March 2011. Atmospheric transport modeling (ATM) of plumes of noble gases and particulates were performed soon after the accident to determine plausible detection locations of any radioactive releases to the atmosphere. We combine sampling data from multiple International Modeling System (IMS) locations in a new way to estimate the magnitude and time sequence of the releases. Dilution factors from the modeled plume at five different detection locations were combined with 57 atmospheric concentration measurements of 133-Xe taken from March 18 to March 23 to estimate the source term. This approach estimates that 59% of the 1.24×1019 Bq of 133-Xe present in the reactors at the time of the earthquake was released to the atmosphere over a three day period. Source term estimates from combinations of detection sites have lower spread than estimates based on measurements at single detection sites. Sensitivity cases based on data from four or more detection locations bound the source term between 35% and 255% of available xenon inventory.

  10. Integrated Removal of NOx with Carbon Monoxide as Reductant, and Capture of Mercury in a Low Temperature Selective Catalytic and Adsorptive Reactor

    SciTech Connect (OSTI)

    Neville Pinto; Panagiotis Smirniotis; Stephen Thiel

    2010-08-31T23:59:59.000Z

    Coal will likely continue to be a dominant component of power generation in the foreseeable future. This project addresses the issue of environmental compliance for two important pollutants: NO{sub x} and mercury. Integration of emission control units is in principle possible through a Low Temperature Selective Catalytic and Adsorptive Reactor (LTSCAR) in which NO{sub x} removal is achieved in a traditional SCR mode but at low temperature, and, uniquely, using carbon monoxide as a reductant. The capture of mercury is integrated into the same process unit. Such an arrangement would reduce mercury removal costs significantly, and provide improved control for the ultimate disposal of mercury. The work completed in this project demonstrates that the use of CO as a reductant in LTSCR is technically feasible using supported manganese oxide catalysts, that the simultaneous warm-gas capture of elemental and oxidized mercury is technically feasible using both nanostructured chelating adsorbents and ceria-titania-based materials, and that integrated removal of mercury and NO{sub x} is technically feasible using ceria-titania-based materials.

  11. Nanowire-integrated microporous silicon membrane for continuous fluid transport in micro cooling device

    SciTech Connect (OSTI)

    So, Hongyun; Pisano, Albert P. [Department of Mechanical Engineering, Berkeley Sensor and Actuator Center, University of California, Berkeley, California 94720 (United States)] [Department of Mechanical Engineering, Berkeley Sensor and Actuator Center, University of California, Berkeley, California 94720 (United States); Cheng, Jim C. [Department of Electrical Engineering, Berkeley Sensor and Actuator Center, University of California, Berkeley, California 94720 (United States)] [Department of Electrical Engineering, Berkeley Sensor and Actuator Center, University of California, Berkeley, California 94720 (United States)

    2013-10-14T23:59:59.000Z

    We report an efficient passive micro pump system combining the physical properties of nanowires and micropores. This nanowire-integrated microporous silicon membrane was created to feed coolant continuously onto the surface of the wick in a micro cooling device to ensure it remains hydrated and in case of dryout, allow for regeneration of the system. The membrane was fabricated by photoelectrochemical etching to form micropores followed by hydrothermal growth of nanowires. This study shows a promising approach to address thermal management challenges for next generation electronic devices with absence of external power.

  12. Final Technical Report - Integrated Hydrogeophysical and Hydrogeologic Driven Parameter Upscaling for Dual-Domain Transport Modeling

    SciTech Connect (OSTI)

    Shafer, John M

    2012-11-05T23:59:59.000Z

    The three major components of this research were: 1. Application of minimally invasive, cost effective hydrogeophysical techniques (surface and borehole), to generate fine scale (~1m or less) 3D estimates of subsurface heterogeneity. Heterogeneity is defined as spatial variability in hydraulic conductivity and/or hydrolithologic zones. 2. Integration of the fine scale characterization of hydrogeologic parameters with the hydrogeologic facies to upscale the finer scale assessment of heterogeneity to field scale. 3. Determination of the relationship between dual-domain parameters and practical characterization data.

  13. Overview of the integration and operation of the USA/9904/B(U)F-85 RTG transportation system (RTGTS) for Cassini and future missions

    SciTech Connect (OSTI)

    Miller, Roger G. [EG and G Mound Applied Technologies P.O. Box 3000 Miamisburg, Ohio 45343-3000 (United States)

    1997-01-10T23:59:59.000Z

    The USA/9904/B(U)F-85, Radioisotope Thermoelectric Generator (RTG) Transportation System (RTGTS) was designed and tested by Westinghouse Hanford Company but was transferred to EG and G Mound Applied Technologies to transport Radioisotope Thermoelectric Generators (RTGs) in support of the Cassini mission. EG and G Mound Applied Technologies is also the RTGTS custodian. Since the RTGTS is a new system, careful scrutiny must be applied not only to the integration of the system into Mound's operations but also the operation of the system so as to prevent any adverse affects to the performance of the RTGs. This paper details specific precautions that have been applied to the integration and operation of the RTGTS to protect the Cassini RTGs during loading, onloading, transportation, offloading, and unloading.

  14. Overview of the integration and operation of the USA/9904/B(U)F-85 RTG transportation system (RTGTS) for Cassini and future missions

    SciTech Connect (OSTI)

    Miller, R.G. [EGG Mound Applied Technologies P.O. Box 3000 Miamisburg, Ohio45343-3000 (United States)

    1997-01-01T23:59:59.000Z

    The USA/9904/B(U)F-85, Radioisotope Thermoelectric Generator (RTG) Transportation System (RTGTS) was designed and tested by Westinghouse Hanford Company but was transferred to EG&G Mound Applied Technologies to transport Radioisotope Thermoelectric Generators (RTGs) in support of the Cassini mission. EG&G Mound Applied Technologies is also the RTGTS custodian. Since the RTGTS is a new system, careful scrutiny must be applied not only to the integration of the system into Mound{close_quote}s operations but also the operation of the system so as to prevent any adverse affects to the performance of the RTGs. This paper details specific precautions that have been applied to the integration and operation of the RTGTS to protect the Cassini RTGs during loading, onloading, transportation, offloading, and unloading. {copyright} {ital 1997 American Institute of Physics.}

  15. Process Intensification with Integrated Water-Gas-Shift Membrane...

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

    Intensification with Integrated Water-Gas-Shift Membrane Reactor Process Intensification with Integrated Water-Gas-Shift Membrane Reactor water-gas-shift.pdf More Documents &...

  16. electrifyingthefuture transportation

    E-Print Network [OSTI]

    Birmingham, University of

    electrifyingthefuture transportation The UK Government's carbon reduction strategy vehicles and the new Birmingham Science City Energy Systems Integration Laboratory (ESIL) will further enhance this work. The laboratory - unique within the UK and world leading - brings together cutting edge

  17. Simulated Performance of the Integrated Passive Neutron Albedo Reactivity and Self-Interrogation Neutron Resonance Densitometry Detector Designed for Spent Fuel Measurement at the Fugen Reactor in Japan

    SciTech Connect (OSTI)

    Ulrich, Timothy J. II [Los Alamos National Laboratory; Lafleur, Adrienne M. [Los Alamos National Laboratory; Menlove, Howard O. [Los Alamos National Laboratory; Swinhoe, Martyn T. [Los Alamos National Laboratory; Tobin, Stephen J. [Los Alamos National Laboratory; Seya, Michio [Los Alamos National Laboratory; Bolind, Alan M. [Los Alamos National Laboratory

    2012-07-16T23:59:59.000Z

    An integrated nondestructive assay instrument, which combined the Passive Neutron Albedo Reactivity (PNAR) and the Self-Interrogation Neutron Resonance Densitometry (SINRD) techniques, is the research focus for a collaborative effort between Los Alamos National Laboratory (LANL) and the Japanese Atomic Energy Agency as part of the Next Generation Safeguard Initiative. We will quantify the anticipated performance of this experimental system in two physical environments: (1) At LANL we will measure fresh Low Enriched Uranium (LEU) assemblies for which the average enrichment can be varied from 0.2% to 3.2% and for which Gd laced rods will be included. (2) At Fugen we will measure spent Mixed Oxide (MOX-B) and LEU spent fuel assemblies from the heavy water moderated Fugen reactor. The MOX-B assemblies will vary in burnup from {approx}3 GWd/tHM to {approx}20 GWd/tHM while the LEU assemblies ({approx}1.9% initial enrichment) will vary from {approx}2 GWd/tHM to {approx}7 GWd/tHM. The estimated count rates will be calculated using MCNPX. These preliminary results will help the finalization of the hardware design and also serve a guide for the experiment. The hardware of the detector is expected to be fabricated in 2012 with measurements expected to take place in 2012 and 2013. This work is supported by the Next Generation Safeguards Initiative, Office of Nuclear Safeguards and Security, National Nuclear Security Administration.

  18. Nuclear reactor engineering: Reactor systems engineering. Fourth edition, Volume Two

    SciTech Connect (OSTI)

    Glasstone, S.; Sesonske, A.

    1994-12-31T23:59:59.000Z

    This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in the design and operation of nuclear power plants. Extensively updated, the fourth edition includes new materials on reactor safety and risk analysis, regulation, fuel management, waste management and operational aspects of nuclear power. This volume contains the following: the systems concept, design decisions, and information tools; energy transport; reactor fuel management and energy cost considerations; environmental effects of nuclear power and waste management; nuclear reactor safety and regulation; power reactor systems; plant operations; and advanced plants and the future.

  19. Integrity assessment of the ferritic / austenitic dissimilar weld joint between intermediate heat exchanger and steam generator in fast reactor

    SciTech Connect (OSTI)

    Jayakumar, T.; Laha, K.; Chandravathi, K. S.; Parameswaran, P.; Goyal, S.; Kumar, J. G.; Mathew, M. D. [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam- 603 102 (India)

    2012-07-01T23:59:59.000Z

    Integrity of the modified 9Cr-1Mo / alloy 800 dissimilar joint welded with Inconel 182 electrodes has been assessed under creep condition based on the detailed analysis of microstructure and stress distribution across the joint by finite element analysis. A hardness peak at the ferritic / austenitic weld interface and a hardness trough at the inter-critical heat affected zone (HAZ) in ferritic base metal developed. Un-tempered martensite was found at the ferritic / austenitic weld interface to impart high hardness in it; whereas annealing of martensitic structure of modified 9Cr-1Mo steel by inter-critical heating during welding thermal cycle resulted in hardness tough in the inter-critical HAZ. Creep tests were carried out on the joint and ferritic steel base metal at 823 K over a stress range of 160-320 MPa. The joint possessed lower creep rupture strength than its ferritic steel base metal. Failure of the joint at relatively lower stresses occurred at the ferritic / austenitic weld interface; whereas it occurred at inter-critical region of HAZ at moderate stresses. Cavity nucleation associated with the weld interface particles led to premature failure of the joint. Finite element analysis of stress distribution across the weld joint considering the micro-mechanical strength inhomogeneity across it revealed higher von-Mises and principal stresses at the weld interface. These stresses induced preferential creep cavitation at the weld interface. Role of precipitate in enhancing creep cavitation at the weld interface has been elucidated based on the FE analysis of stress distribution across it. (authors)

  20. Regional planning and operations architectures as means to foster transportation integration in the Mexico City Metropolitan Area

    E-Print Network [OSTI]

    Ortiz Mantilla, Bernardo Jose, 1977-

    2005-01-01T23:59:59.000Z

    The MCMA complexity in political, institutional, economical, and jurisdictional terms has resulted in limited coordination between MCMA authorities that in conjunction with the limited role of metropolitan transportation ...

  1. MCNP photon transport benchmarking calculations performed at SRP. Revision 1

    SciTech Connect (OSTI)

    White, A.M.

    1989-12-31T23:59:59.000Z

    Monte Carlo methods have long been used at the Savannah River Laboratory (SRL) to perform criticality calculations for many different processes. To perform transport analyses (both neutron and photon) a two-dimensional infinite lattice integral transport code (GLASS) has been used. The neutron transport portion of the code has been benchmarked against other codes and experimental data. The photon transport portion of the code, which is used to calculate gamma redistribution in the event of a loss of moderator and/or coolant, had not been benchmarked against either. For this reason, the Monte Carlo code MCNP was used to benchmark the photon transport portion of the GLASS code. Preceding this, a brief description of the geometry of the Savannah River Plant`s (SRP) reactor cores and how they were modeled using MCNP will be given.

  2. Reactor Safety Research Programs

    SciTech Connect (OSTI)

    Edler, S. K.

    1981-07-01T23:59:59.000Z

    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.

  3. ITERATIVE METHODS FOR NEUTRON TRANSPORT EIGENVALUE PROBLEMS

    E-Print Network [OSTI]

    Graham, Ivan

    Abstract. We discuss iterative methods for computing criticality in nuclear reactors. In general as the inner solver. Key words. neutron transport, criticality, generalised eigenvalue problem, symmetry. Reactor criticality problems. Climate change is a challenging problem of great contemporary interest

  4. Large-scale analytics and optimization in urban transportation : improving public transit and its integration with vehicle-sharing services

    E-Print Network [OSTI]

    Chiraphadhanakul, Virot

    2013-01-01T23:59:59.000Z

    Public transportation is undeniably an effective way to move a large number of people in a city. Its ineffectiveness, such as long travel times, poor coverage, and lack of direct services, however, makes it unappealing to ...

  5. Promoting sustainable transportation through the integration of cycling with public transit : lessons from Copenhagen and Munich for Singapore

    E-Print Network [OSTI]

    Lee, Eugene Siong Aun

    2010-01-01T23:59:59.000Z

    Sustainable modes of transportation like cycling and public transit are always desired, but trying to achieve a mode shift from the automobile to cycling and public transit can be quite a challenge because cycling and ...

  6. Radiation Shielding for Fusion Reactors

    SciTech Connect (OSTI)

    Santoro, R.T.

    1999-10-01T23:59:59.000Z

    Radiation shielding requirements for fusion reactors present different problems than those for fission reactors and accelerators. Fusion devices, particularly tokamak reactors, are complicated by geometry constraints that complicate disposition of fully effective shielding. This paper reviews some of these shielding issues and suggested solutions for optimizing the machine and biological shielding. Radiation transport calculations are essential for predicting and confirming the nuclear performance of the reactor and, as such, must be an essential part of the reactor design process. Development and optimization of reactor components from the first wall and primary shielding to the penetrations and containment shielding must be carried out in a sensible progression. Initial results from one-dimensional transport calculations are used for scoping studies and are followed by detailed two- and three-dimensional analyses to effectively characterize the overall radiation environment. These detail model calculations are essential for accounting for the radiation leakage through ports and other penetrations in the bulk shield. Careful analysis of component activation and radiation damage is cardinal for defining remote handling requirements, in-situ replacement of components, and personnel access at specific locations inside the reactor containment vessel. Radiation shielding requirements for fusion reactors present different problems than those for fission reactors and accelerators. Fusion devices, particularly tokamak reactors, are complicated by geometry constraints that complicate disposition of fully effective shielding. This paper reviews some of these shielding issues and suggested solutions for optimizing the machine and biological shielding. Radiation transport calculations are essential for predicting and confirming the nuclear performance of the reactor and, as such, must be an essential part of the reactor design process. Development and optimization of reactor components from the first wall and primary shielding to the penetrations and containment shielding must be carried out in a sensible progression. Initial results from one-dimensional transport calculations are used for scoping studies and are followed by detailed two- and three-dimensional analyses to effectively characterize the overall radiation environment. These detail model calculations are essential for accounting for the radiation leakage through ports and other penetrations in the bulk shield. Careful analysis of component activation and radiation damage is cardinal for defining remote handling requirements, in-situ replacement of components, and personnel access at specific locations inside the reactor containment vessel.

  7. NETL - Chemical Looping Reactor

    ScienceCinema (OSTI)

    None

    2014-06-26T23:59:59.000Z

    NETL's Chemical Looping Reactor unit is a high-temperature integrated CLC process with extensive instrumentation to improve computational simulations. A non-reacting test unit is also used to study solids flow at ambient temperature. The CLR unit circulates approximately 1,000 pounds per hour at temperatures around 1,800 degrees Fahrenheit.

  8. NETL - Chemical Looping Reactor

    SciTech Connect (OSTI)

    None

    2013-07-24T23:59:59.000Z

    NETL's Chemical Looping Reactor unit is a high-temperature integrated CLC process with extensive instrumentation to improve computational simulations. A non-reacting test unit is also used to study solids flow at ambient temperature. The CLR unit circulates approximately 1,000 pounds per hour at temperatures around 1,800 degrees Fahrenheit.

  9. Power Burst Facility (PBF) Reactor Reactor Decommissioning

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What's Possible forPortsmouth/Paducah Project Office PressPostdoctoraldecadal7Powder DropperReactor

  10. Corrosion-product transport in PWR secondary systems

    SciTech Connect (OSTI)

    Sawochka, S.G.; Copley, S.E.; Pearl, W.L.

    1981-12-01T23:59:59.000Z

    Corrosion product concentration data for the secondary systems of ten pressurized water reactors were obtained employing integrating sampling techniques. Based on these data, mass transport relations were developed for iron, copper, and nickel and employed to evaluate the effect of parameters such as blowdown, condensate polishing, operating pH and oxygen ingress on corrosion product transport and the buildup of sludge in steam generators. Several generic system designs were considered to illustrate application to plant design for the purpose of minimizing corrosion product input to generating equipment.

  11. Graduate Certificate in Transportation

    E-Print Network [OSTI]

    Bertini, Robert L.

    Graduate Certificate in Transportation Nohad A. Toulan School of Urban Studies and Planning of Engineering and Computer Science integrated transportation systems. The Graduate Certificate in Transportation their capabilities. Students in the program can choose among a wide range of relevant courses in transportation

  12. Integrated Ceramic Membrane System for Hydrogen Production

    SciTech Connect (OSTI)

    Schwartz, Joseph; Lim, Hankwon; Drnevich, Raymond

    2010-08-05T23:59:59.000Z

    Phase I was a technoeconomic feasibility study that defined the process scheme for the integrated ceramic membrane system for hydrogen production and determined the plan for Phase II. The hydrogen production system is comprised of an oxygen transport membrane (OTM) and a hydrogen transport membrane (HTM). Two process options were evaluated: 1) Integrated OTM-HTM reactor – in this configuration, the HTM was a ceramic proton conductor operating at temperatures up to 900°C, and 2) Sequential OTM and HTM reactors – in this configuration, the HTM was assumed to be a Pd alloy operating at less than 600°C. The analysis suggested that there are no technical issues related to either system that cannot be managed. The process with the sequential reactors was found to be more efficient, less expensive, and more likely to be commercialized in a shorter time than the single reactor. Therefore, Phase II focused on the sequential reactor system, specifically, the second stage, or the HTM portion. Work on the OTM portion was conducted in a separate program. Phase IIA began in February 2003. Candidate substrate materials and alloys were identified and porous ceramic tubes were produced and coated with Pd. Much effort was made to develop porous substrates with reasonable pore sizes suitable for Pd alloy coating. The second generation of tubes showed some improvement in pore size control, but this was not enough to get a viable membrane. Further improvements were made to the porous ceramic tube manufacturing process. When a support tube was successfully coated, the membrane was tested to determine the hydrogen flux. The results from all these tests were used to update the technoeconomic analysis from Phase I to confirm that the sequential membrane reactor system can potentially be a low-cost hydrogen supply option when using an existing membrane on a larger scale. Phase IIB began in October 2004 and focused on demonstrating an integrated HTM/water gas shift (WGS) reactor to increase CO conversion and produce more hydrogen than a standard water gas shift reactor would. Substantial improvements in substrate and membrane performance were achieved in another DOE project (DE-FC26-07NT43054). These improved membranes were used for testing in a water gas shift environment in this program. The amount of net H2 generated (defined as the difference of hydrogen produced and fed) was greater than would be produced at equilibrium using conventional water gas shift reactors up to 75 psig because of the shift in equilibrium caused by continuous hydrogen removal. However, methanation happened at higher pressures, 100 and 125 psig, and resulted in less net H2 generated than would be expected by equilibrium conversion alone. An effort to avoid methanation by testing in more oxidizing conditions (by increasing CO2/CO ratio in a feed gas) was successful and net H2 generated was higher (40-60%) than a conventional reactor at equilibrium at all pressures tested (up to 125 psig). A model was developed to predict reactor performance in both cases with and without methanation. The required membrane area depends on conditions, but the required membrane area is about 10 ft2 to produce about 2000 scfh of hydrogen. The maximum amount of hydrogen that can be produced in a membrane reactor decreased significantly due to methanation from about 2600 scfh to about 2400 scfh. Therefore, it is critical to eliminate methanation to fully benefit from the use of a membrane in the reaction. Other modeling work showed that operating a membrane reactor at higher temperature provides an opportunity to make the reactor smaller and potentially provides a significant capital cost savings compared to a shift reactor/PSA combination.

  13. Reactor Safety Research Programs

    SciTech Connect (OSTI)

    Dotson, CW

    1980-08-01T23:59:59.000Z

    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.

  14. F Reactor Inspection

    SciTech Connect (OSTI)

    Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

    2014-10-29T23:59:59.000Z

    Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosure and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."

  15. F Reactor Inspection

    ScienceCinema (OSTI)

    Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

    2014-11-24T23:59:59.000Z

    Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosure and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."

  16. Nuclear reactor engineering: Reactor design basics. Fourth edition, Volume One

    SciTech Connect (OSTI)

    Glasstone, S.; Sesonske, A.

    1994-12-31T23:59:59.000Z

    This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in design and operation of nuclear power plants. Extensively updated, the fourth edition includes new material on reactor safety and risk analysis, regulation, fuel management, waste management, and operational aspects of nuclear power. This volume contains the following: energy from nuclear fission; nuclear reactions and radiations; neutron transport; nuclear design basics; nuclear reactor kinetics and control; radiation protection and shielding; and reactor materials.

  17. Reactor assessments of advanced bumpy torus configurations

    SciTech Connect (OSTI)

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

    1983-01-01T23:59:59.000Z

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

  18. Sandia National Laboratories: renewable energy integration

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

    Grid Integration, Infrastructure Security, Microgrid, News, News & Events, Partnership, Renewable Energy, SMART Grid, Transmission Grid Integration, Transportation Energy Under...

  19. Heat dissipating nuclear reactor

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Lazarus, Jonathan D. (Sunnyvale, CA)

    1987-01-01T23:59:59.000Z

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extends from the metal base plate downwardly and outwardly into the earth.

  20. Heat dissipating nuclear reactor

    DOE Patents [OSTI]

    Hunsbedt, A.; Lazarus, J.D.

    1985-11-21T23:59:59.000Z

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

  1. NREL: Transportation Research - Transportation News

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What's Possible for Renewable Energy: Grid IntegrationReportTransmissionResearchNewsTransportation News

  2. Multi-Applications Small Light Water Reactor - NERI Final Report

    SciTech Connect (OSTI)

    S. Michale Modro; James E. Fisher; Kevan D. Weaver; Jose N. Reyes, Jr.; John T. Groome; Pierre Babka; Thomas M. Carlson

    2003-12-01T23:59:59.000Z

    The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle.

  3. A Local Incident Flux Response Expansion Transport Method for Coupling to the Diffusion Method in Cylindrical Geometry

    SciTech Connect (OSTI)

    Dingkang Zhang; Farzad Rahnema; Abderrafi M. Ougouag

    2013-09-01T23:59:59.000Z

    A local incident flux response expansion transport method is developed to generate transport solutions for coupling to diffusion theory codes regardless of their solution method (e.g., fine mesh, nodal, response based, finite element, etc.) for reactor core calculations in both two-dimensional (2-D) and three-dimensional (3-D) cylindrical geometries. In this approach, a Monte Carlo method is first used to precompute the local transport solution (i.e., response function library) for each unique transport coarse node, in which diffusion theory is not valid due to strong transport effects. The response function library is then used to iteratively determine the albedo coefficients on the diffusion-transport interfaces, which are then used as the coupling parameters within the diffusion code. This interface coupling technique allows a seamless integration of the transport and diffusion methods. The new method retains the detailed heterogeneity of the transport nodes and naturally constructs any local solution within them by a simple superposition of local responses to all incoming fluxes from the contiguous coarse nodes. A new technique is also developed for coupling to fine-mesh diffusion methods/codes. The local transport method/module is tested in 2-D and 3-D pebble-bed reactor benchmark problems consisting of an inner reflector, an annular fuel region, and a controlled outer reflector. It is found that the results predicted by the transport module agree very well with the reference fluxes calculated directly by MCNP in both benchmark problems.

  4. Integrated Kinetic Simulation of Laser-Plasma Interactions, Fast-Electron Generation and Transport in Fast Ignition

    SciTech Connect (OSTI)

    Kemp, A; Cohen, B; Divol, L

    2009-11-16T23:59:59.000Z

    We present new results on the physics of short-pulse laser-matter interaction of kilojoule-picosecond pulses at full spatial and temporal scale, using a new approach that combines a 3D collisional electromagnetic Particle-in-Cell code with an MHD-hybrid model of high-density plasma. In the latter, collisions damp out plasma waves, and an Ohm's law with electron inertia effects neglected determines the electric field. In addition to yielding orders of magnitude in speed-up while avoiding numerical instabilities, this allows us to model the whole problem in a single unified framework: the laser-plasma interaction at sub-critical densities, energy deposition at relativistic critical densities, and fast-electron transport in solid densities. Key questions such as the multi-picosecond temporal evolution of the laser energy conversion into hot electrons, the impact of return currents on the laser-plasma interaction, and the effect of self-generated electric and magnetic fields on electron transport will be addressed. We will report applications to current experiments.

  5. EFFECT OF REACTOR HEAT TRANSFER LIMITATIONS ON CO PREFERENTIAL OXIDATION

    E-Print Network [OSTI]

    Besser, Ronald S.

    and conventional packed-bed lab reactors (m-PBR's). Strong evidence has suggested that the reverse water-gas transport limitations of conventional lab reactors [3,4,5,6]: the fast surface chemistry of the exothermic1 EFFECT OF REACTOR HEAT TRANSFER LIMITATIONS ON CO PREFERENTIAL OXIDATION X. Ouyang, R.S. Besser

  6. The MC21 Monte Carlo Transport Code

    SciTech Connect (OSTI)

    Sutton TM, Donovan TJ, Trumbull TH, Dobreff PS, Caro E, Griesheimer DP, Tyburski LJ, Carpenter DC, Joo H

    2007-01-09T23:59:59.000Z

    MC21 is a new Monte Carlo neutron and photon transport code currently under joint development at the Knolls Atomic Power Laboratory and the Bettis Atomic Power Laboratory. MC21 is the Monte Carlo transport kernel of the broader Common Monte Carlo Design Tool (CMCDT), which is also currently under development. The vision for CMCDT is to provide an automated, computer-aided modeling and post-processing environment integrated with a Monte Carlo solver that is optimized for reactor analysis. CMCDT represents a strategy to push the Monte Carlo method beyond its traditional role as a benchmarking tool or ''tool of last resort'' and into a dominant design role. This paper describes various aspects of the code, including the neutron physics and nuclear data treatments, the geometry representation, and the tally and depletion capabilities.

  7. Transportation Licenses Available | Tech Transfer | ORNL

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

    Transportation SHARE Transportation 200301316 Integrated Inverter Control for Multiple Electric Machines 200701874 Power Conversion Apparatus and Method for Hybrid Electric and...

  8. Potential Impact of Interfacial Bonding Efficiency on High-Burnup Spent Nuclear Fuel Vibration Integrity during Normal Transportation

    SciTech Connect (OSTI)

    Jiang, Hao [ORNL; Wang, Jy-An John [ORNL; Wang, Hong [ORNL

    2014-01-01T23:59:59.000Z

    Finite element analysis (FEA) was used to investigate the impacts of interfacial bonding efficiency at pellet pellet and pellet clad interfaces on spent nuclear fuel (SNF) vibration integrity. The FEA simulation results were also validated and benchmarked with reverse bending fatigue test results on surrogate rods consisting of stainless steel (SS) tubes with alumina-pellet inserts. Bending moments (M) are applied to the FEA models to evaluate the system responses of the surrogate rods. From the induced curvature, , the flexural rigidity EI can be estimated as EI=M/ . The impacts of interfacial bonding efficiency on SNF vibration integrity include the moment carrying capacity distribution between pellets and clad and the impact of cohesion on the flexural rigidity of the surrogate rod system. The result also indicates that the immediate consequences of interfacial de-bonding are a load carrying capacity shift from the fuel pellets to the clad and a reduction of the composite rod flexural rigidity. Therefore, the flexural rigidity of the surrogate rod and the bending moment bearing capacity between the clad and fuel pellets are strongly dependent on the efficiency of interfacial bonding at the pellet pellet and pellet clad interfaces. The above-noted phenomenon was calibrated and validated by reverse bending fatigue testing using a surrogate rod system.

  9. Nuclear reactor engineering

    SciTech Connect (OSTI)

    Glasstone, S.; Sesonske, A.

    1981-01-01T23:59:59.000Z

    Chapters are presented concerning energy from nuclear fission; nuclear reactions and radiations; diffusion and slowing-down of neutrons; principles of reactor analysis; nuclear reactor kinetics and control; energy removal; non-fuel reactor materials; the reactor fuel system; radiation protection and environmental effects; nuclear reactor shielding; nuclear reactor safety; and power reactor systems.

  10. Reactor refueling containment system

    DOE Patents [OSTI]

    Gillett, J.E.; Meuschke, R.E.

    1995-05-02T23:59:59.000Z

    A method of refueling a nuclear reactor is disclosed whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced. 2 figs.

  11. Reactor refueling containment system

    DOE Patents [OSTI]

    Gillett, James E. (Greensburg, PA); Meuschke, Robert E. (Pittsburgh, PA)

    1995-01-01T23:59:59.000Z

    A method of refueling a nuclear reactor whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced.

  12. Nuclear Reactors and Technology

    SciTech Connect (OSTI)

    Cason, D.L.; Hicks, S.C. [eds.

    1992-01-01T23:59:59.000Z

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  13. Bioconversion reactor

    DOE Patents [OSTI]

    McCarty, Perry L. (Stanford, CA); Bachmann, Andre (Palo Alto, CA)

    1992-01-01T23:59:59.000Z

    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.

  14. Catalytic reactor

    DOE Patents [OSTI]

    Aaron, Timothy Mark (East Amherst, NY); Shah, Minish Mahendra (East Amherst, NY); Jibb, Richard John (Amherst, NY)

    2009-03-10T23:59:59.000Z

    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.

  15. Nuclear Reactors and Technology; (USA)

    SciTech Connect (OSTI)

    Cason, D.L.; Hicks, S.C. (eds.)

    1991-01-01T23:59:59.000Z

    Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database (EDB) during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency's Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on EDB and Nuclear Science Abstracts (NSA) database. Current information, added daily to EDB, is available to DOE and its contractors through the DOE integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user's needs.

  16. Nuclear Reactor/Hydrogen Process Interface Including the HyPEP Model

    SciTech Connect (OSTI)

    Steven R. Sherman

    2007-05-01T23:59:59.000Z

    The Nuclear Reactor/Hydrogen Plant interface is the intermediate heat transport loop that will connect a very high temperature gas-cooled nuclear reactor (VHTR) to a thermochemical, high-temperature electrolysis, or hybrid hydrogen production plant. A prototype plant called the Next Generation Nuclear Plant (NGNP) is planned for construction and operation at the Idaho National Laboratory in the 2018-2021 timeframe, and will involve a VHTR, a high-temperature interface, and a hydrogen production plant. The interface is responsible for transporting high-temperature thermal energy from the nuclear reactor to the hydrogen production plant while protecting the nuclear plant from operational disturbances at the hydrogen plant. Development of the interface is occurring under the DOE Nuclear Hydrogen Initiative (NHI) and involves the study, design, and development of high-temperature heat exchangers, heat transport systems, materials, safety, and integrated system models. Research and development work on the system interface began in 2004 and is expected to continue at least until the start of construction of an engineering-scale demonstration plant.

  17. Electrorefining {open_quotes}N{close_quotes} reactor fuel

    SciTech Connect (OSTI)

    Gay, E.C.; Miller, W.E.

    1995-02-01T23:59:59.000Z

    Principles of purifying of uranium metal by electrorefining are reviewed. Metal reactor fuel after irradiation is a form of impure uranium. Dissolution and deposition electrorefining processes were developed for spent metal fuel under the Integral Fast Reactor Program. Application of these processes to the conditioning of spent N-reactor fuel slugs is examined.

  18. Twenty-First Water Reactor Safety Information Meeting. Volume 3, Primary system integrity; Aging research, products and applications; Structural and seismic engineering; Seismology and geology: Proceedings

    SciTech Connect (OSTI)

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

    1994-04-01T23:59:59.000Z

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25-27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Selected papers were indexed separately for inclusion in the Energy Science and Technology Database.

  19. MANTRA: An Integral Reactor Physics Experiment to Infer Actinide Capture Cross-sections from Thorium to Californium with Accelerator Mass Spectrometry

    SciTech Connect (OSTI)

    G. Youinou; C. McGrath; G. Imel; M. Paul; R. Pardo; F. Kondev; M. Salvatores; G. Palmiotti

    2011-08-01T23:59:59.000Z

    The principle of the proposed experiment is to irradiate very pure actinide samples in the Advanced Test Reactor at INL and, after a given time, determine the amount of the different transmutation products. The determination of the nuclide densities before and after neutron irradiation will allow inference of effective neutron capture cross-sections. This approach has been used in the past and the novelty of this experiment is that the atom densities of the different transmutation products will be determined using the Accelerator Mass Spectrometry technique at the ATLAS facility located at ANL. It is currently planned to irradiate the following isotopes: 232Th, 235U, 236U, 238U, 237Np, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 241Am, 243Am, 244Cm and 248Cm.

  20. Sandia National Laboratories: Transportation Energy Center

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

    Infrastructure Security, Microgrid, News, News & Events, Partnership, Renewable Energy, SMART Grid, Transmission Grid Integration, Transportation Energy Under an expanded...

  1. Neutronic reactor

    DOE Patents [OSTI]

    Wende, Charles W. J. (Augusta, GA); Babcock, Dale F. (Wilmington, DE); Menegus, Robert L. (Wilmington, DE)

    1983-01-01T23:59:59.000Z

    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.

  2. OVERVIEW OF PROPOSED TRANSPORTATION ENERGY

    E-Print Network [OSTI]

    ......................................................................................................................12 California Freight Energy Demand Model..............................................................................................13 California Transit Energy Demand ModelOVERVIEW OF PROPOSED TRANSPORTATION ENERGY ANALYSES FOR THE 2007 INTEGRATED ENERGY POLICY REPORT

  3. N Reactor - Hanford Site

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC)Integrated Codes |IsLove Your1 SECTIONES2008-54174MoreMuseum Day atMyNERSCVanN Reactor

  4. PID Control Effectiveness for Surface Reactor Concepts

    SciTech Connect (OSTI)

    Dixon, David D. [North Carolina State University, Raleigh, NC (United States); Los Alamos National Laboratory, Los Alamos, NM (United States); Marsh, Christopher L. [United States Naval Academy, Annapolis, MD (United States); Los Alamos National Laboratory, Los Alamos, NM (United States); Poston, David I. [Los Alamos National Laboratory, Los Alamos, NM (United States)

    2007-01-30T23:59:59.000Z

    Control of space and surface fission reactors should be kept as simple as possible, because of the need for high reliability and the difficulty to diagnose and adapt to control system failures. Fortunately, compact, fast-spectrum, externally controlled reactors are very simple in operation. In fact, for some applications it may be possible to design low-power surface reactors without the need for any reactor control after startup; however, a simple proportional, integral, derivative (PID) controller can allow a higher performance concept and add more flexibility to system operation. This paper investigates the effectiveness of a PID control scheme for several anticipated transients that a surface reactor might experience. To perform these analyses, the surface reactor transient code FRINK was modified to simulate control drum movements based on bulk coolant temperature.

  5. The advanced test reactor strategic evaluation program

    SciTech Connect (OSTI)

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

    1989-01-01T23:59:59.000Z

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

  6. Spheromak reactor-design study

    SciTech Connect (OSTI)

    Les, J.M.

    1981-06-30T23:59:59.000Z

    A general overview of spheromak reactor characteristics, such as MHD stability, start up, and plasma geometry is presented. In addition, comparisons are made between spheromaks, tokamaks and field reversed mirrors. The computer code Sphero is also discussed. Sphero is a zero dimensional time independent transport code that uses particle confinement times and profile parameters as input since they are not known with certainty at the present time. More specifically, Sphero numerically solves a given set of transport equations whose solutions include such variables as fuel ion (deuterium and tritium) density, electron density, alpha particle density and ion, electron temperatures.

  7. Washington: Integrated Transportation Programs & Coordinated...

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

    concerns, sustainability by reducing greenhouse gas emissions and the county's carbon footprint. The project began in August 2010 and completed August 2013. The...

  8. Structure of processes in flow reactor and closed reactor: Flow reactor

    E-Print Network [OSTI]

    Greifswald, Ernst-Moritz-Arndt-Universität

    Structure of processes in flow reactor and closed reactor: Flow reactor Closed reactor Active Zone -- chemical quasi- equilibria, similarity principles and macroscopic kinetics", in: Lectures on Plasma Physics

  9. Hybrid adsorptive membrane reactor

    DOE Patents [OSTI]

    Tsotsis, Theodore T. (Huntington Beach, CA); Sahimi, Muhammad (Altadena, CA); Fayyaz-Najafi, Babak (Richmond, CA); Harale, Aadesh (Los Angeles, CA); Park, Byoung-Gi (Yeosu, KR); Liu, Paul K. T. (Lafayette Hill, PA)

    2011-03-01T23:59:59.000Z

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  10. Nuclear reactor engineering

    SciTech Connect (OSTI)

    Glasstone, S.; Sesonske, A.

    1982-07-01T23:59:59.000Z

    A book is reviewed which emphasizes topics directly related to the light water reactor power plant and the fast reactor power system. Current real-world problems are addressed throughout the text, and a chapter on safety includes much of the postThree Mile Island impact on operating systems. Topics covered include Doppler broadening, neutron resonances, multigroup diffusion theory, reactor kinetics, reactor control, energy removal, nonfuel materials, reactor fuel, radiation protection, environmental effects, and reactor safety.

  11. Novel Reactor Relevant RF Actuator Schemes for the Lower

    E-Print Network [OSTI]

    location for RF launchers · Transport in tokamak sends heat and particles to low field side SOL: ­ Forces reactor is challenging: ­ Survivability is a major issue because of the harsh environment high heat solution that both mitigates PMI / coupling problems and improves core wave physics issues. #12;Reactor

  12. Steady-State Gyrokinetics Transport Code (SSGKT), A Scientific Application Partnership with the Framework Application for Core-Edge Transport Simulations, Final Report

    SciTech Connect (OSTI)

    Fahey, Mark R. [Oak Ridge National Laboratory] [Oak Ridge National Laboratory; Candy, Jeff [General Atomics] [General Atomics

    2013-11-07T23:59:59.000Z

    This project initiated the development of TGYRO ? a steady-state Gyrokinetic transport code (SSGKT) that integrates micro-scale GYRO turbulence simulations into a framework for practical multi-scale simulation of conventional tokamaks as well as future reactors. Using a lightweight master transport code, multiple independent (each massively parallel) gyrokinetic simulations are coordinated. The capability to evolve profiles using the TGLF model was also added to TGYRO and represents a more typical use-case for TGYRO. The goal of the project was to develop a steady-state Gyrokinetic transport code (SSGKT) that integrates micro-scale gyrokinetic turbulence simulations into a framework for practical multi-scale simulation of a burning plasma core ? the International Thermonuclear Experimental Reactor (ITER) in particular. This multi-scale simulation capability will be used to predict the performance (the fusion energy gain, Q) given the H-mode pedestal temperature and density. At present, projections of this type rely on transport models like GLF23, which are based on rather approximate fits to the results of linear and nonlinear simulations. Our goal is to make these performance projections with precise nonlinear gyrokinetic simulations. The method of approach is to use a lightweight master transport code to coordinate multiple independent (each massively parallel) gyrokinetic simulations using the GYRO code. This project targets the practical multi-scale simulation of a reactor core plasma in order to predict the core temperature and density profiles given the H-mode pedestal temperature and density. A master transport code will provide feedback to O(16) independent gyrokinetic simulations (each massively parallel). A successful feedback scheme offers a novel approach to predictive modeling of an important national and international problem. Success in this area of fusion simulations will allow US scientists to direct the research path of ITER over the next two decades. The design of an efficient feedback algorithm is a serious numerical challenge. Although the power source and transport balance coding in the master are standard, it is nontrivial to design a feedback loop that can cope with outputs that are both intermittent and extremely expensive. A prototypical feedback scheme has already been successfully demonstrated for a single global GYRO simulation, although the robustness and efficiency are likely far from optimal. Once the transport feedback scheme is perfected, it could, in principle, be embedded into any of the more elaborate transport codes (ONETWO, TRANSP, and CORSICA), or adopted by other FSP-related multi-scale projects.

  13. Overview of the Westinghouse Small Modular Reactor building layout

    SciTech Connect (OSTI)

    Cronje, J. M. [Westinghouse Electric Company LLC, Centurion (South Africa); Van Wyk, J. J.; Memmott, M. J. [Westinghouse Electric Company LLC, Cranberry Township, PA (United States)

    2012-07-01T23:59:59.000Z

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the third in a series of four papers, which describe the design and functionality of the Westinghouse SMR. It focuses in particular upon the plant building layout and modular design of the Westinghouse SMR. In the development of small modular reactors, the building layout is an area where the safety of the plant can be improved by applying new design approaches. This paper will present an overview of the Westinghouse SMR building layout and indicate how the design features improve the safety and robustness of the plant. The Westinghouse SMR is designed with no shared systems between individual reactor units. The main buildings inside the security fence are the nuclear island, the rad-waste building, the annex building, and the turbine building. All safety related equipment is located in the nuclear island, which is a seismic class 1 building. To further enhance the safety and robustness of the design, the reactor, containment, and most of the safety related equipment are located below grade on the nuclear island. This reduces the possibility of severe damage from external threats or natural disasters. Two safety related ultimate heat sink (UHS) water tanks that are used for decay heat removal are located above grade, but are redundant and physically separated as far as possible for improved safety. The reactor and containment vessel are located below grade in the center of the nuclear island. The rad-waste and other radioactive systems are located on the bottom floors to limit the radiation exposure to personnel. The Westinghouse SMR safety trains are completely separated into four unconnected quadrants of the building, with access between quadrants only allowed above grade. This is an improvement to conventional reactor design since it prevents failures of multiple trains during floods or fires and other external events. The main control room is located below grade, with a remote shutdown room in a different quadrant. All defense in depth systems are placed on the nuclear island, primarily above grade, while the safety systems are located on lower floors. The economics of the Westinghouse SMR challenges the established approach of large Light Water Reactors (LWR) that utilized the economies of scale to reach economic competitiveness. To serve the market expectation of smaller capital investment and cost competitive energy, a modular design approach is implemented within the Westinghouse SMR. The Westinghouse SMR building layout integrates the three basic design constraints of modularization; transportation, handling and module-joining technology. (authors)

  14. NREL: Transportation Research - Transportation Secure Data Center

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What's Possible for Renewable Energy: Grid IntegrationReportTransmissionResearchNewsTransportation

  15. Distributed expert systems for nuclear reactor control

    SciTech Connect (OSTI)

    Otaduy, P.J.

    1992-12-01T23:59:59.000Z

    A network of distributed expert systems is the heart of a prototype supervisory control architecture developed at the Oak Ridge National Laboratory (ORNL) for an advanced multimodular reactor. Eight expert systems encode knowledge on signal acquisition, diagnostics, safeguards, and control strategies in a hybrid rule-based, multiprocessing and object-oriented distributed computing environment. An interactive simulation of a power block consisting of three reactors and one turbine provides a realistic, testbed for performance analysis of the integrated control system in real-time. Implementation details and representative reactor transients are discussed.

  16. Distributed expert systems for nuclear reactor control

    SciTech Connect (OSTI)

    Otaduy, P.J.

    1992-01-01T23:59:59.000Z

    A network of distributed expert systems is the heart of a prototype supervisory control architecture developed at the Oak Ridge National Laboratory (ORNL) for an advanced multimodular reactor. Eight expert systems encode knowledge on signal acquisition, diagnostics, safeguards, and control strategies in a hybrid rule-based, multiprocessing and object-oriented distributed computing environment. An interactive simulation of a power block consisting of three reactors and one turbine provides a realistic, testbed for performance analysis of the integrated control system in real-time. Implementation details and representative reactor transients are discussed.

  17. Simulation of household in-home and transportation energy use : an integrated behavioral model for estimating energy consumption at the neighborhood scale

    E-Print Network [OSTI]

    Yu, Feifei, S.M. Massachusetts Institute of Technology

    2013-01-01T23:59:59.000Z

    Household in-home activities and out-of-home transportation are two major sources of urban energy consumption. In light of China's rapid urbanization and income growth, changing lifestyles and consumer patterns - evident ...

  18. Hydrogen and water reactor safety: proceedings

    SciTech Connect (OSTI)

    Not Available

    1982-01-01T23:59:59.000Z

    Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability.

  19. SRS Small Modular Reactors

    SciTech Connect (OSTI)

    None

    2012-04-27T23:59:59.000Z

    The small modular reactor program at the Savannah River Site and the Savannah River National Laboratory.

  20. SRS Small Modular Reactors

    ScienceCinema (OSTI)

    None

    2014-05-21T23:59:59.000Z

    The small modular reactor program at the Savannah River Site and the Savannah River National Laboratory.

  1. Reactor safety method

    DOE Patents [OSTI]

    Vachon, Lawrence J. (Clairton, PA)

    1980-03-11T23:59:59.000Z

    This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

  2. Nuclear reactor

    DOE Patents [OSTI]

    Thomson, Wallace B. (Severna Park, MD)

    2004-03-16T23:59:59.000Z

    A nuclear reactor comprising a cylindrical pressure vessel, an elongated annular core centrally disposed within and spaced from the pressure vessel, and a plurality of ducts disposed longitudinally of the pressure vessel about the periphery thereof, said core comprising an annular active portion, an annular reflector just inside the active portion, and an annular reflector just outside the active a portion, said annular active portion comprising rectangular slab, porous fuel elements radially disposed around the inner reflector and extending the length of the active portion, wedge-shaped, porous moderator elements disposed adjacent one face of each fuel element and extending the length of the fuel element, the fuel and moderator elements being oriented so that the fuel elements face each other and the moderator elements do likewise, adjacent moderator elements being spaced to provide air inlet channels, and adjacent fuel elements being spaced to provide air outlet channels which communicate with the interior of the peripheral ducts, and means for introducing air into the air inlet channels which passes through the porous moderator elements and porous fuel elements to the outlet channel.

  3. Uranium Oxide Aerosol Transport in Porous Graphite

    SciTech Connect (OSTI)

    Blanchard, Jeremy; Gerlach, David C.; Scheele, Randall D.; Stewart, Mark L.; Reid, Bruce D.; Gauglitz, Phillip A.; Bagaasen, Larry M.; Brown, Charles C.; Iovin, Cristian; Delegard, Calvin H.; Zelenyuk, Alla; Buck, Edgar C.; Riley, Brian J.; Burns, Carolyn A.

    2012-01-23T23:59:59.000Z

    The objective of this paper is to investigate the transport of uranium oxide particles that may be present in carbon dioxide (CO2) gas coolant, into the graphite blocks of gas-cooled, graphite moderated reactors. The transport of uranium oxide in the coolant system, and subsequent deposition of this material in the graphite, of such reactors is of interest because it has the potential to influence the application of the Graphite Isotope Ratio Method (GIRM). The GIRM is a technology that has been developed to validate the declared operation of graphite moderated reactors. GIRM exploits isotopic ratio changes that occur in the impurity elements present in the graphite to infer cumulative exposure and hence the reactor’s lifetime cumulative plutonium production. Reference Gesh, et. al., for a more complete discussion on the GIRM technology.

  4. LBB application in the US operating and advanced reactors

    SciTech Connect (OSTI)

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

    1997-04-01T23:59:59.000Z

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

  5. Atmospheric Pressure Reactor System | EMSL

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

    Atmospheric Pressure Reactor System Atmospheric Pressure Reactor System The atmospheric pressure reactor system is designed for testing the efficiency of various catalysts for the...

  6. Smart vehicular transportation systems

    SciTech Connect (OSTI)

    Little, C.Q.; Wilson, C.W.

    1997-05-01T23:59:59.000Z

    This work builds upon established Sandia intelligent systems technology to develop a unique approach for the integration of intelligent system control into the US Highway and urban transportation systems. The Sandia developed concept of the COPILOT controller integrates a human driver with computer control to increase human performance while reducing reliance on detailed driver attention. This research extends Sandia expertise in sensor based, real-time control of robotics systems to high speed transportation systems. Knowledge in the form of maps and performance characteristics of vehicles provides the automatic decision making intelligence needed to plan optimum routes, maintain safe driving speeds and distances, avoid collisions, and conserve fuel.

  7. amphetamine-induced reverse transport: Topics by E-print Network

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

    model for analyzing the transport performance of a Reversed Field Pinch (RFP)-type of thermonuclear fusion reactor has been developed. The study has been focused on determining...

  8. Type B investigation of the iridium contamination event at the High Flux Isotope Reactor on September 7, 1993

    SciTech Connect (OSTI)

    Not Available

    1994-03-01T23:59:59.000Z

    On the title date, at ORNL, area radiation alarms sounded during a routine transfer of a shielding cask (containing 60 Ci{sup 192}Ir) from the HFIR pool side to a transport truck. Small amounts of Ir were released from the cask onto the reactor bay floor. The floor was cleaned, and the cask was shipped to a hot cell at Building 3047 on Oct. 3, 1993. The event was caused by rupture of one of the Ir target rods after it was loaded into the cask for normal transport operations; the rupture was the result of steam generation in the target rod soon after it was placed in the cask (water had entered the target rod through a tiny defect in a weld while it was in the reactor under pressure). While the target rods were in the reactor and reactor pool, there was sufficient cooling to prevent steam generation; when the target rod was loaded into the dry transport cask, the temperature increased enough to result in boiling of the trapped water and produced high enough pressure to result in rupture. The escaping steam ejected some of the Ir pellets. The event was reported as Occurrence Report Number ORO--MMES-X10HFIR-1993-0030, dated Sept. 8, 1993. Analysis indicated that the following conditions were probable causes: less than adequate welding procedures, practices, or techniques, material controls, or inspection methods, or combination thereof, could have led to weld defects, affecting the integrity of target rod IR-75; less than adequate secondary containment in the cask allowed Ir pellets to escape.

  9. Dynamic reactor modeling with applications to SPR and ZEDNA.

    SciTech Connect (OSTI)

    Suo-Anttila, Ahti Jorma

    2011-12-01T23:59:59.000Z

    A dynamic reactor model has been developed for pulse-type reactor applications. The model predicts reactor power, axial and radial fuel expansion, prompt and delayed neutron population, and prompt and delayed gamma population. All model predictions are made as a function of time. The model includes the reactivity effect of fuel expansion on a dynamic timescale as a feedback mechanism for reactor power. All inputs to the model are calculated from first principles, either directly by solving systems of equations, or indirectly from Monte Carlo N-Particle Transport Code (MCNP) derived results. The model does not include any empirical parameters that can be adjusted to match experimental data. Comparisons of model predictions to actual Sandia Pulse Reactor SPR-III pulses show very good agreement for a full range of pulse magnitudes. The model is also applied to Z-pinch externally driven neutron assembly (ZEDNA) type reactor designs to model both normal and off-normal ZEDNA operations.

  10. Automated Transportation Management System (ATMS) | Department...

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

    of Energy's (DOE's) Automated Transportation Management System is an integrated web-based logistics management system allowing users to manage inbound and outbound freight...

  11. Cross-flow electrochemical reactor cells, cross-flow reactors, and use of cross-flow reactors for oxidation reactions

    DOE Patents [OSTI]

    Balachandran, Uthamalingam (Hinsdale, IL); Poeppel, Roger B. (Glen Ellyn, IL); Kleefisch, Mark S. (Naperville, IL); Kobylinski, Thaddeus P. (Lisle, IL); Udovich, Carl A. (Joliet, IL)

    1994-01-01T23:59:59.000Z

    This invention discloses cross-flow electrochemical reactor cells containing oxygen permeable materials which have both electron conductivity and oxygen ion conductivity, cross-flow reactors, and electrochemical processes using cross-flow reactor cells having oxygen permeable monolithic cores to control and facilitate transport of oxygen from an oxygen-containing gas stream to oxidation reactions of organic compounds in another gas stream. These cross-flow electrochemical reactors comprise a hollow ceramic blade positioned across a gas stream flow or a stack of crossed hollow ceramic blades containing a channel or channels for flow of gas streams. Each channel has at least one channel wall disposed between a channel and a portion of an outer surface of the ceramic blade, or a common wall with adjacent blades in a stack comprising a gas-impervious mixed metal oxide material of a perovskite structure having electron conductivity and oxygen ion conductivity. The invention includes reactors comprising first and second zones seprated by gas-impervious mixed metal oxide material material having electron conductivity and oxygen ion conductivity. Prefered gas-impervious materials comprise at least one mixed metal oxide having a perovskite structure or perovskite-like structure. The invention includes, also, oxidation processes controlled by using these electrochemical reactors, and these reactions do not require an external source of electrical potential or any external electric circuit for oxidation to proceed.

  12. Local Transportation

    E-Print Network [OSTI]

    Local Transportation. Transportation from the Airport to Hotel. There are two types of taxi companies that operate at the airport: special and regular taxis (

  13. Greening Transportation

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

    Transportation Goal 2: Greening Transportation LANL supports and encourages employees to reduce their personal greenhouse gas emissions by offering various commuting and work...

  14. Reactor Sharing Program

    SciTech Connect (OSTI)

    Vernetson, W.G.

    1993-01-01T23:59:59.000Z

    Progress achieved at the University of Florida Training Reactor (UFTR) facility through the US Department of Energy's University Reactor Sharing Program is reported for the period of 1991--1992.

  15. Undergraduate reactor control experiment

    SciTech Connect (OSTI)

    Edwards, R.M.; Power, M.A.; Bryan, M. (Pennsylvania State Univ., University Park (United States))

    1992-01-01T23:59:59.000Z

    A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise.

  16. Attrition reactor system

    DOE Patents [OSTI]

    Scott, Charles D. (Oak Ridge, TN); Davison, Brian H. (Knoxvile, TN)

    1993-01-01T23:59:59.000Z

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

  17. Attrition reactor system

    DOE Patents [OSTI]

    Scott, C.D.; Davison, B.H.

    1993-09-28T23:59:59.000Z

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.

  18. ITS version 5.0 :the integrated TIGER series of coupled electron/Photon monte carlo transport codes with CAD geometry.

    SciTech Connect (OSTI)

    Franke, Brian Claude; Kensek, Ronald Patrick; Laub, Thomas William

    2005-09-01T23:59:59.000Z

    ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of linear time-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 5.0, the latest version of ITS, contains (1) improvements to the ITS 3.0 continuous-energy codes, (2) multigroup codes with adjoint transport capabilities, (3) parallel implementations of all ITS codes, (4) a general purpose geometry engine for linking with CAD or other geometry formats, and (5) the Cholla facet geometry library. Moreover, the general user friendliness of the software has been enhanced through increased internal error checking and improved code portability.

  19. compactness in transport theory Mustapha Mokhtar-Kharroubi

    E-Print Network [OSTI]

    Boyer, Edmond

    A main feature of spectra of transport operators in nuclear reactor theory relies on the compactness (orOn L1 compactness in transport theory Mustapha Mokhtar-Kharroubi Département de Mathématiques spectral theory of neutron transport equations on both n-dimensional torus and spatial domains with ...nite

  20. Reactor pressure vessel with forged nozzles

    DOE Patents [OSTI]

    Desai, Dilip R. (Fremont, CA)

    1993-01-01T23:59:59.000Z

    Inlet nozzles for a gravity-driven cooling system (GDCS) are forged with a cylindrical reactor pressure vessel (RPV) section to which a support skirt for the RPV is attached. The forging provides enhanced RPV integrity around the nozzle and substantial reduction of in-service inspection costs by eliminating GDCS nozzle-to-RPV welds.

  1. Advanced Test Reactor Tour

    SciTech Connect (OSTI)

    Miley, Don

    2011-01-01T23:59:59.000Z

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

  2. High solids fermentation reactor

    DOE Patents [OSTI]

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-03-02T23:59:59.000Z

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  3. High solids fermentation reactor

    DOE Patents [OSTI]

    Wyman, Charles E. (Lakewood, CO); Grohmann, Karel (Littleton, CO); Himmel, Michael E. (Littleton, CO); Richard, Christopher J. (Lakewood, CO)

    1993-01-01T23:59:59.000Z

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  4. Improved vortex reactor system

    DOE Patents [OSTI]

    Diebold, James P. (Lakewood, CO); Scahill, John W. (Evergreen, CO)

    1995-01-01T23:59:59.000Z

    An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

  5. Advanced Test Reactor Tour

    ScienceCinema (OSTI)

    Miley, Don

    2013-05-28T23:59:59.000Z

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

  6. EBS Radionuclide Transport Abstraction

    SciTech Connect (OSTI)

    J. Prouty

    2006-07-14T23:59:59.000Z

    The purpose of this report is to develop and analyze the engineered barrier system (EBS) radionuclide transport abstraction model, consistent with Level I and Level II model validation, as identified in Technical Work Plan for: Near-Field Environment and Transport: Engineered Barrier System: Radionuclide Transport Abstraction Model Report Integration (BSC 2005 [DIRS 173617]). The EBS radionuclide transport abstraction (or EBS RT Abstraction) is the conceptual model used in the total system performance assessment (TSPA) to determine the rate of radionuclide releases from the EBS to the unsaturated zone (UZ). The EBS RT Abstraction conceptual model consists of two main components: a flow model and a transport model. Both models are developed mathematically from first principles in order to show explicitly what assumptions, simplifications, and approximations are incorporated into the models used in the TSPA. The flow model defines the pathways for water flow in the EBS and specifies how the flow rate is computed in each pathway. Input to this model includes the seepage flux into a drift. The seepage flux is potentially split by the drip shield, with some (or all) of the flux being diverted by the drip shield and some passing through breaches in the drip shield that might result from corrosion or seismic damage. The flux through drip shield breaches is potentially split by the waste package, with some (or all) of the flux being diverted by the waste package and some passing through waste package breaches that might result from corrosion or seismic damage. Neither the drip shield nor the waste package survives an igneous intrusion, so the flux splitting submodel is not used in the igneous scenario class. The flow model is validated in an independent model validation technical review. The drip shield and waste package flux splitting algorithms are developed and validated using experimental data. The transport model considers advective transport and diffusive transport from a breached waste package. Advective transport occurs when radionuclides that are dissolved or sorbed onto colloids (or both) are carried from the waste package by the portion of the seepage flux that passes through waste package breaches. Diffusive transport occurs as a result of a gradient in radionuclide concentration and may take place while advective transport is also occurring, as well as when no advective transport is occurring. Diffusive transport is addressed in detail because it is the sole means of transport when there is no flow through a waste package, which may dominate during the regulatory compliance period in the nominal and seismic scenarios. The advective transport rate, when it occurs, is generally greater than the diffusive transport rate. Colloid-facilitated advective and diffusive transport is also modeled and is presented in detail in Appendix B of this report.

  7. MOCABA: a general Monte Carlo-Bayes procedure for improved predictions of integral functions of nuclear data

    E-Print Network [OSTI]

    Axel Hoefer; Oliver Buss; Maik Hennebach; Michael Schmid; Dieter Porsch

    2014-11-12T23:59:59.000Z

    MOCABA is a combination of Monte Carlo sampling and Bayesian updating algorithms for the prediction of integral functions of nuclear data, such as reactor power distributions or neutron multiplication factors. Similarly to the established Generalized Linear Least Squares (GLLS) methodology, MOCABA offers the capability to utilize integral experimental data to reduce the prior uncertainty of integral observables. The MOCABA approach, however, does not involve any series expansions and, therefore, does not suffer from the breakdown of first-order perturbation theory for large nuclear data uncertainties. This is related to the fact that, in contrast to the GLLS method, the updating mechanism within MOCABA is applied directly to the integral observables without having to "adjust" any nuclear data. A central part of MOCABA is the nuclear data Monte Carlo program NUDUNA, which performs random sampling of nuclear data evaluations according to their covariance information and converts them into libraries for transport code systems like MCNP or SCALE. What is special about MOCABA is that it can be applied to any integral function of nuclear data, and any integral measurement can be taken into account to improve the prediction of an integral observable of interest. In this paper we present two example applications of the MOCABA framework: the prediction of the neutron multiplication factor of a water-moderated PWR fuel assembly based on 21 criticality safety benchmark experiments and the prediction of the power distribution within a toy model reactor containing 100 fuel assemblies.

  8. Multiscale Strategic Planning Model for the Design of Integrated Ethanol and Gasoline Supply Chain

    E-Print Network [OSTI]

    Grossmann, Ignacio E.

    of concern are related to biomass transportation, supply chain design of bioethanol and its integration

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

    SciTech Connect (OSTI)

    Not Available

    1986-01-01T23:59:59.000Z

    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).

  10. RSMASS-D models: An improved method for estimating reactor and shield mass for space reactor applications

    SciTech Connect (OSTI)

    Marshall, A.C.

    1997-10-01T23:59:59.000Z

    Three relatively simple mathematical models have been developed to estimate minimum reactor and radiation shield masses for liquid-metal-cooled reactors (LMRs), in-core thermionic fuel element (TFE) reactors, and out-of-core thermionic reactors (OTRs). The approach was based on much of the methodology developed for the Reactor/Shield Mass (RSMASS) model. Like the original RSMASS models, the new RSMASS-derivative (RSMASS-D) models use a combination of simple equations derived from reactor physics and other fundamental considerations, along with tabulations of data from more detailed neutron and gamma transport theory computations. All three models vary basic design parameters within a range specified by the user to achieve a parameter choice that yields a minimum mass for the power level and operational time of interest. The impact of critical mass, fuel damage, and thermal limitations are accounted for to determine the required fuel mass. The effect of thermionic limitations are also taken into account for the thermionic reactor models. All major reactor component masses are estimated, as well as instrumentation and control (I&C), boom, and safety system masses. A new shield model was developed and incorporated into all three reactor concept models. The new shield model is more accurate and simpler to use than the approach used in the original RSMASS model. The estimated reactor and shield masses agree with the mass predictions from separate detailed calculations within 15 percent for all three models.

  11. Reactor vessel support system

    DOE Patents [OSTI]

    Golden, Martin P. (Trafford, PA); Holley, John C. (McKeesport, PA)

    1982-01-01T23:59:59.000Z

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

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

    SciTech Connect (OSTI)

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

    2012-07-01T23:59:59.000Z

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

  13. Reactor pressure vessel vented head

    DOE Patents [OSTI]

    Sawabe, James K. (San Jose, CA)

    1994-01-11T23:59:59.000Z

    A head for closing a nuclear reactor pressure vessel shell includes an arcuate dome having an integral head flange which includes a mating surface for sealingly mating with the shell upon assembly therewith. The head flange includes an internal passage extending therethrough with a first port being disposed on the head mating surface. A vent line includes a proximal end disposed in flow communication with the head internal passage, and a distal end disposed in flow communication with the inside of the dome for channeling a fluid therethrough. The vent line is fixedly joined to the dome and is carried therewith when the head is assembled to and disassembled from the shell.

  14. Reactor water cleanup system

    DOE Patents [OSTI]

    Gluntz, D.M.; Taft, W.E.

    1994-12-20T23:59:59.000Z

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

  15. Spinning fluids reactor

    SciTech Connect (OSTI)

    Miller, Jan D; Hupka, Jan; Aranowski, Robert

    2012-11-20T23:59:59.000Z

    A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

  16. Determining Reactor Neutrino Flux

    E-Print Network [OSTI]

    Jun Cao

    2012-03-08T23:59:59.000Z

    Flux is an important source of uncertainties for a reactor neutrino experiment. It is determined from thermal power measurements, reactor core simulation, and knowledge of neutrino spectra of fuel isotopes. Past reactor neutrino experiments have determined the flux to (2-3)% precision. Precision measurements of mixing angle $\\theta_{13}$ by reactor neutrino experiments in the coming years will use near-far detector configurations. Most uncertainties from reactor will be canceled out. Understanding of the correlation of uncertainties is required for $\\theta_{13}$ experiments. Precise determination of reactor neutrino flux will also improve the sensitivity of the non-proliferation monitoring and future reactor experiments. We will discuss the flux calculation and recent progresses.

  17. ITERATED INTEGRALS OF SUPERCONNECTIONS KIYOSHI IGUSA

    E-Print Network [OSTI]

    Igusa, Kiyoshi

    ITERATED INTEGRALS OF SUPERCONNECTIONS KIYOSHI IGUSA Abstract. Iterated integration of a connection gives the holonomy or parallel transport of the connection. Iterated integration of a superconnection gives something else which we call a "superconnection parallel transport." We ask under what conditions

  18. Applications of Integral Benchmark Data

    SciTech Connect (OSTI)

    Giuseppe Palmiotti; Teruhiko Kugo; Fitz Trumble; Albert C. (Skip) Kahler; Dale Lancaster

    2014-10-09T23:59:59.000Z

    The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) provide evaluated integral benchmark data that may be used for validation of reactor physics / nuclear criticality safety analytical methods and data, nuclear data testing, advanced modeling and simulation, and safety analysis licensing activities. The handbooks produced by these programs are used in over 30 countries. Five example applications are presented in this paper: (1) Use of IRPhEP Data in Uncertainty Analyses and Cross Section Adjustment, (2) Uncertainty Evaluation Methods for Reactor Core Design at JAEA Using Reactor Physics Experimental Data, (3) Application of Benchmarking Data to a Broad Range of Criticality Safety Problems, (4) Cross Section Data Testing with ICSBEP Benchmarks, and (5) Use of the International Handbook of Evaluated Reactor Physics Benchmark Experiments to Support the Power Industry.

  19. REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS

    SciTech Connect (OSTI)

    Nichols, T.; Beals, D.; Sternat, M.

    2011-07-18T23:59:59.000Z

    Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Many research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and radiochemical analysis. The fuel assembly was modeled using MONTEBURNS(MCNP5/ ORIGEN2.2) and MCNPX/CINDER90. The results from the models have been compared to each other and to the measured data.

  20. Electrochemistry of Water-Cooled Nuclear Reactors

    SciTech Connect (OSTI)

    Macdonald, Dgiby; Urquidi-Macdonald, Mirna; Pitt, Jonathan

    2006-08-08T23:59:59.000Z

    This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or "radiation fields" around the primary loop and the vessel, as a function of the operating parameters and the water chemistry.

  1. Role of nuclear reactors in future military satellites

    SciTech Connect (OSTI)

    Buden, D.; Angelo, J.A. Jr.

    1982-01-01T23:59:59.000Z

    Future military capabilities will be profoundly influenced by emerging Shuttle Era space technology. Regardless of the specific direction or content of tomorrow's military space program, it is clear that advanced space transportation systems, orbital support facilities, and large-capacity power subsystems will be needed to create the generally larger, more sophisticated military space systems of the future. This paper explores the critical role that space nuclear reactors should play in America's future space program and reviews the current state of nuclear reactor power plant technology. Space nuclear reactor technologies have the potential of satisfying power requirements ranging from 10 kW/sub (e)/ to 100 MW/sub (e)/.

  2. Ceramic membrane reactor with two reactant gases at different pressures

    DOE Patents [OSTI]

    Balachandran, Uthamalingam (Hinsdale, IL); Mieville, Rodney L. (Glen Ellyn, IL)

    2001-01-01T23:59:59.000Z

    The invention is a ceramic membrane reactor for syngas production having a reaction chamber, an inlet in the reactor for natural gas intake, a plurality of oxygen permeating ceramic slabs inside the reaction chamber with each slab having a plurality of passages paralleling the gas flow for transporting air through the reaction chamber, a manifold affixed to one end of the reaction chamber for intake of air connected to the slabs, a second manifold affixed to the reactor for removing the oxygen depleted air, and an outlet in the reaction chamber for removing syngas.

  3. Improved vortex reactor system

    DOE Patents [OSTI]

    Diebold, J.P.; Scahill, J.W.

    1995-05-09T23:59:59.000Z

    An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

  4. Tokamak reactor first wall

    DOE Patents [OSTI]

    Creedon, R.L.; Levine, H.E.; Wong, C.; Battaglia, J.

    1984-11-20T23:59:59.000Z

    This invention relates to an improved first wall construction for a tokamak fusion reactor vessel, or other vessels subjected to similar pressure and thermal stresses.

  5. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, Juhani (Karhula, FI)

    1996-01-01T23:59:59.000Z

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.

  6. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, J.

    1996-03-19T23:59:59.000Z

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

  7. Core follow calculation with the nTRACER numerical reactor and verification using power reactor measurement data

    SciTech Connect (OSTI)

    Jung, Y. S.; Joo, H. G. [Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul (Korea, Republic of); Yoon, J. I. [KEPCO Nuclear Fuel, 1047 Daedukdae-ro, Yuseong-gu, Daejeon (Korea, Republic of)

    2013-07-01T23:59:59.000Z

    The nTRACER direct whole core transport code employing the planar MOC solution based 3-D calculation method, the subgroup method for resonance treatment, the Krylov matrix exponential method for depletion, and a subchannel thermal/hydraulic calculation solver was developed for practical high-fidelity simulation of power reactors. Its accuracy and performance is verified by comparing with the measurement data obtained for three pressurized water reactor cores. It is demonstrated that accurate and detailed multi-physic simulation of power reactors is practically realizable without any prior calculations or adjustments. (authors)

  8. Aerosol Resuspension Model for MELCOR for Fusion and Very High Temperature Reactor Applications

    SciTech Connect (OSTI)

    B.J. Merrill

    2011-01-01T23:59:59.000Z

    Dust is generated in fusion reactors from plasma erosion of plasma facing components within the reactor’s vacuum vessel (VV) during reactor operation. This dust collects in cooler regions on interior surfaces of the VV. Because this dust can be radioactive, toxic, and/or chemically reactive, it poses a safety concern, especially if mobilized by the process of resuspension during an accident and then transported as an aerosol though out the reactor confinement building, and possibly released to the environment. A computer code used at the Idaho National Laboratory (INL) to model aerosol transport for safety consequence analysis is the MELCOR code. A primary reason for selecting MELCOR for this application is its aerosol transport capabilities. The INL Fusion Safety Program (FSP) organization has made fusion specific modifications to MELCOR. Recent modifications include the implementation of aerosol resuspension models in MELCOR 1.8.5 for Fusion. This paper presents the resuspension models adopted and the initial benchmarking of these models.

  9. Integration of Nontraditional Isotopic Systems Into Reaction...

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

    Water-Rock Interaction, and Impacts of Water Chemistry on Reservoir Sustainability Integration of Nontraditional Isotopic Systems Into Reaction-Transport Models of EGS For...

  10. Irradiation behavior of metallic fast reactor fuels

    SciTech Connect (OSTI)

    Pahl, R.G.; Porter, D.L.; Crawford, D.C.; Walters, L.C.

    1991-01-01T23:59:59.000Z

    Metallic fuels were the first fuels chosen for liquid metal cooled fast reactors (LMR's). In the late 1960's world-wide interest turned toward ceramic LMR fuels before the full potential of metallic fuel was realized. However, during the 1970's the performance limitations of metallic fuel were resolved in order to achieve a high plant factor at the Argonne National Laboratory's Experimental Breeder Reactor II. The 1980's spawned renewed interest in metallic fuel when the Integral Fast Reactor (IFR) concept emerged at Argonne National Laboratory. A fuel performance demonstration program was put into place to obtain the data needed for the eventual licensing of metallic fuel. This paper will summarize the results of the irradiation program carried out since 1985.

  11. Brookhaven Graphite Research Reactor Workshop

    Broader source: Energy.gov [DOE]

    The Brookhaven Graphite Research Reactor (BGRR) was the first reactor built in the U.S. for peacetime atomic research following World War II.  Construction began in 1947 and the reactor started...

  12. Subsurface Uranium Fate and Transport: Integrated Experiments and Modeling of Coupled Biogeochemical Mechanisms of Nanocrystalline Uraninite Oxidation by Fe(III)-(hydr)oxides - Project Final Report

    SciTech Connect (OSTI)

    Peyton, Brent M. [Montana State University; Timothy, Ginn R. [University of California Davis; Sani, Rajesh K. [South Dakota School of Mines and Technology

    2013-08-14T23:59:59.000Z

    Subsurface bacteria including sulfate reducing bacteria (SRB) reduce soluble U(VI) to insoluble U(IV) with subsequent precipitation of UO2. We have shown that SRB reduce U(VI) to nanometer-sized UO2 particles (1-5 nm) which are both intra- and extracellular, with UO2 inside the cell likely physically shielded from subsequent oxidation processes. We evaluated the UO2 nanoparticles produced by Desulfovibrio desulfuricans G20 under growth and non-growth conditions in the presence of lactate or pyruvate and sulfate, thiosulfate, or fumarate, using ultrafiltration and HR-TEM. Results showed that a significant mass fraction of bioreduced U (35-60%) existed as a mobile phase when the initial concentration of U(VI) was 160 µM. Further experiments with different initial U(VI) concentrations (25 - 900 ?M) in MTM with PIPES or bicarbonate buffers indicated that aggregation of uraninite depended on the initial concentrations of U(VI) and type of buffer. It is known that under some conditions SRB-mediated UO2 nanocrystals can be reoxidized (and thus remobilized) by Fe(III)-(hydr)oxides, common constituents of soils and sediments. To elucidate the mechanism of UO2 reoxidation by Fe(III) (hydr)oxides, we studied the impact of Fe and U chelating compounds (citrate, NTA, and EDTA) on reoxidation rates. Experiments were conducted in anaerobic batch systems in PIPES buffer. Results showed EDTA significantly accelerated UO2 reoxidation with an initial rate of 9.5?M day-1 for ferrihydrite. In all cases, bicarbonate increased the rate and extent of UO2 reoxidation with ferrihydrite. The highest rate of UO2 reoxidation occurred when the chelator promoted UO2 and Fe(III) (hydr)oxide dissolution as demonstrated with EDTA. When UO2 dissolution did not occur, UO2 reoxidation likely proceeded through an aqueous Fe(III) intermediate as observed for both NTA and citrate. To complement to these laboratory studies, we collected U-bearing samples from a surface seep at the Rifle field site and have measured elevated U concentrations in oxic iron-rich sediments. To translate experimental results into numerical analysis of U fate and transport, a reaction network was developed based on Sani et al. (2004) to simulate U(VI) bioreduction with concomitant UO2 reoxidation in the presence of hematite or ferrihydrite. The reduction phase considers SRB reduction (using lactate) with the reductive dissolution of Fe(III) solids, which is set to be microbially mediated as well as abiotically driven by sulfide. Model results show the oxidation of HS– by Fe(III) directly competes with UO2 reoxidation as Fe(III) oxidizes HS– preferentially over UO2. The majority of Fe reduction is predicted to be abiotic, with ferrihydrite becoming fully consumed by reaction with sulfide. Predicted total dissolved carbonate concentrations from the degradation of lactate are elevated (log(pCO2) ~ –1) and, in the hematite system, yield close to two orders-of-magnitude higher U(VI) concentrations than under initial carbonate concentrations of 3 mM. Modeling of U(VI) bioreduction with concomitant reoxidation of UO2 in the presence of ferrihydrite was also extended to a two-dimensional field-scale groundwater flow and biogeochemically reactive transport model for the South Oyster site in eastern Virginia. This model was developed to simulate the field-scale immobilization and subsequent reoxidation of U by a biologically mediated reaction network.

  13. The development of a remote monitoring system for the Nuclear Science Center reactor

    E-Print Network [OSTI]

    Jiltchenkov, Dmitri Victorovich

    2002-01-01T23:59:59.000Z

    With funding provided by Nuclear Energy Research Initiative (NERI), design of Secure, Transportable, Autonomous Reactors (STAR) to aid countries with insufficient energy supplies is underway. The development of a new monitoring system that allows...

  14. The design of a compact integral medium size PWR : the CIRIS

    E-Print Network [OSTI]

    Shirvan, Koroush

    2010-01-01T23:59:59.000Z

    The International Reactor Innovative and Secure (IRIS) is an advanced medium size, modular integral light water reactor design, rated currently at 1000 MWt. IRIS design has been under development by over 20 organizations ...

  15. Integrated environmental degradation model for Fe-Ni-Cr alloys in irradiated aqueous solutions

    E-Print Network [OSTI]

    Pleune, Thomas Todd, 1974-

    1999-01-01T23:59:59.000Z

    An integrated model has been developed to evaluate the effect of reactor flux, fluence, and other operating conditions on crack growth rates in austenitic stainless steels in boiling water reactor (BWR) environments. The ...

  16. Measurements of 222Rn, 220Rn, and CO Emissions in Natural CO2 Fields in Wyoming: MVA Techniques for Determining Gas Transport and Caprock Integrity

    SciTech Connect (OSTI)

    Kaszuba, John; Sims, Kenneth

    2014-09-30T23:59:59.000Z

    An integrated field-laboratory program evaluated the use of radon and CO2 flux measurements to constrain source and timescale of CO2 fluxes in environments proximate to CO2 storage reservoirs. By understanding the type and depth of the gas source, the integrity of a CO2 storage reservoir can be assessed and monitored. The concept is based on correlations of radon and CO2 fluxes observed in volcanic systems. This fundamental research is designed to advance the science of Monitoring, Verification, and Accounting (MVA) and to address the Carbon Storage Program goal of developing and validating technologies to ensure 99 percent storage performance. Graduate and undergraduate students conducted the research under the guidance of the Principal Investigators; in doing so they were provided with training opportunities in skills required for implementing and deploying CCS technologies. Although a final method or “tool” was not developed, significant progress was made. The field program identified issues with measuring radon in environments rich in CO2. Laboratory experiments determined a correction factor to apply to radon measurements made in CO2-bearing environments. The field program also identified issues with radon and CO2-flux measurements in soil gases at a natural CO2 analog. A systematic survey of radon and CO2 flux in soil gases at the LaBarge CO2 Field in Southwest Wyoming indicates that measurements of 222Rn (radon), 220Rn (thoron), and CO2 flux may not be a robust method for monitoring the integrity of a CO2 storage reservoir. The field program was also not able to correlate radon and CO2 flux in the CO2-charged springs of the Thermopolis hydrothermal system. However, this part of the program helped to motivate the aforementioned laboratory experiments that determined correction factors for measuring radon in CO2-rich environments. A graduate student earned a Master of Science degree for this part of the field program; she is currently employed with a geologic consulting company. Measurement of radon in springs has improved significantly since the field program first began; however, in situ measurement of 222Rn and particularly 220Rn in springs is problematic. Future refinements include simultaneous salinity measurements and systematic corrections, or adjustments to the partition coefficient as needed for more accurate radon concentration determination. A graduate student earned a Master of Science degree for this part of the field program; he is currently employed with a geologic consulting company. Both graduate students are poised to begin work in a CCS technology area. Laboratory experiments evaluated important process-level fundamentals that effect measurements of radon and CO2. Laboratory tests established that fine-grained source minerals yield higher radon emissivity compared to coarser-sized source minerals; subtleties in the dataset suggest that grain size alone is not fully representative of all the processes controlling the ability of radon to escape its mineral host. Emissivity for both 222Rn and 220Rn increases linearly with temperature due to reaction of rocks with water, consistent with faster diffusion and enhanced mineral dissolution at higher temperatures. The presence of CO2 changes the relative importance of the factors that control release of radon. Emissivity for both 222Rn and 220Rn in CO2-bearing experiments is greater at all temperatures compared to the experiments without CO2, but emissivity does not increase as a simple function of temperature. Governing processes may include a balance between enhanced dissolution versus carbonate mineral formation in CO2-rich waters.

  17. REACTOR OPERATIONS AND CONTROL

    E-Print Network [OSTI]

    Pázsit, Imre

    REACTOR OPERATIONS AND CONTROL KEYWORDS: core calculations, neural networks, control rod elevation of a control rod, or a group of control rods, is an important parameter from the viewpoint of reactor control DETERMINATION OF PWR CONTROL ROD POSITION BY CORE PHYSICS AND NEURAL NETWORK METHODS NINOS S. GARIS* and IMRE

  18. Reed Reactor Facility. Final report

    SciTech Connect (OSTI)

    Frantz, S.G.

    1994-12-31T23:59:59.000Z

    This report discusses the operation and maintenance of the Reed Reactor Facility. The Reed reactor is mostly used for education and train purposes.

  19. Reactor & Nuclear Systems Publications | ORNL

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

    Science Home | Science & Discovery | Nuclear Science | Publications and Reports | Reactor and Nuclear Systems Publications SHARE Reactor and Nuclear Systems Publications The...

  20. Nanoengineered membranes for controlled transport

    DOE Patents [OSTI]

    Doktycz, Mitchel J. (Oak Ridge, TN) [Oak Ridge, TN; Simpson, Michael L. (Knoxville, TN) [Knoxville, TN; McKnight, Timothy E. (Greenback, TN) [Greenback, TN; Melechko, Anatoli V. (Oak Ridge, TN) [Oak Ridge, TN; Lowndes, Douglas H. (Knoxville, TN) [Knoxville, TN; Guillorn, Michael A. (Knoxville, TN) [Knoxville, TN; Merkulov, Vladimir I. (Oak Ridge, TN) [Oak Ridge, TN

    2010-01-05T23:59:59.000Z

    A nanoengineered membrane for controlling material transport (e.g., molecular transport) is disclosed. The membrane includes a substrate, a cover definining a material transport channel between the substrate and the cover, and a plurality of fibers positioned in the channel and connected to an extending away from a surface of the substrate. The fibers are aligned perpendicular to the surface of the substrate, and have a width of 100 nanometers or less. The diffusion limits for material transport are controlled by the separation of the fibers. In one embodiment, chemical derivitization of carbon fibers may be undertaken to further affect the diffusion limits or affect selective permeability or facilitated transport. For example, a coating can be applied to at least a portion of the fibers. In another embodiment, individually addressable carbon nanofibers can be integrated with the membrane to provide an electrical driving force for material transport.

  1. RSMASS-D: Reactor and shield mass minimization models

    SciTech Connect (OSTI)

    Marshall, A.C. (Department 0410 Sandia National Laboratories (USA) NE-52 The Department of Energy, Germantown Building, Washington, D.C. 20545 (USA)); Gallup, D.R. (Division 6472 Sandia National Laboratories, P.O. Box 5800, Albuquerque, New Mexico 87185 (USA))

    1991-01-01T23:59:59.000Z

    Three relatively simple mathematical models have been developed to estimate minimum reactor and radiation shield masses for liquid metal cooled reactors (LMR's), in-core thermionic reactors (TI's) and out-of-core thermionic reactors (OTR's). The approach was based on much of the methodology developed for the RSMASS model (Marshall 1986). The models use a combination of simple equations derived from reactor physics and other fundamental considerations along with tabulations of data from more detailed neutron and gamma transport theory computations. All three models vary basic design parameters within an allowed range to achieve a parameter choice which yields a minimum mass for the power level and operational time of interest. The impact of critical mass, fuel damage and thermal limitations are accounted for in the computations. Thermionic requirements are also accounted for in the thermionic reactor models. All major reactor component masses are estimated as well as instrumentation and control (I C), boom and safety system mass. A new shield model was developed and incorporated into all three models. The new shield model is more accurate and simpler to use than the approach used in the original RSMASS model. The estimated reactor and shield masses agree with the mass predictions from detailed calculations within 16 percent for all three models.

  2. Reactor dosimetry and RPV life management

    SciTech Connect (OSTI)

    Belousov, S.; Ilieva, K.; Mitev, M. [Inst. for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences, Tsarigradsko 72, 1784 Sofia (Bulgaria)

    2011-07-01T23:59:59.000Z

    Reactor dosimetry (RD) is a tool that provides data for neutron fluence accumulated over the reactor pressure vessel (RPV) during the reactor operation. This information, however, is not sufficient for RPV lifetime assessment. The life management of RPV is a multidisciplinary task. To assess whether the RPV steel properties at the current stage (for actual accumulated neutron fluence) of reactor operation are still 'safe enough,' the dependence of material properties on the fluence must be known; this is a task for material science (MS). Moreover, the mechanical loading over the RPV during normal operation and accidence have to be known, as well, for evaluation, if the RPV material integrity in this loading condition and existing cracks is provided. The crack loading path in terms of stress intensity factor is carried out by structural analyses (SA). Pressure and temperature distribution over RPV used in these analyses are obtained from a thermal hydraulic (TH) calculation. The conjunction of RD and other disciplines in RPV integrity assessment is analyzed in accordance with the FFP (fitness for purpose) approach. It could help to improve the efficiency in multi-disciplinary tasks solutions. (authors)

  3. Impact of conversion to mixed-oxide fuels on reactor structural components

    SciTech Connect (OSTI)

    Yahr, G.T.

    1997-04-01T23:59:59.000Z

    The use of mixed-oxide (MOX) fuel to replace conventional uranium fuel in commercial light-water power reactors will result in an increase in the neutron flux. The impact of the higher flux on the structural integrity of reactor structural components must be evaluated. This report briefly reviews the effects of radiation on the mechanical properties of metals. Aging degradation studies and reactor operating experience provide a basis for determining the areas where conversion to MOX fuels has the potential to impact the structural integrity of reactor components.

  4. Transportation Institutional Plan

    SciTech Connect (OSTI)

    Not Available

    1986-08-01T23:59:59.000Z

    This Institutional Plan is divided into three chapters. Chapter 1 provides background information, discusses the purposes of the Plan and the policy guidance for establishing the transportation system, and describes the projected system and the plans for its integrated development. Chapter 2 discusses the major participants who must interact to build the system. Chapter 3 suggests mechanisms for interaction that will foster wide participation in program planning and implementation and provides a framework for managing and resolving the issues related to development and operation of the transportation system. A list of acronyms and a glossary are included for the reader's convenience. Also included in this Plan are four appendices. Of particular importance is Appendix A, which includes detailed discussion of specific transportation issues. Appendices B, C, and D provide supporting material to assist the reader in understanding the roles of the involved institutions.

  5. Nuclear Reactor Technologies | Department of Energy

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What's Possible for Renewable Energy:Nanowire3627 Federal Register /76 LosExperimentalSecurityReactor

  6. Processes and Procedures for Application of CFD to Nuclear Reactor Safety Analysis

    SciTech Connect (OSTI)

    Richard W. Johnson; Richard R. Schultz; Patrick J. Roache; Ismail B. Celik; William D. Pointer; Yassin A. Hassan

    2006-09-01T23:59:59.000Z

    Traditionally, nuclear reactor safety analysis has been performed using systems analysis codes such as RELAP5, which was developed at the INL. However, goals established by the Generation IV program, especially the desire to increase efficiency, has lead to an increase in operating temperatures for the reactors. This increase pushes reactor materials to operate towards their upper temperature limits relative to structural integrity. Because there will be some finite variation of the power density in the reactor core, there will be a potential for local hot spots to occur in the reactor vessel. Hence, it has become apparent that detailed analysis will be required to ensure that local ‘hot spots’ do not exceed safety limits. It is generally accepted that computational fluid dynamics (CFD) codes are intrinsically capable of simulating fluid dynamics and heat transport locally because they are based on ‘first principles.’ Indeed, CFD analysis has reached a fairly mature level of development, including the commercial level. However, CFD experts are aware that even though commercial codes are capable of simulating local fluid and thermal physics, great care must be taken in their application to avoid errors caused by such things as inappropriate grid meshing, low-order discretization schemes, lack of iterative convergence and inaccurate time-stepping. Just as important is the choice of a turbulence model for turbulent flow simulation. Turbulence models model the effects of turbulent transport of mass, momentum and energy, but are not necessarily applicable for wide ranges of flow types. Therefore, there is a well-recognized need to establish practices and procedures for the proper application of CFD to simulate flow physics accurately and establish the level of uncertainty of such computations. The present document represents contributions of CFD experts on what the basic practices, procedures and guidelines should be to aid CFD analysts to obtain accurate estimates of the flow and energy transport as applied to nuclear reactor safety. However, it is expected that these practices and procedures will require updating from time to time as research and development affect them or replace them with better procedures. The practices and procedures are categorized into five groups. These are: 1.Code Verification 2.Code and Calculation Documentation 3.Reduction of Numerical Error 4.Quantification of Numerical Uncertainty (Calculation Verification) 5.Calculation Validation. These five categories have been identified from procedures currently required of CFD simulations such as those required for publication of a paper in the ASME Journal of Fluids Engineering and from the literature such as Roache [1998]. Code verification refers to the demonstration that the equations of fluid and energy transport have been correctly coded in the CFD code. Code and calculation documentation simply means that the equations and their discretizations, etc., and boundary and initial conditions used to pose the fluid flow problem are fully described in available documentation. Reduction of numerical error refers to practices and procedures to lower numerical errors to negligible or very low levels as is reasonably possible (such as avoiding use of first-order discretizations). The quantification of numerical uncertainty is also known as calculation verification. This means that estimates are made of numerical error to allow the characterization of the numerical

  7. Transportation Agency Tool to Analyze Benefits of Living Snow Fences

    E-Print Network [OSTI]

    Minnesota, University of

    Transportation Agency Tool to Analyze Benefits of Living Snow Fences 5/31/12 Transportation Agency/31/12 Transportation Agency Tool to Analyze Benefits of Living Snow Fences Center for Integrated Natural Resources, Mobility, & Transportation Authority Benefits, Farmer Costs, & Carbon Impacts Focus Groups and Surveys

  8. Nuclear reactor control column

    SciTech Connect (OSTI)

    Bachovchin, D.M.

    1982-08-10T23:59:59.000Z

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest crosssectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

  9. Nuclear reactor control column

    DOE Patents [OSTI]

    Bachovchin, Dennis M. (Plum Borough, PA)

    1982-01-01T23:59:59.000Z

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

  10. Fast Breeder Reactor studies

    SciTech Connect (OSTI)

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01T23:59:59.000Z

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  11. Spherical torus fusion reactor

    DOE Patents [OSTI]

    Martin Peng, Y.K.M.

    1985-10-03T23:59:59.000Z

    The object of this invention is to provide a compact torus fusion reactor with dramatic simplification of plasma confinement design. Another object of this invention is to provide a compact torus fusion reactor with low magnetic field and small aspect ratio stable plasma confinement. In accordance with the principles of this invention there is provided a compact toroidal-type plasma confinement fusion reactor in which only the indispensable components inboard of a tokamak type of plasma confinement region, mainly a current conducting medium which carries electrical current for producing a toroidal magnet confinement field about the toroidal plasma region, are retained.

  12. Nuclear reactor reflector

    DOE Patents [OSTI]

    Hopkins, Ronald J. (Pensacola, FL); Land, John T. (Pensacola, FL); Misvel, Michael C. (Pensacola, FL)

    1994-01-01T23:59:59.000Z

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled.

  13. Nuclear reactor reflector

    DOE Patents [OSTI]

    Hopkins, R.J.; Land, J.T.; Misvel, M.C.

    1994-06-07T23:59:59.000Z

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled. 12 figs.

  14. Microfluidic electrochemical reactors

    DOE Patents [OSTI]

    Nuzzo, Ralph G. (Champaign, IL); Mitrovski, Svetlana M. (Urbana, IL)

    2011-03-22T23:59:59.000Z

    A microfluidic electrochemical reactor includes an electrode and one or more microfluidic channels on the electrode, where the microfluidic channels are covered with a membrane containing a gas permeable polymer. The distance between the electrode and the membrane is less than 500 micrometers. The microfluidic electrochemical reactor can provide for increased reaction rates in electrochemical reactions using a gaseous reactant, as compared to conventional electrochemical cells. Microfluidic electrochemical reactors can be incorporated into devices for applications such as fuel cells, electrochemical analysis, microfluidic actuation, pH gradient formation.

  15. Homogenization of the criticality spectral equation in neutron transport

    E-Print Network [OSTI]

    Bal, Guillaume

    for the neutron transport equation in a periodic heterogeneous domain, modeling the criticality study of nuclearHomogenization of the criticality spectral equation in neutron transport Gr'egoire Allaire \\Lambda problem. This result justifies and improves the engineering procedure used in practice for nuclear reactor

  16. Simulation Analysis Center for Transportation Analysis

    E-Print Network [OSTI]

    policy makers in the planning for the loading of spent nuclear fuel at commercial utility reactors Oak Ridge National Laboratory managed by UT-Battelle, LLC for the U.S. Department of Energy under experience in simulation modeling for logistics, transportation, security and supply chain management. ORNL

  17. EBS Radionuclide Transport Abstraction

    SciTech Connect (OSTI)

    J.D. Schreiber

    2005-08-25T23:59:59.000Z

    The purpose of this report is to develop and analyze the engineered barrier system (EBS) radionuclide transport abstraction model, consistent with Level I and Level II model validation, as identified in ''Technical Work Plan for: Near-Field Environment and Transport: Engineered Barrier System: Radionuclide Transport Abstraction Model Report Integration'' (BSC 2005 [DIRS 173617]). The EBS radionuclide transport abstraction (or EBS RT Abstraction) is the conceptual model used in the total system performance assessment for the license application (TSPA-LA) to determine the rate of radionuclide releases from the EBS to the unsaturated zone (UZ). The EBS RT Abstraction conceptual model consists of two main components: a flow model and a transport model. Both models are developed mathematically from first principles in order to show explicitly what assumptions, simplifications, and approximations are incorporated into the models used in the TSPA-LA. The flow model defines the pathways for water flow in the EBS and specifies how the flow rate is computed in each pathway. Input to this model includes the seepage flux into a drift. The seepage flux is potentially split by the drip shield, with some (or all) of the flux being diverted by the drip shield and some passing through breaches in the drip shield that might result from corrosion or seismic damage. The flux through drip shield breaches is potentially split by the waste package, with some (or all) of the flux being diverted by the waste package and some passing through waste package breaches that might result from corrosion or seismic damage. Neither the drip shield nor the waste package survives an igneous intrusion, so the flux splitting submodel is not used in the igneous scenario class. The flow model is validated in an independent model validation technical review. The drip shield and waste package flux splitting algorithms are developed and validated using experimental data. The transport model considers advective transport and diffusive transport from a breached waste package. Advective transport occurs when radionuclides that are dissolved or sorbed onto colloids (or both) are carried from the waste package by the portion of the seepage flux that passes through waste package breaches. Diffusive transport occurs as a result of a gradient in radionuclide concentration and may take place while advective transport is also occurring, as well as when no advective transport is occurring. Diffusive transport is addressed in detail because it is the sole means of transport when there is no flow through a waste package, which may dominate during the regulatory compliance period in the nominal and seismic scenarios. The advective transport rate, when it occurs, is generally greater than the diffusive transport rate. Colloid-facilitated advective and diffusive transport is also modeled and is presented in detail in Appendix B of this report.

  18. Deterministic Modeling of the High Temperature Test Reactor

    SciTech Connect (OSTI)

    Ortensi, J.; Cogliati, J. J.; Pope, M. A.; Ferrer, R. M.; Ougouag, A. M.

    2010-06-01T23:59:59.000Z

    Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability of the Next Generation Nuclear Power (NGNP) project. In order to examine INL’s current prismatic reactor deterministic analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19 column thin annular core, and the fully loaded core critical condition with 30 columns. Special emphasis is devoted to the annular core modeling, which shares more characteristics with the NGNP base design. The DRAGON code is used in this study because it offers significant ease and versatility in modeling prismatic designs. Despite some geometric limitations, the code performs quite well compared to other lattice physics codes. DRAGON can generate transport solutions via collision probability (CP), method of characteristics (MOC), and discrete ordinates (Sn). A fine group cross section library based on the SHEM 281 energy structure is used in the DRAGON calculations. HEXPEDITE is the hexagonal z full core solver used in this study and is based on the Green’s Function solution of the transverse integrated equations. In addition, two Monte Carlo (MC) based codes, MCNP5 and PSG2/SERPENT, provide benchmarking capability for the DRAGON and the nodal diffusion solver codes. The results from this study show a consistent bias of 2–3% for the core multiplication factor. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The ENDF/B VII graphite and U235 cross sections appear to be the main source of the error. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement stems from the fact that during the experiments the control rods were adjusted to maintain criticality, whereas in the model, the rod positions were fixed. In addition, this work includes a brief study of a cross section generation approach that seeks to decouple the domain in order to account for neighbor effects. This spectral interpenetration is a dominant effect in annular HTR physics. This analysis methodology should be further explored in order to reduce the error that is systematically propagated in the traditional generation of cross sections.

  19. Reactor pressure vessel vented head

    DOE Patents [OSTI]

    Sawabe, J.K.

    1994-01-11T23:59:59.000Z

    A head for closing a nuclear reactor pressure vessel shell includes an arcuate dome having an integral head flange which includes a mating surface for sealingly mating with the shell upon assembly therewith. The head flange includes an internal passage extending therethrough with a first port being disposed on the head mating surface. A vent line includes a proximal end disposed in flow communication with the head internal passage, and a distal end disposed in flow communication with the inside of the dome for channeling a fluid therethrough. The vent line is fixedly joined to the dome and is carried therewith when the head is assembled to and disassembled from the shell. 6 figures.

  20. Production capabilities in US nuclear reactors for medical radioisotopes

    SciTech Connect (OSTI)

    Mirzadeh, S.; Callahan, A.P.; Knapp, F.F. Jr. [Oak Ridge National Lab., TN (United States); Schenter, R.E. [Westinghouse Hanford Co., Richland, WA (United States)

    1992-11-01T23:59:59.000Z

    The availability of reactor-produced radioisotopes in the United States for use in medical research and nuclear medicine has traditionally depended on facilities which are an integral part of the US national laboratories and a few reactors at universities. One exception is the reactor in Sterling Forest, New York, originally operated as part of the Cintichem (Union Carbide) system, which is currently in the process of permanent shutdown. Since there are no industry-run reactors in the US, the national laboratories and universities thus play a critical role in providing reactor-produced radioisotopes for medical research and clinical use. The goal of this survey is to provide a comprehensive summary of these production capabilities. With the temporary shutdown of the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) in November 1986, the radioisotopes required for DOE-supported radionuclide generators were made available at the Brookhaven National Laboratory (BNL) High Flux Beam Reactor (HFBR). In March 1988, however, the HFBR was temporarily shut down which forced investigators to look at other reactors for production of the radioisotopes. During this period the Missouri University Research Reactor (MURR) played an important role in providing these services. The HFIR resumed routine operation in July 1990 at 85 MW power, and the HFBR resumed operation in June 1991, at 30 MW power. At the time of the HFBR shutdown, there was no available comprehensive overview which could provide information on status of the reactors operating in the US and their capabilities for radioisotope production. The obvious need for a useful overview was thus the impetus for preparing this survey, which would provide an up-to-date summary of those reactors available in the US at both the DOE-funded national laboratories and at US universities where service irradiations are currently or expected to be conducted.

  1. Parametric study on maximum transportable distance and cost for thermal energy transportation using various coolants

    SciTech Connect (OSTI)

    Su-Jong Yoon; Piyush Sabharwall

    2014-07-01T23:59:59.000Z

    The operation temperature of advanced nuclear reactors is generally higher than commercial light water reactors and thermal energy from advanced nuclear reactor can be used for various purposes such as district heating, desalination, hydrogen production and other process heat applications, etc. The process heat industry/facilities will be located outside the nuclear island due to safety measures. This thermal energy from the reactor has to be transported a fair distance. In this study, analytical analysis was conducted to identify the maximum distance that thermal energy could be transported using various coolants such as molten-salts, helium and water by varying the pipe diameter and mass flow rate. The cost required to transport each coolant was also analyzed. The coolants analyzed are molten salts (such as: KClMgCl2, LiF-NaF-KF (FLiNaK) and KF-ZrF4), helium and water. Fluoride salts are superior because of better heat transport characteristics but chloride salts are most economical for higher temperature transportation purposes. For lower temperature water is a possible alternative when compared with He, because low pressure He requires higher pumping power which makes the process very inefficient and economically not viable for both low and high temperature application.

  2. Hypothetical Reactor Accident Study

    E-Print Network [OSTI]

    POPULATIONS; IODINE 131; MELTDOWN; METEOROLOGY; NUCLEAR POWER PLANTS; P CODES; PWR TYPE REACTORS; RADIATION in a Typical BWR and in a typical PWR. Comparison with WASH-1400 by C F . Højerup 202 APPENDIX 3. Calculation

  3. P Reactor Grouting

    SciTech Connect (OSTI)

    None

    2010-01-01T23:59:59.000Z

    Filling the P Reactor with grout. This seals the radioactive material and reduces the environmental footprint left from the Cold War. Project sponsored by the Recovery Act at the Savannah River Site.

  4. Nuclear reactor control

    SciTech Connect (OSTI)

    Ingham, R.V.

    1980-01-01T23:59:59.000Z

    A liquid metal cooled fast breeder nuclear reactor has power setback means for use in an emergency. On initiation of a trip-signal a control rod is injected into the core in two stages, firstly, by free fall to effect an immediate power-set back to a safe level and, secondly, by controlled insertion. Total shut-down of the reactor under all emergencies is avoided. 4 claims.

  5. Polymerization reactor control

    SciTech Connect (OSTI)

    Ray, W.H.

    1985-01-01T23:59:59.000Z

    The principal difficulties in achieving good control of polymerization reactors are related to inadequate on-line measurement, a lack of understanding of the dynamics of the process, the highly sensitive and nonlinear behavior of these reactors, and the lack of well-developed techniques for the control of nonlinear processes. Some illustrations of these problems and a discussion of potential techniques for overcoming some of these difficulties is provided.

  6. Molten metal reactors

    DOE Patents [OSTI]

    Bingham, Dennis N; Klingler, Kerry M; Turner, Terry D; Wilding, Bruce M

    2013-11-05T23:59:59.000Z

    A molten metal reactor for converting a carbon material and steam into a gas comprising hydrogen, carbon monoxide, and carbon dioxide is disclosed. The reactor includes an interior crucible having a portion contained within an exterior crucible. The interior crucible includes an inlet and an outlet; the outlet leads to the exterior crucible and may comprise a diffuser. The exterior crucible may contain a molten alkaline metal compound. Contained between the exterior crucible and the interior crucible is at least one baffle.

  7. Reactor- Nuclear Science Center 

    E-Print Network [OSTI]

    Unknown

    2011-08-17T23:59:59.000Z

    A COMPARISON OF NUCLEAR REACTOR CONTROL ROOM DISPLAY PANELS A Thesis by FRANCES RENAE BOWERS Submitted to the Graduate College of Texas A&M University in partial fulfillment of the requirements for the degree of MASTER OF SCIENCE May 1988... Major Subject: Industrial Engineering A COMPARISON OF NUCLEAR REACTOR CONTROL ROOM DISPLAY PANELS A Thesis by FRANCES RENAE BOWERS Approved as to style and content by: Rod er . oppa (Cha' of 'ttee) R. Quinn Brackett (Member) rome . Co gleton...

  8. Transportation Market Distortions

    E-Print Network [OSTI]

    Litman, Todd

    2006-01-01T23:59:59.000Z

    of Highways, Volpe National Transportation Systems Center (Evaluating Criticism of Transportation Costing, VictoriaFrom Here: Evaluating Transportation Diversity, Victoria

  9. A vectorized heat transfer model for solid reactor cores

    SciTech Connect (OSTI)

    Rider, W.J.; Cappiello, M.W.; Liles, D.R.

    1990-01-01T23:59:59.000Z

    The new generation of nuclear reactors includes designs that are significantly different from light water reactors. Among these new reactor designs is the Modular High-Temperature Gas-Cooled Reactor (MHTGR). In addition, nuclear thermal rockets share a number of similarities with terrestrial HTGRs and would be amenable to similar types of analyses. In these reactors, the heat transfer in the solid core mass is of primary interest in design and safety assessment. One significant safety feature of these reactors is the capability to withstand a loss of pressure and forced cooling in the primary system and still maintain peak fuel temperatures below the safe threshold for retaining the fission products. To accurately assess the performance of gas-cooled reactors during these types of transients, a Helium/Hydrogen Cooled Reactor Analysis (HERA) computer code has been developed. HERA has the ability to model arbitrary geometries in three dimensions, which allows the user to easily analyze reactor cores constructed of prismatic graphite elements. The code accounts for heat generation in the fuel, control rods and other structures; conduction and radiation across gaps; convection to the coolant; and a variety of boundary conditions. The numerical solution scheme has been optimized for vector computers, making long transient analyses economical. Time integration is either explicit or implicit, which allows the use of the model to accurately calculate both short- or long-term transients with an efficient use of computer time. Both the basic spatial and temporal integration schemes have been benchmarked against analytical solutions. Also, HERA has been used to analyze a depressurized loss of forced cooling transient in a HTGR with a very detailed three-dimensional input model. The results compare favorably with other means of analysis and provide further validation of the models and methods. 18 refs., 11 figs.

  10. activity transport: Topics by E-print Network

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

    and social welfare, equity, human health and ecological integrity. A narrow definition of sustainable transport tends to favour individual technological solutions, while a broader...

  11. Transportation Shock and Vibration Literature Review

    SciTech Connect (OSTI)

    Maheras, Steven J.; Lahti, Erik A.; Ross, Steven B.

    2013-06-06T23:59:59.000Z

    This report fulfills the M4 milestone M4FT-13OR08220112, "Report Documenting Experimental Activities." The purpose of this report is to document the results of a literature review conducted of studies related to the vibration and shock associated with the normal conditions of transport for rail shipments of used nuclear fuel from commercial light-water reactors. As discussed in Adkins (2013), the objective of this report is to determine if adequate data exist that would enable the impacts of the shock and vibration associated with the normal conditions of transport on commercial light-water reactor used nuclear fuel shipped in current generation rail transportation casks to be realistically modeled.

  12. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    SciTech Connect (OSTI)

    P. Delmolino

    2005-05-06T23:59:59.000Z

    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  13. Zero Power Reactor simulation | Argonne National Laboratory

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

    Zero Power Reactor simulation Share Description Ever wanted to see a nuclear reactor core in action? Here's a detailed simulation of the Zero Power Reactor experiment, run by...

  14. The development of an integrated multistage fluid bed retorting process. [Kentort II process

    SciTech Connect (OSTI)

    Carter, S.D.; Taulbee, D.N.; Robl, T.L.; Hower, J.C.

    1992-08-01T23:59:59.000Z

    This report summarizes the progress made on the development of an integrated multistage fluidized bed retorting process (KENTORT II) during the period of April 1, 1992 through June 30, 1992. The KENTORT II process includes integral fluidized bed zones for pyrolysis, gasification, and combustion of the oil shale. The purpose of this program is to design and test the KENTORT II process at the 50-lb/hr scale. The raw oil shale sample for the program was mined, prepared, characterized and stored this quarter. The shale that was chosen was from the high-grade zone of the Devonian Cleveland Member of the Ohio Shale in Montgomery County, Kentucky. The shale was mined and then transported to the contractor's crushing facility where it was crushed, double-screened, and loaded into 85 55-gal barrels. The barrels, containing a total of 25-30 tons of shale, were transported to the (CAER) Center for Applied Energy Research where the shale was double-screened, analyzed and stored. A major objective of the program is the study of solid-induced secondary coking and cracking reactions. A valved fluidized bed reactor has been the primary apparatus used for this study prior to this quarter, but two additional techniques have been initiated this quarter for the study of other aspects of this issue. First, the two-stage hydropyrolysis reactor at the University of Strathclyde, Glasgow, Scotland, was used to study the coking tendency of shale oil vapors under a wide range of pyrolysis and hydropyrolysis conditions. This work enabled us to examine secondary reactions under high pressure conditions (up to 150 bar) which were previously unavailable. Second, the development of a fixed bed reactor system was initiated at the CAER to study the coking and cracking characteristics of model compounds. A fixed bed apparatus was necessary because the conversion of model compounds was too low in the fluidized bed apparatus.

  15. Current Abstracts Nuclear Reactors and Technology

    SciTech Connect (OSTI)

    Bales, J.D.; Hicks, S.C. [eds.

    1993-01-01T23:59:59.000Z

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  16. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

    SciTech Connect (OSTI)

    Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

    2010-09-01T23:59:59.000Z

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.

  17. Decision Models for Bulk Energy Transportation Networks

    E-Print Network [OSTI]

    Tesfatsion, Leigh

    emissions prices? How would CO2 regulations impact coal, gas, electricity, & SO2 markets? 3. Disruptions1 Decision Models for Bulk Energy Transportation Networks Electrical Engineering Professor Jim Mc: · integrated fuel, electricity networks · environmental impacts · electricity commodity markets · behavior

  18. Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

    SciTech Connect (OSTI)

    Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E. [Sandia National Labs., Albuquerque, NM (United States)] Sandia National Labs., Albuquerque, NM (United States); Tills, J. [J. Tills and Associates, Inc., Sandia Park, NM (United States)] J. Tills and Associates, Inc., Sandia Park, NM (United States)

    1997-12-01T23:59:59.000Z

    The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.

  19. Washington: Integrated Transportation Programs & Coordinated Regional

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA group currentBradley Nickell Director ofDepartment ofof EnergyPond |100

  20. Integration for Seamless Transport | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are8COaBulkTransmissionSitingProcess.pdfGetecGtel JumpCounty,Jump7OpenInnovativeTechnologies IFCTSensing

  1. Transportation and Stationary Power Integration: Workshop Proceedings |

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn'tOriginEducationVideoStrategic|IndustrialCenter GetsEnergy

  2. Washington: Integrated Transportation Programs & Coordinated Regional

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your DensityEnergy U.S.-China Electric Vehicle and03/02Report |toVEHICLEof Energy Success Story-AFederal

  3. Supertruck - Improving Transportation Efficiency through Integrated

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn'tOriginEducationVideoStrategic Safety GoalsEnergy BeginsSupercomputingTractor &

  4. Methanation assembly using multiple reactors

    DOE Patents [OSTI]

    Jahnke, Fred C.; Parab, Sanjay C.

    2007-07-24T23:59:59.000Z

    A methanation assembly for use with a water supply and a gas supply containing gas to be methanated in which a reactor assembly has a plurality of methanation reactors each for methanating gas input to the assembly and a gas delivery and cooling assembly adapted to deliver gas from the gas supply to each of said methanation reactors and to combine water from the water supply with the output of each methanation reactor being conveyed to a next methanation reactor and carry the mixture to such next methanation reactor.

  5. Reactor pressure vessel. Status report

    SciTech Connect (OSTI)

    Elliot, B.J.; Hackett, E.M.; Lee, A.D. [and others

    1996-10-01T23:59:59.000Z

    This report describes the issues raised as a result of the staffs review of Generic Letter (GL) 92-01, Revision 1, responses and plant-specific reactor pressure vessel (RPV) assessments and the actions taken or work in progress to address these issues. In addition, the report describes actions taken by the staff and the nuclear industry to develop a thermal annealing process for use at U.S. commercial nuclear power plants. This process is intended to be used as a means of mitigating the effects of neutron radiation on the fracture toughness of RPV materials. The Nuclear Regulatory Commission (NRC) issued GL 92-01, Revision 1, Supplement 1, to obtain information needed to assess compliance with regulatory requirements and licensee commitments regarding RPV integrity. GL 92-01, Revision 1, Supplement 1, was issued as a result of generic issues that were raised in the NRC staff`s reviews of licensee responses to GL 92-01, Revision 1, and plant-specific RPV evaluations. In particular, an integrated review of all data submitted in response to GL 92-01, Revision 1, indicated that licensees may not have considered all relevant data in their RPV assessments. This report is representative of submittals to and evaluations by the staff as of September 30, 1996. An update of this report will be issued at a later date.

  6. Regional groundwater flow model for C, K. L. and P reactor areas, Savannah River Site, Aiken, SC

    SciTech Connect (OSTI)

    Flach, G.P.

    2000-02-11T23:59:59.000Z

    A regional groundwater flow model encompassing approximately 100 mi2 surrounding the C, K, L, and P reactor areas has been developed. The reactor flow model is designed to meet the planning objectives outlined in the General Groundwater Strategy for Reactor Area Projects by providing a common framework for analyzing groundwater flow, contaminant migration and remedial alternatives within the Reactor Projects team of the Environmental Restoration Department. The model provides a quantitative understanding of groundwater flow on a regional scale within the near surface aquifers and deeper semi-confined to confined aquifers. The model incorporates historical and current field characterization data up through Spring 1999. Model preprocessing is automated so that future updates and modifications can be performed quickly and efficiently. The CKLP regional reactor model can be used to guide characterization, perform scoping analyses of contaminant transport, and serve as a common base for subsequent finer-scale transport and remedial/feasibility models for each reactor area.

  7. Natural Fueling of a Tokamak Fusion Reactor

    E-Print Network [OSTI]

    Wan, Weigang; Chen, Yang; Perkins, Francis W

    2009-01-01T23:59:59.000Z

    A natural fueling mechanism that helps to maintain the main core deuterium and tritium (DT) density profiles in a tokamak fusion reactor is discussed. In H-mode plasmas dominated by ion- temperature gradient (ITG) driven turbulence, cold DT ions near the edge will naturally pinch radially inward towards the core. This mechanism is due to the quasi-neutral heat flux dominated nature of ITG turbulence and still applies when trapped and passing kinetic electron effects are included. Fueling using shallow pellet injection or supersonic gas jets is augmented by an inward pinch of could DT fuel. The natural fueling mechanism is demonstrated using the three-dimensional toroidal electromagnetic gyrokinetic turbulence code GEM and is analyzed using quasilinear theory. Profiles similar to those used for conservative ITER transport modeling that have a completely flat density profile are examined and it is found that natural fueling actually reduces the linear growth rates and energy transport.

  8. A TEN MEGAWATT BOILING HETEROGENEOUS PACKAGE POWER REACTOR. Reactor...

    Office of Scientific and Technical Information (OSTI)

    A reactor and associated power plant designed to produce 1.05 Mwh and 3.535 Mwh of steam for heating purposes are described. The total thermal output of the reactor is 10 Mwh....

  9. Predicting Reactor Antineutrino Emissions Using New Precision Beta Spectroscopy

    SciTech Connect (OSTI)

    Asner, David M.; Burns, Kimberly A.; Greenfield, Bryce A.; Kos, Marek S.; Orrell, John L.; Schram, Malachi; VanDevender, Brent A.; Wootan, David W.

    2013-05-01T23:59:59.000Z

    Neutrino experiments at nuclear reactors are currently vital to the study of neutrino oscillations. The observed antineutrino rates at reactors are typically lower than model expectations. This observed deficit is called the “reactor neutrino anomaly”. A new understanding of neutrino physics may be required to explain this deficit, though model estimation uncertainties may also play a role in the apparent discrepancy. PNNL is currently investigating an experimental technique that promises reduced uncertainties for measured data to support these hypotheses and interpret reactor antineutrino measurements. The experimental approach is to 1) direct a proton accelerator beam on a metal target to produce a source of neutrons, 2) use spectral tailoring to modify the neutron spectrum to closely simulate the energy distribution of a power reactor neutron spectrum, 3) irradiate isotopic fission foils (235U, 238U, 239Pu, 241Pu) in this neutron spectrum so that fissions occur at energies representative of a reactor, 4) transport the beta particles released by the fission products in the foils to a beta spectrometer, 5) measure the beta energy spectrum, and 6) invert the measured beta energy spectrum to an antineutrino energy spectrum. A similar technique using a beta spectrometer and isotopic fission foils was pioneered in the 1980’s at the ILL thermal reactor. Those measurements have been the basis for interpreting all subsequent antineutrino measurements at reactors. A basic constraint in efforts to reduce uncertainties in predicting the antineutrino emission from reactor cores is any underlying limitation of the original measurements. This may include beta spectrum energy resolution, the absolute normalization of beta emission to number of fission, statistical counting uncertainties, lack of 238U data, the purely thermal nature of the IIL reactor neutrons used, etc. An accelerator-based neutron source that can be tailored to match various reactor neutron spectra provides an advantage for control in studying how changes in the neutron spectra (i.e. "in the reactor core") affects the resulting fission product beta spectrum. Furthermore, the 238U antineutrino spectrum, which has not been measured, can be studied directly because of the enhanced 1 MeV fast neutron flux available at the accelerator source. A facility such as the Project X Injector Experiment (PXIE) 30 MeV proton linear accelerator at Fermilab is being considered for this experiment. The hypothesis is that a new approach utilizing the flexibility of an accelerator neutron source with spectral tailoring coupled with a careful design of an isotopic fission target and beta spectrometer and the inversion of the beta spectrum to the neutrino spectrum will allow further reduction in the uncertainties associated with prediction of the reactor antineutrino spectrum.

  10. Reactor for exothermic reactions

    DOE Patents [OSTI]

    Smith, Jr., Lawrence A. (Bellaire, TX); Hearn, Dennis (Houston, TX); Jones, Jr., Edward M. (Friendswood, TX)

    1993-01-01T23:59:59.000Z

    A liquid phase process for oligomerization of C.sub.4 and C.sub.5 isoolefins or the etherification thereof with C.sub.1 to C.sub.6 alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120.degree. to 300.degree. F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

  11. Thermionic Reactor Design Studies

    SciTech Connect (OSTI)

    Schock, Alfred

    1994-08-01T23:59:59.000Z

    Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their (thermionic reactor) performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling.

  12. Reactor for exothermic reactions

    DOE Patents [OSTI]

    Smith, L.A. Jr.; Hearn, D.; Jones, E.M. Jr.

    1993-03-02T23:59:59.000Z

    A liquid phase process is described for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

  13. Nuclear reactor safety device

    DOE Patents [OSTI]

    Hutter, Ernest (Wilmette, IL)

    1986-01-01T23:59:59.000Z

    A safety device is disclosed for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of an upward thermal excursion. Such safety device comprises a laminated helical ribbon configured as a tube-like helical coil having contiguous helical turns with slidably abutting edges. The helical coil is disclosed as a portion of a drive member connected axially to the control rod. The laminated ribbon is formed of outer and inner laminae. The material of the outer lamina has a greater thermal coefficient of expansion than the material of the inner lamina. In the event of an upward thermal excursion, the laminated helical coil curls inwardly to a smaller diameter. Such inward curling causes the total length of the helical coil to increase by a substantial increment, so that the control rod is axially repositioned by a corresponding amount to reduce the power output of the reactor.

  14. Fusion reactor control

    SciTech Connect (OSTI)

    Plummer, D.A.

    1995-12-31T23:59:59.000Z

    The plasma kinetic temperature and density changes, each per an injected fuel density rate increment, control the energy supplied by a thermonuclear fusion reactor in a power production cycle. This could include simultaneously coupled control objectives for plasma current, horizontal and vertical position, shape and burn control. The minimum number of measurements required, use of indirect (not plasma parameters) system measurements, and distributed control procedures for burn control are to be verifiable in a time dependent systems code. The International Thermonuclear Experimental Reactor (ITER) has the need to feedback control both the fusion output power and the driven plasma current, while avoiding damage to diverter plates. The system engineering of fusion reactors must be performed to assure their development expeditiously and effectively by considering reliability, availability, maintainability, environmental impact, health and safety, and cost.

  15. Simulator platform for fast reactor operation and safety technology demonstration

    SciTech Connect (OSTI)

    Vilim, R. B.; Park, Y. S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J. (Nuclear Engineering Division)

    2012-07-30T23:59:59.000Z

    A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.

  16. Design study of the deep-sea reactor X

    SciTech Connect (OSTI)

    Iida, Hiromasa (Japan Atomic Energy Research Inst., Ibaraki (Japan)); Ishizaka, Yuichi (Mitsubishi Atomic Power Industries, Inc., Tokyo (Japan)); Kim, Y.C.; Yamaguchi, Chouichi (Japan Research Inst., Ltd., Tokyo (Japan))

    1994-07-01T23:59:59.000Z

    The deep-sea reactor X (DRX) is a small nuclear plant designed to provide undersea power sources. It has the full advantages of nuclear reactors and can provide large power capacity and does not require oxygen for power production. An application conceivable in the near future is that for a submersible. The Japan Atomic Energy Research Institute is conducting a design study of a 150-kW(electric) DRX plant for a deep-sea research vessel. It has a so-called integrated pressurized water reactor,'' having a steam generator inside the reactor vessel. A pressure shell includes a turbine and a generator as well as a reactor vessel, resulting in a very compact electricity producing plant. It should be easy to operate and have high passive safety characteristics; namely, a short startup time, good reactor response to power demand changes, and passive core flooding and decay heat removal in case of an accident. Transient analyses including those for load follow-up, reactor startup, and accidents have been conducted. The results show that the DRX has excellent inherent characteristics satisfying those requirements.

  17. Reactor safety research programs. Quarterly report Apr-Jun 81

    SciTech Connect (OSTI)

    Edler, S.K.

    1981-09-01T23:59:59.000Z

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from April 1 through June 30, 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 pipe-to-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-of-coolant 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.

  18. Reactor Safety Research Programs Quarterly Report April- June 1981

    SciTech Connect (OSTI)

    Edler, S. K.

    1981-09-01T23:59:59.000Z

    This document summarizes the work performed by Pacific Northwest laboratory (PNL} from April1 through June 30, 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 pipe-to-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-of-coolant 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, lspra, 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.

  19. Closure head for a nuclear reactor

    DOE Patents [OSTI]

    Wade, Elman E. (South Huntingdon, PA)

    1980-01-01T23:59:59.000Z

    A closure head for a nuclear reactor includes a stationary outer ring integral with the reactor vessel with a first rotatable plug disposed within the stationary outer ring and supported from the stationary outer ring by a bearing assembly. A sealing system is associated with the bearing assembly to seal the annulus defined between the first rotatable plug and the stationary outer ring. The sealing system comprises tubular seal elements disposed in the annulus with load springs contacting the tubular seal elements so as to force the tubular seal elements against the annulus in a manner to seal the annulus. The sealing system also comprises a sealing fluid which is pumped through the annulus and over the tubular seal elements causing the load springs to compress thereby reducing the friction between the tubular seal elements and the rotatable components while maintaining a gas-tight seal therebetween.

  20. Fast quench reactor method

    DOE Patents [OSTI]

    Detering, Brent A. (Idaho Falls, ID); Donaldson, Alan D. (Idaho Falls, ID); Fincke, James R. (Idaho Falls, ID); Kong, Peter C. (Idaho Falls, ID); Berry, Ray A. (Idaho Falls, ID)

    1999-01-01T23:59:59.000Z

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream.

  1. Fast quench reactor method

    DOE Patents [OSTI]

    Detering, B.A.; Donaldson, A.D.; Fincke, J.R.; Kong, P.C.; Berry, R.A.

    1999-08-10T23:59:59.000Z

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream. 8 figs.

  2. Fusion reactor pumped laser

    DOE Patents [OSTI]

    Jassby, Daniel L. (Princeton, NJ)

    1988-01-01T23:59:59.000Z

    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.

  3. Reactor Neutrino Experiments

    E-Print Network [OSTI]

    Jun Cao

    2007-12-06T23:59:59.000Z

    Precisely measuring $\\theta_{13}$ is one of the highest priority in neutrino oscillation study. Reactor experiments can cleanly determine $\\theta_{13}$. Past reactor neutrino experiments are reviewed and status of next precision $\\theta_{13}$ experiments are presented. Daya Bay is designed to measure $\\sin^22\\theta_{13}$ to better than 0.01 and Double Chooz and RENO are designed to measure it to 0.02-0.03. All are heading to full operation in 2010. Recent improvements in neutrino moment measurement are also briefed.

  4. Perspectives on reactor safety

    SciTech Connect (OSTI)

    Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)

    1994-03-01T23:59:59.000Z

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  5. THE MATERIALS OF FAST BREEDER REACTORS

    E-Print Network [OSTI]

    Olander, Donald R.

    2013-01-01T23:59:59.000Z

    times larger in a fast reactor than in a thermal reactor,structural metals in a fast reactor will be subject to farof fuel ele- ments in fast reactors which are roughly one

  6. Innovative design of uranium startup fast reactors

    E-Print Network [OSTI]

    Fei, Tingzhou

    2012-01-01T23:59:59.000Z

    Sodium Fast Reactors are one of the three candidates of GEN-IV fast reactors. Fast reactors play an important role in saving uranium resources and reducing nuclear wastes. Conventional fast reactors rely on transuranic ...

  7. Reactor physics verification of the MCNP6 unstructured mesh capability

    SciTech Connect (OSTI)

    Burke, T. P. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Boulevard, Ann Arbor, MI 48109 (United States); Kiedrowski, B. C.; Martz, R. L. [X-Computational Physics Division, Monte Carlo Codes Group, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Martin, W. R. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Boulevard, Ann Arbor, MI 48109 (United States)

    2013-07-01T23:59:59.000Z

    The Monte Carlo software package MCNP6 has the ability to transport particles on unstructured meshes generated from the Computed-Aided Engineering software Abaqus. Verification is performed using benchmarks with features relevant to reactor physics - Big Ten and the C5G7 computational benchmark. Various meshing strategies are tested and results are compared to reference solutions. Computational performance results are also given. The conclusions show MCNP6 is capable of producing accurate calculations for reactor physics geometries and the computational requirements for small lattice benchmarks are reasonable on modern computing platforms. (authors)

  8. Space-reactor electric systems: subsystem technology assessment

    SciTech Connect (OSTI)

    Anderson, R.V.; Bost, D.; Determan, W.R.

    1983-03-29T23:59:59.000Z

    This report documents the subsystem technology assessment. For the purpose of this report, five subsystems were defined for a space reactor electric system, and the report is organized around these subsystems: reactor; shielding; primary heat transport; power conversion and processing; and heat rejection. The purpose of the assessment was to determine the current technology status and the technology potentials for different types of the five subsystems. The cost and schedule needed to develop these potentials were estimated, and sets of development-compatible subsystems were identified.

  9. Reactor operation environmental information document

    SciTech Connect (OSTI)

    Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

    1989-12-01T23:59:59.000Z

    The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

  10. Power calibrations for TRIGA reactors

    SciTech Connect (OSTI)

    Whittemore, W.L.; Razvi, J.; Shoptaugh, J.R. Jr. [General Atomics, San Diego, CA (United States)

    1988-07-01T23:59:59.000Z

    The purpose of this paper is to establish a framework for the calorimetric power calibration of TRIGA reactors so that reliable results can be obtained with a precision better than {+-} 5%. Careful application of the same procedures has produced power calibration results that have been reproducible to {+-} 1.5%. The procedures are equally applicable to the Mark I, Mark II and Mark III reactors as well as to reactors having much larger reactor tanks and to TRIGA reactors capable of forced cooling up to 3 MW in some cases and 15 MW in another case. In the case of forced cooled TRIGA reactors, the calorimetric power calibration is applicable in the natural convection mode for these reactors using exactly the same procedures as are discussed below for the smaller TRIGA reactors (< 2 MW)

  11. Current Reactor Physics Benchmark Activities at the Idaho National Laboratory

    SciTech Connect (OSTI)

    John D. Bess; Margaret A. Marshall; Mackenzie L. Gorham; Joseph Christensen; James C. Turnbull; Kim Clark

    2011-11-01T23:59:59.000Z

    The International Reactor Physics Experiment Evaluation Project (IRPhEP) [1] and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) [2] were established to preserve integral reactor physics and criticality experiment data for present and future research. These valuable assets provide the basis for recording, developing, and validating our integral nuclear data, and experimental and computational methods. These projects are managed through the Idaho National Laboratory (INL) and the Organisation for Economic Co-operation and Development Nuclear Energy Agency (OECD-NEA). Staff and students at the Department of Energy - Idaho (DOE-ID) and INL are engaged in the development of benchmarks to support ongoing research activities. These benchmarks include reactors or assemblies that support Next Generation Nuclear Plant (NGNP) research, space nuclear Fission Surface Power System (FSPS) design validation, and currently operational facilities in Southeastern Idaho.

  12. Neoclassical simulations of fusion alpha particles in pellet charge exchange experiments on the Tokamak Fusion Test Reactor

    E-Print Network [OSTI]

    Redi, Martha H.

    on the Tokamak Fusion Test Reactor M. H. Redia , S. H. Batha, R. V. Budny, D. S. Darrow, F. M. Levinton, D. C. Mc Test Reactor [Phys. Plas. 5, 1577 (1998)] are found to be in good agreement with measurements includes the neoclassical transport processes, a recent first-principles model for stochastic ripple loss

  13. Neoclassical simulations of fusion alpha particles in pellet charge exchange experiments on the Tokamak Fusion Test Reactor

    E-Print Network [OSTI]

    on the Tokamak Fusion Test Reactor M. H. Redi a , S. H. Batha, M. G. Bell, R. V. Budny, D. S. Darrow, F. M on the Tokamak Fusion Test Reactor (TFTR) [Phys. Plas. 5 , 1577 (1998)] are found to be in good agreement code which includes the neoclassical transport processes, a recent first­principles model

  14. Neoclassical simulations of fusion alpha particles in pellet charge exchange experiments on the Tokamak Fusion Test Reactor

    E-Print Network [OSTI]

    on the Tokamak Fusion Test Reactor M. H. Redia , S. H. Batha, M. G. Bell, R. V. Budny, D. S. Darrow, F. M on the Tokamak Fusion Test Reactor (TFTR) [Phys. Plas. 5, 1577 (1998)] are found to be in good agreement code which includes the neoclassical transport processes, a recent first-principles model

  15. A Generalized Adjoint Framework for Sensitivity and Global Error Estimation in Time-Dependent Nuclear Reactor Simulations 1

    E-Print Network [OSTI]

    Anitescu, Mihai

    -Dependent Nuclear Reactor Simulations 1 H. F. Striplinga, , M. Anitescub , M. L. Adamsa aNuclear Engineering Bateman and transport equations, which govern the time-dependent neutronic behavior of a nuclear reactor framework for computing the adjoint variable to nuclear engineering problems gov- erned by a set

  16. verification & Validation of High-Order Short-Characteristics-Based Deterministic Transport Methodology on Unstructured Grids

    SciTech Connect (OSTI)

    Azmy, Yousry; Wang, Yaqi

    2013-12-20T23:59:59.000Z

    The research team has developed a practical, high-order, discrete-ordinates, short characteristics neutron transport code for three-dimensional configurations represented on unstructured tetrahedral grids that can be used for realistic reactor physics applications at both the assembly and core levels. This project will perform a comprehensive verification and validation of this new computational tool against both a continuous-energy Monte Carlo simulation (e.g. MCNP) and experimentally measured data, an essential prerequisite for its deployment in reactor core modeling. Verification is divided into three phases. The team will first conduct spatial mesh and expansion order refinement studies to monitor convergence of the numerical solution to reference solutions. This is quantified by convergence rates that are based on integral error norms computed from the cell-by-cell difference between the code’s numerical solution and its reference counterpart. The latter is either analytic or very fine- mesh numerical solutions from independent computational tools. For the second phase, the team will create a suite of code-independent benchmark configurations to enable testing the theoretical order of accuracy of any particular discretization of the discrete ordinates approximation of the transport equation. For each tested case (i.e. mesh and spatial approximation order), researchers will execute the code and compare the resulting numerical solution to the exact solution on a per cell basis to determine the distribution of the numerical error. The final activity comprises a comparison to continuous-energy Monte Carlo solutions for zero-power critical configuration measurements at Idaho National Laboratory’s Advanced Test Reactor (ATR). Results of this comparison will allow the investigators to distinguish between modeling errors and the above- listed discretization errors introduced by the deterministic method, and to separate the sources of uncertainty.

  17. Reactor operation safety information document

    SciTech Connect (OSTI)

    Not Available

    1990-01-01T23:59:59.000Z

    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  18. Reed Reactor Facility Annual Report

    SciTech Connect (OSTI)

    Frantz, Stephen G.

    2000-09-01T23:59:59.000Z

    This is the report of the operations, experiments, modifications, and other aspects of the Reed Reactor Facility for the year.

  19. A Transportation Risk Assessment Tool for Analyzing the Transport of Spent Nuclear Fuel and High-Level Radioactive Waste to the Proposed Yucca Mountain Repository

    SciTech Connect (OSTI)

    Ralph Best; T. Winnard; S. Ross; R. Best

    2001-08-17T23:59:59.000Z

    The Yucca Mountain Transportation Database was developed as a data management tool for assembling and integrating data from multiple sources to compile the potential transportation impacts presented in the Draft Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada (DEIS). The database uses the results from existing models and codes such as RADTRAN, RISKIND, INTERLINE, and HIGHWAY to estimate transportation-related impacts of transporting spent nuclear fuel and high-level radioactive waste from commercial reactors and U. S. Department of Energy (DOE) facilities to Yucca Mountain. The source tables in the database are compendiums of information from many diverse sources including: radionuclide quantities for each waste type; route and route characteristics for rail, legal-weight truck, heavy haul. truck, and barge transport options; state-specific accident and fatality rates for routes selected for analysis; packaging and shipment data by waste type; unit risk factors; the complex behavior of the packaged waste forms in severe transport accidents; and the effects of exposure to radiation or the isotopic specific effects of radionclides should they be released in severe transportation accidents. The database works together with the codes RADTRAN (Neuhauser, et al, 1994) and RISKlND (Yuan, et al, 1995) to calculate incident-free dose and accident risk. For the incident-free transportation scenario, the database uses RADTRAN and RISKIND-generated data to calculate doses to offlink populations, onlink populations, people at stops, crews, inspectors, workers at intermodal transfer stations, guards at overnight stops, and escorts, as well as non-radioactive pollution health effects. For accident scenarios, the database uses RADTRAN-generated data to calculate dose risks based on ingestion, inhalation, resuspension, immersion (cloudshine), and groundshine as well as non-radioactive traffic fatalities. The Yucca Mountain EIS Transportation Database was developed using Microsoft Access 97{trademark} software and the Microsoft Windows NT{trademark} operating system. The database consists of tables for storing data, forms for selecting data for querying, and queries for retrieving the data in a predefined format. Database queries retrieve records based on input parameters and are used to calculate incident-free and accident doses using unit risk factors obtained from RADTRAN results. The next section briefly provides some background that led to the development of the database approach used in preparing the Yucca Mountain DEIS. Subsequent sections provide additional details on the database structure and types of impacts calculated using the database.

  20. A review of experiments and results from the transient reactor test (TREAT) facility.

    SciTech Connect (OSTI)

    Deitrich, L. W.

    1998-07-28T23:59:59.000Z

    The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop.

  1. ASME Material Challenges for Advanced Reactor Concepts

    SciTech Connect (OSTI)

    Piyush Sabharwall; Ali Siahpush

    2013-07-01T23:59:59.000Z

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

  2. Reactor Pressure Vessel Head Packaging & Disposal

    SciTech Connect (OSTI)

    Wheeler, D. M.; Posivak, E.; Freitag, A.; Geddes, B.

    2003-02-26T23:59:59.000Z

    Reactor Pressure Vessel (RPV) Head replacements have come to the forefront due to erosion/corrosion and wastage problems resulting from the susceptibility of the RPV Head alloy steel material to water/boric acid corrosion from reactor coolant leakage through the various RPV Head penetrations. A case in point is the recent Davis-Besse RPV Head project, where detailed inspections in early 2002 revealed significant wastage of head material adjacent to one of the Control Rod Drive Mechanism (CRDM) nozzles. In lieu of making ASME weld repairs to the damaged head, Davis-Besse made the decision to replace the RPV Head. The decision was made on the basis that the required weld repair would be too extensive and almost impractical. This paper presents the packaging, transport, and disposal considerations for the damaged Davis-Besse RPV Head. It addresses the requirements necessary to meet Davis Besse needs, as well as the regulatory criteria, for shipping and burial of the head. It focuses on the radiological characterization, shipping/disposal package design, site preparation and packaging, and the transportation and emergency response plans that were developed for the Davis-Besse RPV Head project.

  3. Transportation Politics and Policy

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What'sis Taking Over OurThe Iron Spin Transition in2, 2003ToolsearchTransportation Equipment

  4. Total absorption spectroscopy study of $^{92}$Rb decay: a major contributor to reactor antineutrino flux

    E-Print Network [OSTI]

    A. -A. Zakari-Issoufou; M. Fallot; A. Porta; A. Algora; J. L. Tain; E. Valencia; S. Rice; V. M Bui; S. Cormon; M. Estienne; J. Agramunt; J. Äystö; M. Bowry; J. A. Briz; R. Caballero-Folch; D. Cano-Ott; A. Cucoanes; V. -V. Elomaa; T. Eronen; E. Estévez; G. F. Farrelly; A. R. Garcia; W. Gelletly; M. B Gomez-Hornillos; V. Gorlychev; J. Hakala; A. Jokinen; M. D. Jordan; A. Kankainen; P. Karvonen; V. S. Kolhinen; F. G Kondev; T. Martinez; E. Mendoza; F. Molina; I. Moore; A. Perez; Zs. Podolyák; H. Penttilä; P. H. Regan; M. Reponen; J. Rissanen; B. Rubio; T. Shiba; A. A. Sonzogni; C. Weber; IGISOL collaboration

    2015-05-20T23:59:59.000Z

    The antineutrino spectra measured in recent experiments at reactors are inconsistent with calculations based on the conversion of integral beta spectra recorded at the ILL reactor. $^{92}$Rb makes the dominant contribution to the reactor spectrum in the 5-8 MeV range but its decay properties are in question. We have studied $^{92}$Rb decay with total absorption spectroscopy. Previously unobserved beta feeding was seen in the 4.5-5.5 region and the GS to GS feeding was found to be 87.5(25)%. The impact on the reactor antineutrino spectra calculated with the summation method is shown and discussed.

  5. Total absorption spectroscopy study of $^{92}$Rb decay: a major contributor to reactor antineutrino flux

    E-Print Network [OSTI]

    Zakari-Issoufou, A -A; Porta, A; Algora, A; Tain, J L; Valencia, E; Rice, S; Bui, V M; Cormon, S; Estienne, M; Agramunt, J; Äystö, J; Bowry, M; Briz, J A; Caballero-Folch, R; Cano-Ott, D; Cucoanes, A; Elomaa, V -V; Eronen, T; Estévez, E; Farrelly, G F; Garcia, A R; Gelletly, W; Gomez-Hornillos, M B; Gorlychev, V; Hakala, J; Jokinen, A; Jordan, M D; Kankainen, A; Karvonen, P; Kolhinen, V S; Kondev, F G; Martinez, T; Mendoza, E; Molina, F; Moore, I; Perez, A; Podolyák, Zs; Penttilä, H; Regan, P H; Reponen, M; Rissanen, J; Rubio, B; Shiba, T; Sonzogni, A A; Weber, C

    2015-01-01T23:59:59.000Z

    The antineutrino spectra measured in recent experiments at reactors are inconsistent with calculations based on the conversion of integral beta spectra recorded at the ILL reactor. $^{92}$Rb makes the dominant contribution to the reactor spectrum in the 5-8 MeV range but its decay properties are in question. We have studied $^{92}$Rb decay with total absorption spectroscopy. Previously unobserved beta feeding was seen in the 4.5-5.5 region and the GS to GS feeding was found to be 87.5(25)%. The impact on the reactor antineutrino spectra calculated with the summation method is shown and discussed.

  6. Thermal Reactor Safety

    SciTech Connect (OSTI)

    Not Available

    1980-06-01T23:59:59.000Z

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

  7. Nuclear reactor building

    DOE Patents [OSTI]

    Gou, Perng-Fei (Saratoga, CA); Townsend, Harold E. (Campbell, CA); Barbanti, Giancarlo (Sirtori, IT)

    1994-01-01T23:59:59.000Z

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed thereabove. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define therebetween an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin.

  8. Nuclear reactor building

    DOE Patents [OSTI]

    Gou, P.F.; Townsend, H.E.; Barbanti, G.

    1994-04-05T23:59:59.000Z

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed there above. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define there between an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin. 4 figures.

  9. Fossil fuel furnace reactor

    DOE Patents [OSTI]

    Parkinson, William J. (Los Alamos, NM)

    1987-01-01T23:59:59.000Z

    A fossil fuel furnace reactor is provided for simulating a continuous processing plant with a batch reactor. An internal reaction vessel contains a batch of shale oil, with the vessel having a relatively thin wall thickness for a heat transfer rate effective to simulate a process temperature history in the selected continuous processing plant. A heater jacket is disposed about the reactor vessel and defines a number of independent controllable temperature zones axially spaced along the reaction vessel. Each temperature zone can be energized to simulate a time-temperature history of process material through the continuous plant. A pressure vessel contains both the heater jacket and the reaction vessel at an operating pressure functionally selected to simulate the continuous processing plant. The process yield from the oil shale may be used as feedback information to software simulating operation of the continuous plant to provide operating parameters, i.e., temperature profiles, ambient atmosphere, operating pressure, material feed rates, etc., for simulation in the batch reactor.

  10. Reactor Monitoring with Neutrinos

    E-Print Network [OSTI]

    M. Cribier

    2007-04-06T23:59:59.000Z

    The fundamental knowledge on neutrinos acquired in the recent years open the possibility of applied neutrino physics. Among it the automatic and non intrusive monitoring of nuclear reactor by its antineutrino signal could be very valuable to IAEA in charge of the control of nuclear power plants. Several efforts worldwide have already started.

  11. Transportation System Concept of Operations

    SciTech Connect (OSTI)

    N. Slater-Thompson

    2006-08-16T23:59:59.000Z

    The Nuclear Waste Policy Act of 1982 (NWPA), as amended, authorized the DOE to develop and manage a Federal system for the disposal of SNF and HLW. OCRWM was created to manage acceptance and disposal of SNF and HLW in a manner that protects public health, safety, and the environment; enhances national and energy security; and merits public confidence. This responsibility includes managing the transportation of SNF and HLW from origin sites to the Repository for disposal. The Transportation System Concept of Operations is the core high-level OCRWM document written to describe the Transportation System integrated design and present the vision, mission, and goals for Transportation System operations. By defining the functions, processes, and critical interfaces of this system early in the system development phase, programmatic risks are minimized, system costs are contained, and system operations are better managed, safer, and more secure. This document also facilitates discussions and understanding among parties responsible for the design, development, and operation of the Transportation System. Such understanding is important for the timely development of system requirements and identification of system interfaces. Information provided in the Transportation System Concept of Operations includes: the functions and key components of the Transportation System; system component interactions; flows of information within the system; the general operating sequences; and the internal and external factors affecting transportation operations. The Transportation System Concept of Operations reflects OCRWM's overall waste management system policies and mission objectives, and as such provides a description of the preferred state of system operation. The description of general Transportation System operating functions in the Transportation System Concept of Operations is the first step in the OCRWM systems engineering process, establishing the starting point for the lower level descriptions. of subsystems and components, and the Transportation System Requirements Document. Other program and system documents, plans, instructions, and detailed designs will be consistent with and informed by the Transportation System Concept of Operations. The Transportation System Concept of Operations is a living document, enduring throughout the OCRWM systems engineering lifecycle. It will undergo formal approval and controlled revisions as appropriate while the Transportation System matures. Revisions will take into account new policy decisions, new information available through system modeling, engineering investigations, technical analyses and tests, and the introduction of new technologies that can demonstrably improve system performance.

  12. Surrogate Spent Nuclear Fuel Vibration Integrity Investigation

    SciTech Connect (OSTI)

    Wang, Jy-An John [ORNL; Wang, Hong [ORNL; Bevard, Bruce Balkcom [ORNL; Howard, Rob L [ORNL

    2014-01-01T23:59:59.000Z

    Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During transportation, SNF experiences unique conditions and challenges to cladding integrity due to the vibrational and impact loading encountered during road or rail shipment. ORNL has been developing testing capabilities that can be used to improve our understanding of the impacts of vibration loading on SNF integrity, especially for high burn-up SNF in normal transportation operation conditions. This information can be used to meet nuclear industry and U.S. Nuclear Regulatory Commission needs in the area of safety of SNF storage and transportation operations.

  13. Reactor vessel support system. [LMFBR

    DOE Patents [OSTI]

    Golden, M.P.; Holley, J.C.

    1980-05-09T23:59:59.000Z

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  14. Information Technology and Intelligent Transportation: A

    E-Print Network [OSTI]

    Yehudai, Amiram

    Motorization Growth: Road Vehicle Populations by Region, 1996 and 2020 Source: EIA, International Energy Integration (VII) Initiative · Graduate Program in Computational Transportation Science · Conductive IT trends Integration (VII): A Federal Initiative · Vision ­ Information about all roads, all the time · To control

  15. A Fast Multigrid Algorithm for Isotropic Transport Problems I: Pure Scattering \\Lambda

    E-Print Network [OSTI]

    Oliveira, Suely

    geometry. These problems are important in many application such as nuclear reactor design, radiation techniques traditionally used in numerical transport theory (e.g., see Morel and Manteuffel [15]). Since

  16. Parallel algorithms for Monte Carlo particle transport simulation on exascale computing architectures

    E-Print Network [OSTI]

    Romano, Paul K. (Paul Kollath)

    2013-01-01T23:59:59.000Z

    Monte Carlo particle transport methods are being considered as a viable option for high-fidelity simulation of nuclear reactors. While Monte Carlo methods offer several potential advantages over deterministic methods, there ...

  17. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    SciTech Connect (OSTI)

    Bragg-Sitton, Shannon M. [NASA Marshall Space Flight Center, Nuclear and Advanced Propulsion Branch, ER-11, MSFC, AL 35812 (United States); Morton, T. J. [Department of Chemical and Nuclear Engineering, University of New Mexico, Albuquerque, NM 87131 (United States)

    2006-01-20T23:59:59.000Z

    Non-nuclear testing can be a valuable tool in the development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe (HP) cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system. Reactivity feedback calculations were then based on a bulk reactivity feedback coefficient and measured average core temperature. This paper presents preliminary results from similar dynamic testing of a direct drive gas cooled reactor system (DDG), demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. Although the HP and DDG designs both utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility. Planned system upgrades to allow implementation of higher fidelity dynamic testing are also discussed. Proposed DDG testing will utilize a higher fidelity point kinetics model to control core power transients, and reactivity feedback will be based on localized feedback coefficients and several independent temperature measurements taken within the core block. This paper presents preliminary test results and discusses the methodology that will be implemented in follow-on DDG testing and the additional instrumentation required to implement high fidelity dynamic testing.

  18. Small Modular Nuclear Reactors | Department of Energy

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What'sis Taking Over Our Instagram Secretary Moniz9MorganYou are here Home » SmallEnergyReactor

  19. Reactor Neutrino Oscillation Experiments | ornl.gov

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC)IntegratedSpeedingTechnicalPurchase,Reaction-Driven Restructuring of BimetallicReactor

  20. Transport Corrections in Nodal Diffusion Codes for HTR Modeling

    SciTech Connect (OSTI)

    Abderrafi M. Ougouag; Frederick N. Gleicher

    2010-08-01T23:59:59.000Z

    The cores and reflectors of High Temperature Reactors (HTRs) of the Next Generation Nuclear Plant (NGNP) type are dominantly diffusive media from the point of view of behavior of the neutrons and their migration between the various structures of the reactor. This means that neutron diffusion theory is sufficient for modeling most features of such reactors and transport theory may not be needed for most applications. Of course, the above statement assumes the availability of homogenized diffusion theory data. The statement is true for most situations but not all. Two features of NGNP-type HTRs require that the diffusion theory-based solution be corrected for local transport effects. These two cases are the treatment of burnable poisons (BP) in the case of the prismatic block reactors and, for both pebble bed reactor (PBR) and prismatic block reactor (PMR) designs, that of control rods (CR) embedded in non-multiplying regions near the interface between fueled zones and said non-multiplying zones. The need for transport correction arises because diffusion theory-based solutions appear not to provide sufficient fidelity in these situations.

  1. Nuclear reactor construction with bottom supported reactor vessel

    DOE Patents [OSTI]

    Sharbaugh, John E. (Bullskin Township, Fayette County, PA)

    1987-01-01T23:59:59.000Z

    An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core within the pool. The reactor vessel has an open top end, a closed flat bottom end wall and a continuous cylindrical closed side wall interconnecting the top end and bottom end wall. The reactor also has a generally cylindrical concrete containment structure surrounding the reactor vessel and being formed by a cylindrical side wall spaced outwardly from the reactor vessel side wall and a flat base mat spaced below the reactor vessel bottom end wall. A central support pedestal is anchored to the containment structure base mat and extends upwardly therefrom to the reactor vessel and upwardly therefrom to the reactor core so as to support the bottom end wall of the reactor vessel and the lower end of the reactor core in spaced apart relationship above the containment structure base mat. Also, an annular reinforced support structure is disposed in the reactor vessel on the bottom end wall thereof and extends about the lower end of the core so as to support the periphery thereof. In addition, an annular support ring having a plurality of inward radially extending linear members is disposed between the containment structure base mat and the bottom end of the reactor vessel wall and is connected to and supports the reactor vessel at its bottom end on the containment structure base mat so as to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event. The reactor construction also includes a bed of insulating material in sand-like granular form, preferably being high density magnesium oxide particles, disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall on the containment structure base mat so as to insulate the reactor vessel bottom end wall from the containment structure base mat and allow the reactor vessel bottom end wall to freely expand as it heats up while providing continuous support thereof. Further, a deck is supported upon the side wall of the containment structure above the top open end of the reactor vessel, and a plurality of serially connected extendible and retractable annular bellows extend between the deck and the top open end of the reactor vessel and flexibly and sealably interconnect the reactor vessel at its top end to the deck. An annular guide ring is disposed on the containment structure and extends between its side wall and the top open end of the reactor vessel for providing lateral support of the reactor vessel top open end by limiting imposition of lateral loads on the annular bellows by the occurrence of a lateral seismic event.

  2. Transportation and its Infrastructure

    E-Print Network [OSTI]

    2007-01-01T23:59:59.000Z

    40 pp. IEA, 2004c: Biofuels for Transport: An Internationalthe ACT Map scenario, transport biofuels production reachesestimates that biofuels’ share of transport fuel could

  3. Advanced simulation of electron heat transport in fusion plasmas , S Klasky2

    E-Print Network [OSTI]

    Lin, Zhihong

    first heat electrons. First-principles simulations of electron turbulence are much more challenging due reactor design relies on the extrapolation of the transport level from present-day fusion experiments to much larger future devices, such as the International Thermonuclear Experimental Reactor (ITER)1

  4. Iterative methods for neutron transport eigenvalue Fynn Scheben and Ivan G. Graham

    E-Print Network [OSTI]

    Graham, Ivan

    iterative methods for computing criticality in nuclear reactors. In general this requires the solution words. neutron transport, criticality, generalised eigenvalue problem, symmetry, inexact inverse iteration. AMS subject classifications. 45C05, 65F15, 65N25, 65R20, 65Z05, 82D75 1. Reactor criticality

  5. RMC - A Monte Carlo code for reactor physics analysis

    SciTech Connect (OSTI)

    Wang, K.; Li, Z.; She, D.; Liang, J.; Xu, Q.; Qiu, A.; Yu, J.; Sun, J.; Fan, X.; Yu, G. [Department of Engineering Physics, Tsinghua University, Liuqing Building, Beijing, 100084 (China)

    2013-07-01T23:59:59.000Z

    A new Monte Carlo neutron transport code RMC has been being developed by Department of Engineering Physics, Tsinghua University, Beijing as a tool for reactor physics analysis on high-performance computing platforms. To meet the requirements of reactor analysis, RMC now has such functions as criticality calculation, fixed-source calculation, burnup calculation and kinetics simulations. Some techniques for geometry treatment, new burnup algorithm, source convergence acceleration, massive tally and parallel calculation, and temperature dependent cross sections processing are researched and implemented in RMC to improve the efficiency. Validation results of criticality calculation, burnup calculation, source convergence acceleration, tallies performance and parallel performance shown in this paper prove the capabilities of RMC in dealing with reactor analysis problems with good performances. (authors)

  6. A Multivariate Time Series Method for Monte Carlo Reactor Analysis

    SciTech Connect (OSTI)

    Taro Ueki

    2008-08-14T23:59:59.000Z

    A robust multivariate time series method has been established for the Monte Carlo calculation of neutron multiplication problems. The method is termed Coarse Mesh Projection Method (CMPM) and can be implemented using the coarse statistical bins for acquisition of nuclear fission source data. A novel aspect of CMPM is the combination of the general technical principle of projection pursuit in the signal processing discipline and the neutron multiplication eigenvalue problem in the nuclear engineering discipline. CMPM enables reactor physicists to accurately evaluate major eigenvalue separations of nuclear reactors with continuous energy Monte Carlo calculation. CMPM was incorporated in the MCNP Monte Carlo particle transport code of Los Alamos National Laboratory. The great advantage of CMPM over the traditional Fission Matrix method is demonstrated for the three space-dimensional modeling of the initial core of a pressurized water reactor.

  7. Spent nuclear fuel discharges from US reactors 1993

    SciTech Connect (OSTI)

    Not Available

    1995-02-01T23:59:59.000Z

    The Energy Information Administration (EIA) of the U.S. Department of Energy (DOE) administers the Nuclear Fuel Data Survey, Form RW-859. This form is used to collect data on fuel assemblies irradiated at commercial nuclear reactors operating in the United States, and the current inventories and storage capacities of those reactors. These data are important to the design and operation of the equipment and facilities that DOE will use for the future acceptance, transportation, and disposal of spent fuels. The data collected and presented identifies trends in burnup, enrichment, and spent nuclear fuel discharged form commercial light-water reactor as of December 31, 1993. The document covers not only spent nuclear fuel discharges; but also site capacities and inventories; canisters and nonfuel components; and assembly type characteristics.

  8. The Role of Uncertainty Quantification for Reactor Physics

    SciTech Connect (OSTI)

    Massimo Salvatores; Giuseppe Palmiotti; G. Aliberti

    2015-01-01T23:59:59.000Z

    The role of uncertainty quantification has been stressed and has been the object of several assessments in the past (see e.g. References 1 and 2 among many others), in particular in relation to design requirements for safety assessments, design margins definition and optimization, both for the reactor core and for the associated fuel cycles. The use of integral experiments has been advocated since many years, and recently re-assessed in order to reduce uncertainties and to define new reduced ''a-posteriori'' uncertainties. While uncertainty quantification in the case of existing power plants benefits from a large data base of operating reactor experimental results, innovative reactor systems (reactor and associated fuel cycles) should rely on limited power reactor experiment data bases and on a number of past integral experiments that should be shown to be representative enough. Moreover, in some cases, in particular related to innovative fuel cycle performance and feasibility assessment, nuclear data uncertainties are the only available information. Uncertainty quantification in that case becomes a tool for detecting potential show stoppers associated to specfic fuel cycle strategies, besides the challenges related to fuel properties, fuel processing chemistry and material performance issues.

  9. Transportation implications of a closed fuel cycle.

    SciTech Connect (OSTI)

    Bullard, Tim (University of Nevada - Reno); Bays, Samuel (Idaho National Laboratory); Dennis, Matthew L.; Weiner, Ruth F.; Sorenson, Ken Bryce; Dixon, Brent (Idaho National Laboratory); Greiner, Miles (University of Nevada - Reno)

    2010-10-01T23:59:59.000Z

    Transportation for each step of a closed fuel cycle is analyzed in consideration of the availability of appropriate transportation infrastructure. The United States has both experience and certified casks for transportation that may be required by this cycle, except for the transport of fresh and used MOX fuel and fresh and used Advanced Burner Reactor (ABR) fuel. Packaging that had been used for other fuel with somewhat similar characteristics may be appropriate for these fuels, but would be inefficient. Therefore, the required neutron and gamma shielding, heat dissipation, and criticality were calculated for MOX and ABR fresh and spent fuel. Criticality would not be an issue, but the packaging design would need to balance neutron shielding and regulatory heat dissipation requirements.

  10. Retrospective dosimetry analyses of reactor vessel cladding samples

    SciTech Connect (OSTI)

    Greenwood, L. R.; Soderquist, C. Z. [Battelle Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Fero, A. H. [Westinghouse Electric Company, Cranberry Twp., PA 16066 (United States)

    2011-07-01T23:59:59.000Z

    Reactor pressure vessel cladding samples for Ringhals Units 3 and 4 in Sweden were analyzed using retrospective reactor dosimetry techniques. The objective was to provide the best estimates of the neutron fluence for comparison with neutron transport calculations. A total of 51 stainless steel samples consisting of chips weighing approximately 100 to 200 mg were removed from selected locations around the pressure vessel and were sent to Pacific Northwest National Laboratory for analysis. The samples were fully characterized and analyzed for radioactive isotopes, with special interest in the presence of Nb-93m. The RPV cladding retrospective dosimetry results will be combined with a re-evaluation of the surveillance capsule dosimetry and with ex-vessel neutron dosimetry results to form a comprehensive 3D comparison of measurements to calculations performed with 3D deterministic transport code. (authors)

  11. Investigation of vessel exterior air cooling for a HLMC reactor

    SciTech Connect (OSTI)

    Sienicki, J. J.; Spencer, B. W.

    2000-01-13T23:59:59.000Z

    The Secure Transportable Autonomous Reactor (STAR) concept under development at Argonne National Laboratory provides a small (300 MWt) reactor module for steam supply that incorporates design features to attain proliferation resistance, heightened passive safety, and improved cost competitiveness through extreme simplification. Examples are the achievement of 100%+ natural circulation heat removal from the low power density/low pressure drop ultra-long lifetime core and utilization of lead-bismuth eutectic (LBE) coolant enabling elimination of main coolant pumps as well as the need for an intermediate heat transport circuit. It is required to provide a passive means of removing decay heat and effecting reactor cooldown in the event that the normal steam generator heat sink, including its normal shutdown heat removal mode, is postulated to be unavailable. In the present approach, denoted as the Reactor Exterior Cooling System (RECS), passive decay heat removal is provided by cooling the outside of the containment/guard vessel with air. RECS is similar to the Reactor Vessel Auxiliary Cooling System (RVACS) incorporated into the PRISM design. However, to enhance the heat removal, RECS incorporates fins on the containment vessel exterior to enhance heat transfer to air as well as removable steel venetian conductors that provide a conduction heat transfer path across the reactor vessel-containment vessel gap to enhance heat transfer between the vessels. The objective of the present work is to investigate the effectiveness of air cooling in removing heat from the vessel and limiting the coolant temperature increase following a sudden complete loss of the steam generator heat sink.

  12. Investigation of vessel exterior air cooling for an HLMC reactor

    SciTech Connect (OSTI)

    Sienicki, J.J.; Spencer, B.W.

    2000-07-01T23:59:59.000Z

    The secure transportable autonomous reactor (STAR) concept under development at Argonne National Laboratory provides a small [300-MW(thermal)] reactor module for steam supply that incorporates design features to attain proliferation resistance, heightened passive safety, and improved cost competitiveness through extreme simplification. Examples are the achievement of 100% + natural-circulation heat removal from the low-power-density/low-pressure-drop ultralong lifetime core and utilization of lead-bismuth eutectic (LBE) coolant enabling elimination of main coolant pumps as well as the need for an intermediate heat transport circuit. It is required to provide a passive means of removing decay heat and effecting reactor cooldown in the event that the normal steam generator heat sink, including its normal shutdown heat removal mode, is postulated to be unavailable. In the present approach, denoted as the reactor exterior cooling system (RECS), passive decay heat removal is provided by cooling the outside of the containment/guard vessel with air. RECS is similar to the reactor vessel auxiliary cooling system (RVACS) incorporated into the PRISM design. However, to enhance the heat removal, RECS incorporates fins on the containment vessel exterior to enhance heat transfer to air as well as removable steel venetian conductors that provide a conduction heat transfer path across the reactor vessel-containment vessel gap to enhance heat transfer between the vessels. The objective of the present work is to investigate the effectiveness of air cooling in removing heat from the vessel and limiting the coolant temperature increase following a sudden complete loss of the steam generator heat sink.

  13. Nuclear reactor safety device

    DOE Patents [OSTI]

    Hutter, E.

    1983-08-15T23:59:59.000Z

    A safety device is described for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of a thermal excursion. It comprises a laminated strip helically configured to form a tube, said tube being in operative relation to said control rod. The laminated strip is formed of at least two materials having different thermal coefficients of expansion, and is helically configured such that the material forming the outer lamina of the tube has a greater thermal coefficient of expansion than the material forming the inner lamina of said tube. In the event of a thermal excursion the laminated strip will tend to curl inwardly so that said tube will increase in length, whereby as said tube increases in length it exerts a force on said control rod to axially reposition said control rod with respect to said core.

  14. Fusion reactor pumped laser

    DOE Patents [OSTI]

    Jassby, D.L.

    1987-09-04T23:59:59.000Z

    A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam. 10 figs.

  15. Spherical torus fusion reactor

    DOE Patents [OSTI]

    Peng, Yueng-Kay M. (Oak Ridge, TN)

    1989-01-01T23:59:59.000Z

    A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

  16. The development of an integrated multistage fluid bed retorting process. Technical report, April 1, 1992--June 30, 1992

    SciTech Connect (OSTI)

    Carter, S.D.; Taulbee, D.N.; Robl, T.L.; Hower, J.C.

    1992-08-01T23:59:59.000Z

    This report summarizes the progress made on the development of an integrated multistage fluidized bed retorting process (KENTORT II) during the period of April 1, 1992 through June 30, 1992. The KENTORT II process includes integral fluidized bed zones for pyrolysis, gasification, and combustion of the oil shale. The purpose of this program is to design and test the KENTORT II process at the 50-lb/hr scale. The raw oil shale sample for the program was mined, prepared, characterized and stored this quarter. The shale that was chosen was from the high-grade zone of the Devonian Cleveland Member of the Ohio Shale in Montgomery County, Kentucky. The shale was mined and then transported to the contractor`s crushing facility where it was crushed, double-screened, and loaded into 85 55-gal barrels. The barrels, containing a total of 25-30 tons of shale, were transported to the (CAER) Center for Applied Energy Research where the shale was double-screened, analyzed and stored. A major objective of the program is the study of solid-induced secondary coking and cracking reactions. A valved fluidized bed reactor has been the primary apparatus used for this study prior to this quarter, but two additional techniques have been initiated this quarter for the study of other aspects of this issue. First, the two-stage hydropyrolysis reactor at the University of Strathclyde, Glasgow, Scotland, was used to study the coking tendency of shale oil vapors under a wide range of pyrolysis and hydropyrolysis conditions. This work enabled us to examine secondary reactions under high pressure conditions (up to 150 bar) which were previously unavailable. Second, the development of a fixed bed reactor system was initiated at the CAER to study the coking and cracking characteristics of model compounds. A fixed bed apparatus was necessary because the conversion of model compounds was too low in the fluidized bed apparatus.

  17. Reactor Safety Research Programs Quarterly Report January - March 1980

    SciTech Connect (OSTI)

    Hagen, C. M

    1980-10-01T23:59:59.000Z

    This document summarizes the work performed by Pacific Northwest Laboratory from January 1 through March 31, 1980, 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 serviceinduced 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, Ispra, 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.

  18. Reactor Safety Research Programs Quarterly Report July- September 1980

    SciTech Connect (OSTI)

    Edler, S. K.

    1980-12-01T23:59:59.000Z

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from April 1 through June 30, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission {NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility 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 (PWR) 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 loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; blowdown and reflood tests in the test facility at Cadarache, France; 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.

  19. Reactor Safety Research Programs Quarterly Report October - December 1981

    SciTech Connect (OSTI)

    Edler, S. K.

    1982-03-01T23:59:59.000Z

    This document summarizes the work performed by Pacific Northwest laboratory (PNL) from October 1 through December 31, 1981, for the Division of Accident Evaluation, 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 serviceinduced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-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-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and post accident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  20. Reactor Safety Research Programs Quarterly Report April -June 1980

    SciTech Connect (OSTI)

    Edler, S. K.

    1980-11-01T23:59:59.000Z

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from April 1 through June 30, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission {NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility 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 (PWR) 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 loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; blowdown and reflood tests in the test facility at Cadarache, France; 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.

  1. Reactor Safety Research Programs Quarterly Report July - September 1981

    SciTech Connect (OSTI)

    Edler, S. K.

    1982-01-01T23:59:59.000Z

    This document summarizes the work performed by Pacific Northwest laboratory (PNL) from July 1 through September 30, 1981, for the Division of Accident Evaluation, 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 pipe-to-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-of-coolant 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, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  2. Reactor Safety Research Programs Quarterly Report October - December 1980

    SciTech Connect (OSTI)

    Edler, S K

    1981-04-01T23:59:59.000Z

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from October 1 through December 31, 1980, 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 structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NOE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) 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 loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor 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.

  3. Reactor technology assessment and selection utilizing systems engineering approach

    SciTech Connect (OSTI)

    Zolkaffly, Muhammed Zulfakar; Han, Ki-In [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2014-02-12T23:59:59.000Z

    The first Nuclear power plant (NPP) deployment in a country is a complex process that needs to consider technical, economic and financial aspects along with other aspects like public acceptance. Increased interest in the deployment of new NPPs, both among newcomer countries and those with expanding programs, necessitates the selection of reactor technology among commercially available technologies. This paper reviews the Systems Decision Process (SDP) of Systems Engineering and applies it in selecting the most appropriate reactor technology for the deployment in Malaysia. The integrated qualitative and quantitative analyses employed in the SDP are explored to perform reactor technology assessment and to select the most feasible technology whose design has also to comply with the IAEA standard requirements and other relevant requirements that have been established in this study. A quick Malaysian case study result suggests that the country reside with PWR (pressurized water reactor) technologies with more detailed study to be performed in the future for the selection of the most appropriate reactor technology for Malaysia. The demonstrated technology assessment also proposes an alternative method to systematically and quantitatively select the most appropriate reactor technology.

  4. Reactor Subsystem Simulation for Nuclear Hybrid Energy Systems

    SciTech Connect (OSTI)

    Shannon Bragg-Sitton; J. Michael Doster; Alan Rominger

    2012-09-01T23:59:59.000Z

    Preliminary system models have been developed by Idaho National Laboratory researchers and are currently being enhanced to assess integrated system performance given multiple sources (e.g., nuclear + wind) and multiple applications (i.e., electricity + process heat). Initial efforts to integrate a Fortran-based simulation of a small modular reactor (SMR) with the balance of plant model have been completed in FY12. This initial effort takes advantage of an existing SMR model developed at North Carolina State University to provide initial integrated system simulation for a relatively low cost. The SMR subsystem simulation details are discussed in this report.

  5. Synfuels from fusion: using the tandem mirror reactor and a thermochemical cycle to produce hydrogen

    SciTech Connect (OSTI)

    Werner, R.W. (ed.)

    1982-11-01T23:59:59.000Z

    This study is concerned with the following area: (1) the tandem mirror reactor and its physics; (2) energy balance; (3) the lithium oxide canister blanket system; (4) high-temperature blanket; (5) energy transport system-reactor to process; (6) thermochemical hydrogen processes; (7) interfacing the GA cycle; (8) matching power and temperature demands; (9) preliminary cost estimates; (10) synfuels beyond hydrogen; and (11) thermodynamics of the H/sub 2/SO/sub 4/-H/sub 2/O system. (MOW)

  6. Small space reactor power systems for unmanned solar system exploration missions

    SciTech Connect (OSTI)

    Bloomfield, H.S.

    1987-12-01T23:59:59.000Z

    A preliminary feasibility study of the application of small nuclear reactor space power systems to the Mariner Mark II Cassini spacecraft/mission was conducted. The purpose of the study was to identify and assess the technology and performance issues associated with the reactor power system/spacecraft/mission integration. The Cassini mission was selected because study of the Saturn system was identified as a high priority outer planet exploration objective. Reactor power systems applied to this mission were evaluated for two different uses. First, a very small 1 kWe reactor power system was used as an RTG replacement for the nominal spacecraft mission science payload power requirements while still retaining the spacecraft's usual bipropellant chemical propulsion system. The second use of reactor power involved the additional replacement of the chemical propulsion system with a small reactor power system and an electric propulsion system. The study also provides an examination of potential applications for the additional power available for scientific data collection. The reactor power system characteristics utilized in the study were based on a parametric mass model that was developed specifically for these low power applications. The model was generated following a neutronic safety and operational feasibility assessment of six small reactor concepts solicited from U.S. industry. This assessment provided the validation of reactor safety for all mission phases and generatad the reactor mass and dimensional data needed for the system mass model.

  7. Nuclear reactor sealing system

    DOE Patents [OSTI]

    McEdwards, James A. (Calabasas, CA)

    1983-01-01T23:59:59.000Z

    A liquid metal-cooled nuclear reactor sealing system. The nuclear reactor includes a vessel sealed at its upper end by a closure head. The closure head comprises at least two components, one of which is rotatable; and the two components define an annulus therebetween. The sealing system includes at least a first and second inflatable seal disposed in series in an upper portion of the annulus. The system further includes a dip seal extending into a body of insulation located adjacent a bottom portion of the closure head. The dip seal comprises a trough formed by a lower portion of one of the components, and a seal blade pendently supported from the other component and extending downwardly into the trough. A body of liquid metal is contained in the trough which submerges a portion of the seal blade. The seal blade is provided with at least one aperture located above the body of liquid metal for providing fluid communication between the annulus intermediate the dip seal and the inflatable seals, and a body of cover gas located inside the vessel. There also is provided means for introducing a purge gas into the annulus intermediate the inflatable seals and the seal blade. The purge gas is introduced in an amount sufficient to substantially reduce diffusion of radioactive cover gas or sodium vapor up to the inflatable seals. The purge gas mixes with the cover gas in the reactor vessel where it can be withdrawn from the vessel for treatment and recycle to the vessel.

  8. NREL: Transportation Research - Transportation and Hydrogen Newsletter...

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

    The Future of Sustainable Transportation This is the January 2015 issue of the Transportation and Hydrogen Newsletter. Illustration of an electric vehicle Illustration of an...

  9. Radiation Transport Calculations and Simulations

    SciTech Connect (OSTI)

    Fasso, Alberto; /SLAC; Ferrari, A.; /CERN

    2011-06-30T23:59:59.000Z

    This article is an introduction to the Monte Carlo method as used in particle transport. After a description at an elementary level of the mathematical basis of the method, the Boltzmann equation and its physical meaning are presented, followed by Monte Carlo integration and random sampling, and by a general description of the main aspects and components of a typical Monte Carlo particle transport code. In particular, the most common biasing techniques are described, as well as the concepts of estimator and detector. After a discussion of the different types of errors, the issue of Quality Assurance is briefly considered.

  10. Westinghouse Small Modular Reactor nuclear steam supply system design

    SciTech Connect (OSTI)

    Memmott, M. J.; Harkness, A. W.; Van Wyk, J. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

    2012-07-01T23:59:59.000Z

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam generator, and eight reactor coolant pumps (RCP). The containment vessel is 27.1 m (89 ft) long and 9.8 m (32 ft) in diameter, and is designed to withstand pressures up to 1.7 MPa (250 psi). It is completely submerged in a pool of water serving as a heat sink and radiation shield. Housed within the containment are four combined core makeup tanks (CMT)/passive residual heat removal (PRHR) heat exchangers, two in-containment pools (ICP), two ICP tanks and four valves which function as the automatic depressurization system (ADS). The PRHR heat exchangers are thermally connected to two different ultimate heat sink (UHS) tanks which provide transient cooling capabilities. (authors)

  11. Coulomb interactions between dust particles in plasma etching reactors

    SciTech Connect (OSTI)

    Hwang, H.H.; Kushner, M.J. [Univ. of Illinois, Urbana, IL (United States). Dept. of Electrical and Computer Engineering

    1996-12-31T23:59:59.000Z

    Wafer contamination by particles, or dust, in plasma processing reactors remains a continuing concern in the microelectronics industry. Particles charge negatively in low temperature plasmas and resemble electrically floating bodies. The transport of these particles in plasma processing reactors is dominated by electrostatic, ion-drag, fluid-drag, and thermophoretic forces. Under conditions where the particle density is large, Debye shielding may be insufficient to isolate the particles, leading to particle-particle Coulomb interactions. Such interactions are likely to occur in trapping locations, which are typically near the plasma-sheath boundaries in Reactive Ion Etching (RIE) discharges. Particles that experience Coulomb interactions may display collective behavior, an extreme example being a Coulomb liquid or solid. Particle transport in plasma processing reactors has been studied extensively to predict rates of wafer contamination thought to date particle-particle interactions have not been addressed. In this paper, the authors discuss results from a computer model for dust particle transport in RIE discharges where particle-particle Coulomb interactions are included.

  12. Analytical modelling of hydrogen transport in reactor containments

    E-Print Network [OSTI]

    Manno, V.

    1983-01-01T23:59:59.000Z

    There are two diffusion processes, molecular and turbulent, which should be modelled in different ways. Molecular diffusion is modelled by Wilke's formula for the multi-component gas diffusion, where the diffusion constants ...

  13. Modular Pebble Bed Reactor High Temperature Gas Reactor

    E-Print Network [OSTI]

    Built · Site Assembled · On--line Refueling · Modules added to meet demand. · No Reprocessing · High Experiments · Non-Proliferation · Safeguards · Waste Disposal · Reactor Research/ Demonstration Facility

  14. Arnold Schwarzenegger INTEGRATING NEW AND EMERGING

    E-Print Network [OSTI]

    GRID INFRASTRUCTURE A Report on a Smart Grid for California PIERFINALPROJECTREPORT Prepared For Technologies into the California Smart Grid Infrastructure. EPRI, Palo Alto, CA; California Energy Commission Technologies · Transportation Integrating New and Emerging Technologies into the California Smart Grid

  15. Thermionic Reactor Design Studies

    SciTech Connect (OSTI)

    Schock, Alfred

    1994-06-01T23:59:59.000Z

    During the 1960's and early 70's the author performed extensive design studies, analyses, and tests aimed at thermionic reactor concepts that differed significantly from those pursued by other investigators. Those studies, like most others under Atomic Energy Commission (AEC and DOE) and the National Aeronautics and Space Administration (NASA) sponsorship, were terminated in the early 1970's. Some of this work was previously published, but much of it was never made available in the open literature. U.S. interest in thermionic reactors resumed in the early 80's, and was greatly intensified by reports about Soviet ground and flight tests in the late 80's. This recent interest resulted in renewed U.S. thermionic reactor development programs, primarily under Department of Defense (DOD) and Department of Energy (DOE) sponsorship. Since most current investigators have not had an opportunity to study all of the author's previous work, a review of the highlights of that work may be of value to them. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling. Where the author's concepts differed from the later Topaz-2 design was in the relative location of the emitter and the collector. Placing the fueled emitter on the outside of the cylindrical diodes permits much higher axial conductances to reduce ohmic losses in the electrodes of full-core-height diodes. Moreover, placing the fuel on the outside of the diode makes possible reactors with much higher fuel volume fractions, which enable power-flattened fast reactors scalable to very low power levels without the need for life-limiting hydride moderators or the use of efficiency-limiting driver fuel. In addition, with the fuel on the outside its swelling does not increase the emitter diameter or reduce the interelectrode gap. This should permit long lifetimes even with closer spacings, which can significantly improve the system efficiences. This was confirmed by coupled neutronic, thermal, thermionic, and electrical system analyses - some of which are presented in this paper - and by subsequent experiments. A companion paper presented next describes the fabrication and testing of full-scale converter elements, both fueled and unfueled, and summarizes the test results obtained. There is a duplicate copy in the file.

  16. REPRESENTATIVE SOURCE TERMS AND THE INFLUENCE OF REACTOR ATTRIBUTES ON FUNCTIONAL CONTAINMENT IN MODULAR HIGH-TEMPERATURE GAS-COOLED REACTORS

    SciTech Connect (OSTI)

    Petti, D. A.; Hobbins, R. R.; Lowry, Peter P.; Gougar, Hans

    2013-11-01T23:59:59.000Z

    Modular high-temperature gas-cooled reactors (MHTGRs) offer a high degree of passive safety. The low power density of the reactor and the high heat capacity of the graphite core result in slow transients that do not challenge the integrity of the robust TRISO fuel. Another benefit of this fuel form and the surrounding graphite is their superior ability to retain fission products under all anticipated normal and off-normal conditions, which limits reactor accident source terms to very low values. In this paper, we develop estimates of the source term for a generic MHTGR to illustrate the performance of the radionuclide barriers that comprise the MHTGR functional containment. We also examine the influence of initial fuel quality, fuel performance/failure, reactor outlet temperature, and retention outside of the reactor core on the resultant source term to the environment.

  17. Representative Source Terms and the Influence of Reactor Attributes on Functional Containment in Modular High-Temperature Gas-Cooled Reactors

    SciTech Connect (OSTI)

    D. A. Petti; Hans Gougar; Dick Hobbins; Pete Lowry

    2013-11-01T23:59:59.000Z

    Modular high-temperature gas-cooled reactors (MHTGRs) offer a high degree of passive safety. The low power density of the reactor and the high heat capacity of the graphite core result in slow transients that do not challenge the integrity of the robust TRISO fuel. Another benefit of this fuel form and the surrounding graphite is their superior ability to retain fission products under all anticipated normal and off-normal conditions, which limits reactor accident source terms to very low values. In this paper, we develop estimates of the source term for a generic MHTGR to illustrate the performance of the radionuclide barriers that comprise the MHTGR functional containment. We also examine the influence of initial fuel quality, fuel performance/failure, reactor outlet temperature, and retention outside of the reactor core on the resultant source term to the environment.

  18. Results from simulated upper-plenum aerosol transport tests

    SciTech Connect (OSTI)

    Wright, A.L.; Pattison, W.L.

    1986-01-01T23:59:59.000Z

    A series of eight aerosol transport experiments, designated as Aerosol Transport Tests (ATT) A101 through A108, has recently been completed at the Oak Ridge National Laboratory (ORNL). These tests provide a data base for validation of aerosol transport modeling used in the TRAP-MELT2 computer code (Jordan and Kuhlman, 1985), which was developed at Battelle Columbus Laboratories to calculate aerosol/fission-product transport in the reactor coolant system in postulated light-water reactor (LWR) core-melt accidents. Results from tests A103 and A104 have been summarized in a previous paper (Wright and Pattison, 1985a); the present paper discusses results from tests A105 through A108.

  19. Designing Reactors to Facilitate Decommissioning

    SciTech Connect (OSTI)

    Richard H. Meservey

    2006-06-01T23:59:59.000Z

    Critics of nuclear power often cite issues with tail-end-of-the-fuel-cycle activities as reasons to oppose the building of new reactors. In fact, waste disposal and the decommissioning of large nuclear reactors have proven more challenging than anticipated. In the early days of the nuclear power industry the design and operation of various reactor systems was given a great deal of attention. Little effort, however, was expended on end-of-the-cycle activities, such as decommissioning and disposal of wastes. As early power and test reactors have been decommissioned difficulties with end-of-the-fuel-cycle activities have become evident. Even the small test reactors common at the INEEL were not designed to facilitate their eventual decontamination, decommissioning, and dismantlement. The results are that decommissioning of these facilities is expensive, time consuming, relatively hazardous, and generates large volumes of waste. This situation clearly supports critics concerns about building a new generation of power reactors.

  20. Transport Reactor Development Unit Modification to Provide a Syngas Slipstream at Elevated Conditions to Enable Separation of 100 LB/D of Hydrogen by Hydrogen Separation Membranes Year - 6 Activity 1.15 - Development of a National Center for Hydrogen Technology

    SciTech Connect (OSTI)

    Schlasner, Steven

    2012-03-01T23:59:59.000Z

    Gasification of coal when associated with carbon dioxide capture and sequestration has the potential to provide low-cost as well as low-carbon hydrogen for electric power, fuels or chemicals production. The key element to the success of this concept is inexpensive, effective separation of hydrogen from carbon dioxide in synthesis gas. Many studies indicate that membrane technology is one of the most, if not the most, economical means of accomplishing separation; however, the advancement of hydrogen separation membrane technology is hampered by the absence of experience or demonstration that the technology is effective economically and environmentally at larger scales. While encouraging performance has been observed at bench scale (less than 12 lb/d hydrogen), it would be imprudent to pursue a largescale demonstration without testing at least one intermediate scale, such as 100 lb/d hydrogen. Among its many gasifiers, the Energy & Environmental Research Center is home to the transport reactor demonstration unit (TRDU), a unit capable of firing 200—500 lb/hr of coal to produce 400 scfm of synthesis gas containing more than 200 lb/d of hydrogen. The TRDU and associated downstream processing equipment has demonstrated the capability of producing a syngas over a wide range of temperatures and contaminant levels — some of which approximate conditions of commercial-scale gasifiers. Until this activity, however, the maximum pressure of the TRDU’ s product syngas was 120 psig, well below the 400+ psig pressures of existing large gasifiers. This activity installed a high-temperature compressor capable of accepting the range of TRDU products up to 450°F and compressing them to 500 psig, a pressure comparable to some large scale gasifiers. Thus, with heating or cooling downstream of the TRDU compressor, the unit is now able to present a near-raw to clean gasifier synthesis gas containing more than 100 lb/d of hydrogen at up to 500 psig over a wide range of temperatures to hydrogen separation membranes or other equipment for development and demonstration.

  1. Techniques to Access Databases and Integrate Data for Hydrologic Modeling

    SciTech Connect (OSTI)

    Whelan, Gene; Tenney, Nathan D.; Pelton, Mitchell A.; Coleman, Andre M.; Ward, Duane L.; Droppo, James G.; Meyer, Philip D.; Dorow, Kevin E.; Taira, Randal Y.

    2009-06-17T23:59:59.000Z

    This document addresses techniques to access and integrate data for defining site-specific conditions and behaviors associated with ground-water and surface-water radionuclide transport applicable to U.S. Nuclear Regulatory Commission reviews. Environmental models typically require input data from multiple internal and external sources that may include, but are not limited to, stream and rainfall gage data, meteorological data, hydrogeological data, habitat data, and biological data. These data may be retrieved from a variety of organizations (e.g., federal, state, and regional) and source types (e.g., HTTP, FTP, and databases). Available data sources relevant to hydrologic analyses for reactor licensing are identified and reviewed. The data sources described can be useful to define model inputs and parameters, including site features (e.g., watershed boundaries, stream locations, reservoirs, site topography), site properties (e.g., surface conditions, subsurface hydraulic properties, water quality), and site boundary conditions, input forcings, and extreme events (e.g., stream discharge, lake levels, precipitation, recharge, flood and drought characteristics). Available software tools for accessing established databases, retrieving the data, and integrating it with models were identified and reviewed. The emphasis in this review was on existing software products with minimal required modifications to enable their use with the FRAMES modeling framework. The ability of four of these tools to access and retrieve the identified data sources was reviewed. These four software tools were the Hydrologic Data Acquisition and Processing System (HDAPS), Integrated Water Resources Modeling System (IWRMS) External Data Harvester, Data for Environmental Modeling Environmental Data Download Tool (D4EM EDDT), and the FRAMES Internet Database Tools. The IWRMS External Data Harvester and the D4EM EDDT were identified as the most promising tools based on their ability to access and retrieve the required data, and their ability to integrate the data into environmental models using the FRAMES environment.

  2. Heat pipe cooled reactors for multi-kilowatt space power supplies

    SciTech Connect (OSTI)

    Ranken, W.A.; Houts, M.G.

    1995-01-01T23:59:59.000Z

    Three nuclear reactor space power system designs are described that demonstrate how the use of high temperature heat pipes for reactor heat transport, combined with direct conversion of heat to electricity, can result in eliminating pumped heat transport loops for both primary reactor cooling and heat rejection. The result is a significant reduction in system complexity that leads to very low mass systems with high reliability, especially in the power range of 1 to 20 kWe. In addition to removing heat exchangers, electromagnetic pumps, and coolant expansion chambers, the heat pipe/direct conversion combination provides such capabilities as startup from the frozen state, automatic rejection of reactor decay heat in the event of emergency or accidental reactor shutdown, and the elimination of single point failures in the reactor cooling system. The power system designs described include a thermoelectric system that can produce 1 to 2 kWe, a bimodal modification of this system to increase its power level to 5 kWe and incorporate high temperature hydrogen propulsion capability, and a moderated thermionic reactor concept with 5 to 20 kWe power output that is based on beryllium modules that thermally couple cylindrical thermionic fuel elements (TFEs) to radiator heat pipes.

  3. Fast quench reactor and method

    DOE Patents [OSTI]

    Detering, B.A.; Donaldson, A.D.; Fincke, J.R.; Kong, P.C.

    1998-05-12T23:59:59.000Z

    A fast quench reactor includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This ``freezes`` the desired end product(s) in the heated equilibrium reaction stage. 7 figs.

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

    SciTech Connect (OSTI)

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

    2012-07-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    David Petti; Jim Kinsey; Dave Alberstein

    2014-01-01T23:59:59.000Z

    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. Neutrino Oscillation Studies with Reactors

    E-Print Network [OSTI]

    Petr Vogel; Liangjian Wen; Chao Zhang

    2015-03-03T23:59:59.000Z

    Nuclear reactors are one of the most intense, pure, controllable, cost-effective, and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavors are quantum mechanical mixtures. Over the past several decades reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle $\\theta_{13}$. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

  7. Neutrino Oscillation Studies with Reactors

    E-Print Network [OSTI]

    Petr Vogel; Liangjian Wen; Chao Zhang

    2015-04-27T23:59:59.000Z

    Nuclear reactors are one of the most intense, pure, controllable, cost-effective, and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavors are quantum mechanical mixtures. Over the past several decades reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle $\\theta_{13}$. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

  8. Neutrino Oscillation Studies with Reactors

    E-Print Network [OSTI]

    Vogel, Petr; Zhang, Chao

    2015-01-01T23:59:59.000Z

    Nuclear reactors are one of the most intense, pure, controllable, cost-effective, and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavors are quantum mechanical mixtures. Over the past several decades reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle $\\theta_{13}$. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

  9. Progress Update: Reactor Disassembly Grouting

    ScienceCinema (OSTI)

    Cody, Tom

    2012-06-14T23:59:59.000Z

    Grouting the P&R reactors in order to remove these basins as an environmental threat. This will end the Cold War legacy and end the environmental footprint.

  10. Progress Update: Reactor Disassembly Grouting

    SciTech Connect (OSTI)

    Cody, Tom

    2010-01-01T23:59:59.000Z

    Grouting the P&R reactors in order to remove these basins as an environmental threat. This will end the Cold War legacy and end the environmental footprint.

  11. Neutrino oscillation studies with reactors

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

    Vogel, P. [California Inst. of Technology (CalTech), Pasadena, CA (United States). Kellog Radiation Lab.; Wen, L.J. [Chinese Academy of Sciences (CAS), Beijing (China). Inst. of High Energy Physics (IHEP); Zhang, C. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2015-04-27T23:59:59.000Z

    Nuclear reactors are one of the most intense, pure, controllable, cost-effective and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavours are quantum mechanical mixtures. Over the past several decades, reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle ?13. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

  12. Thermonuclear Reflect AB-Reactor

    E-Print Network [OSTI]

    Alexander Bolonkin

    2008-03-26T23:59:59.000Z

    The author offers a new kind of thermonuclear reflect reactor. The remarkable feature of this new reactor is a three net AB reflector, which confines the high temperature plasma. The plasma loses part of its energy when it contacts with the net but this loss can be compensated by an additional permanent plasma heating. When the plasma is rarefied (has a small density), the heat flow to the AB reflector is not large and the temperature in the triple reflector net is lower than 2000 - 3000 K. This offered AB-reactor has significantly less power then the currently contemplated power reactors with magnetic or inertial confinement (hundreds-thousands of kW, not millions of kW). But it is enough for many vehicles and ships and particularly valuable for tunnelers, subs and space apparatus, where air to burn chemical fuel is at a premium or simply not available. The author has made a number of innovations in this reactor, researched its theory, developed methods of computation, made a sample computation of typical project. The main point of preference for the offered reactor is its likely cheapness as a power source. Key words: Micro-thermonuclear reactor, Multi-reflex AB-thermonuclear reactor, Self-magnetic AB-thermonuclear reactor, aerospace thermonuclear engine.

  13. Grid Integration

    SciTech Connect (OSTI)

    Not Available

    2008-09-01T23:59:59.000Z

    Summarizes the goals and activities of the DOE Solar Energy Technologies Program efforts within its grid integration subprogram.

  14. Preapplication safety evaluation report for the Power Reactor Innovative Small Module (PRISM) liquid-metal reactor. Final report

    SciTech Connect (OSTI)

    Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A. [Nuclear Regulatory Commission, Washington, DC (United States). Associate Directorate for Advanced Reactors and License Renewal

    1994-02-01T23:59:59.000Z

    This preapplication safety evaluation report (PSER) presents the results of the preapplication desip review for die Power Reactor Innovative Small Module (PRISM) liquid-mew (sodium)-cooled reactor, Nuclear Regulatory Commission (NRC) Project No. 674. The PRISM conceptual desip was submitted by the US Department of Energy in accordance with the NRC`s ``Statement of Policy for the Regulation of Advanced Nuclear Power Plants`` (51 Federal Register 24643). This policy provides for the early Commission review and interaction with designers and licensees. The PRISM reactor desip is a small, modular, pool-type, liquid-mew (sodium)-cooled reactor. The standard plant design consists of dim identical power blocks with a total electrical output rating of 1395 MWe- Each power block comprises three reactor modules, each with a thermal rating of 471 MWt. Each module is located in its own below-grade silo and is co to its own intermediate heat transport system and steam generator system. The reactors utilize a metallic-type fuel, a ternary alloy of U-Pu-Zr. The design includes passive reactor shutdown and passive decay heat removal features. The PSER is the NRC`s preliminary evaluation of the safety features in the PRISM design, including the projected research and development programs required to support the design and the proposed testing needs. Because the NRC review was based on a conceptual design, the PSER did not result in an approval of the design. Instead it identified certain key safety issues, provided some guidance on applicable licensing criteria, assessed the adequacy of the preapplicant`s research and development programs, and concluded that no obvious impediments to licensing the PRISM design had been identified.

  15. OXYGEN TRANSPORT CERAMIC MEMBRANES

    SciTech Connect (OSTI)

    Dr. Sukumar Bandopadhyay; Dr. Nagendra Nagabhushana

    2003-01-01T23:59:59.000Z

    In the present quarter, the possibility of using a more complex interfacial engineering approach to the development of reliable and stable oxygen transport perovskite ceramic membranes/metal seals is discussed. Experiments are presented and ceramic/metal interactions are characterized. Crack growth and fracture toughness of the membrane in the reducing conditions are also discussed. Future work regarding this approach is proposed are evaluated for strength and fracture in oxygen gradient conditions. Oxygen gradients are created in tubular membranes by insulating the inner surface from the reducing environment by platinum foils. Fracture in these test conditions is observed to have a gradient in trans and inter-granular fracture as opposed to pure trans-granular fracture observed in homogeneous conditions. Fracture gradients are reasoned to be due to oxygen gradient set up in the membrane, variation in stoichiometry across the thickness and due to varying decomposition of the parent perovskite. The studies are useful in predicting fracture criterion in actual reactor conditions and in understanding the initial evolution of fracture processes.

  16. Reactor coolant pump flywheel

    DOE Patents [OSTI]

    Finegan, John Raymond; Kreke, Francis Joseph; Casamassa, John Joseph

    2013-11-26T23:59:59.000Z

    A flywheel for a pump, and in particular a flywheel having a number of high density segments for use in a nuclear reactor coolant pump. The flywheel includes an inner member and an outer member. A number of high density segments are provided between the inner and outer members. The high density segments may be formed from a tungsten based alloy. A preselected gap is provided between each of the number of high density segments. The gap accommodates thermal expansion of each of the number of segments and resists the hoop stress effect/keystoning of the segments.

  17. Nuclear reactor control assembly

    SciTech Connect (OSTI)

    Negron, S.B.

    1991-06-11T23:59:59.000Z

    This patent describes an assembly for providing global power control in a nuclear reactor having the core split into two halves. It comprises a disk assembly formed from at least two disks each machined with an identical surface hole pattern such that rotation of one disk relative to the other causes the hole pattern to open or close, the disk assembly being positioned substantially at the longitudinal center of and coaxial with the core halves; and means for rotating at least one of the disks relative to the other.

  18. Nuclear reactor control apparatus

    SciTech Connect (OSTI)

    Sridhar, B.N.

    1981-08-28T23:59:59.000Z

    Nuclear reactor core safety rod release apparatus comprises a control rod having a detent notch in the form of an annular peripheral recess at its upper end, a control rod support tube for raising and lowering the control rod under normal conditions, latches pivotally mounted on the control support tube with free ends thereof normally disposed in the recess in the control rod, and cam means for pivoting the latches out of the recess in the control rod when a scram condition occurs. One embodiment of the invention comprises an additonal magnetically-operated latch for releasing the control rod under two different conditions, one involving seismic shock.

  19. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, C.D.; Marasco, J.A.

    1996-02-27T23:59:59.000Z

    A fluidized bed reactor system is described which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary and tertiary particulate phases, continuously introduced and removed simultaneously in the cocurrent and countercurrent mode, act in a role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Means for introducing and removing the sorbent phases include feed screw mechanisms and multivane slurry valves. 3 figs.

  20. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, Charles D. (Oak Ridge, TN)

    1993-01-01T23:59:59.000Z

    A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves.

  1. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, Charles D. (Oak Ridge, TN); Marasco, Joseph A. (Kingston, TN)

    1995-01-01T23:59:59.000Z

    A fluidized bed reactor system utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves.

  2. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, Charles D. (Oak Ridge, TN); Marasco, Joseph A. (Kingston, TN)

    1996-01-01T23:59:59.000Z

    A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary and tertiary particulate phases, continuously introduced and removed simultaneously in the cocurrent and countercurrent mode, act in a role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Means for introducing and removing the sorbent phases include feed screw mechanisms and multivane slurry valves.

  3. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, C.D.

    1993-12-14T23:59:59.000Z

    A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase is described. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves. 3 figures.

  4. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, C.D.; Marasco, J.A.

    1995-04-25T23:59:59.000Z

    A fluidized bed reactor system utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves. 3 figs.

  5. Nuclear reactor control apparatus

    DOE Patents [OSTI]

    Sridhar, Bettadapur N. (Cupertino, CA)

    1983-11-01T23:59:59.000Z

    Nuclear reactor core safety rod release apparatus comprises a control rod having a detent notch in the form of an annular peripheral recess at its upper end, a control rod support tube for raising and lowering the control rod under normal conditions, latches pivotally mounted on the control support tube with free ends thereof normally disposed in the recess in the control rod, and cam means for pivoting the latches out of the recess in the control rod when a scram condition occurs. One embodiment of the invention comprises an additional magnetically-operated latch for releasing the control rod under two different conditions, one involving seismic shock.

  6. Stochastic Transport

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What'sis Taking Over Our Instagram Secretary900Steep SlopeStochastic Domain-Wall

  7. Fast Reactor Fuel Type and Reactor Safety Performance

    SciTech Connect (OSTI)

    R. Wigeland; J. Cahalan

    2009-09-01T23:59:59.000Z

    Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of ‘inherent safety’ concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and thermophysical properties of the fuel and their compatibility with the reactor coolant, with corresponding differences in the challenges presented to the reactor developers. Accident phenomena are discussed for the sodium-cooled fast reactor based on the mechanistic progression of conditions from accident initiation to accident termination, whether a benign state is achieved or more severe consequences are expected. General principles connecting accident phenomena and fuel properties are developed from the oxide and metal fuel safety analyses, providing guidelines that can be used as part of the evaluation for selection of fuel type for the sodium-cooled fast reactor.

  8. Integration of Hydra-TH in VERA (L2 Milestone THM.CFD.P5.01)

    SciTech Connect (OSTI)

    Christon, Mark A. [Los Alamos National Laboratory; Baksoi, Jozsef [Los Alamos National Laboratory; Barnett, Nathan [formerly in CCS-2; Francois, Marianne M. [Los Alamos National Laboratory; Lowrie, Robert B. [Los Alamos National Laboratory; Sankaran, Ramanan [Scientific Computing Group, Oak Ridge National Laboratory

    2012-06-21T23:59:59.000Z

    This report describes the work carried out for completion of the Thermal Hydraulics Methods (THM) Level 2 Milestone THM.CFD.P5.01 for the Consortium for Advanced Simulation of Light Water Reactors (CASL). This milestone focused primarily on the initial integration of Hydra-TH in VERA. The primary objective for this milestone was the integration of Hydra-TH as a standalone executable in VERA. A series of code extensions/modifications have been made to Hydra-TH to facilitate integration of Hydra-TH in VERA and to permit future tighter integration and physics coupling. A total of 61 serial and 64 parallel regression tests have been supplied with Hydra-TH. These tests are are being executed in the TriBITS environment. Once the VERA team enables the full suite of tests, the results can be posted to the VERA CDash site. Future work will consider the use of the LIME 2.0 interface for tighter integration in VERA with additional efforts focused on multiphysics coupling with radiation transport, fuel performance, and solid/structural mechanics.

  9. Granular Dynamics in Pebble Bed Reactor Cores

    E-Print Network [OSTI]

    Laufer, Michael Robert

    2013-01-01T23:59:59.000Z

    in a pebble-bed nuclear reactor,” Phys. Rev. E, vol. 74, no.cycles of the pebble bed reactor,” Nuclear Engineering andoptimization of pebble-bed reactors,” Annals of Nuclear

  10. On-Board Vehicle, Cost Effective Hydrogen Enhancement Technology for Transportation PEM Fuel Cells

    SciTech Connect (OSTI)

    Thomas H. Vanderspurt; Zissis Dardas; Ying She; Mallika Gummalla; Benoit Olsommer

    2005-12-30T23:59:59.000Z

    Final Report of On-Board Vehicle, Cost Effective Hydrogen Enhancement Technology for Transportation PEM Fuel Cells. The objective of this effort was to technologically enable a compact, fast start-up integrated Water Gas Shift-Pd membrane reactor for integration into an On Board Fuel Processing System (FPS) for an automotive 50 kWe PEM Fuel Cell (PEM FC). Our approach was to: (1) use physics based reactor and system level models to optimize the design through trade studies of the various system design and operating parameters; and (2) synthesize, characterize and assess the performance of advanced high flux, high selectivity, Pd alloy membranes on porous stainless steel tubes for mechanical strength and robustness. In parallel and not part of this program we were simultaneously developing air tolerant, high volumetric activity, thermally stable Water Gas Shift catalysts for the WGS/membrane reactor. We identified through our models the optimum WGS/membrane reactor configuration, and best Pd membrane/FPS and PEM FC integration scheme. Such a PEM FC power plant was shown through the models to offer 6% higher efficiency than a system without the integrated membrane reactor. The estimated FPS response time was < 1 minute to 50% power on start-up, 5 sec transient response time, 1140 W/L power density and 1100 W/kg specific power with an estimated production cost of $35/kW. Such an FPS system would have a Catalytic Partial Oxidation System (CPO) rather than the slower starting Auto-Thermal Reformer (ATR). We found that at optimum WGS reactor configuration that H{sub 2} recovery efficiencies of 95% could be achieved at 6 atm WGS pressure. However optimum overall fuel to net electrical efficiency ({approx}31%) is highest at lower fuel processor efficiency (67%) with 85% H{sub 2} recovery because less parasitic power is needed. The H{sub 2} permeance of {approx}45 m{sup 3}/m{sup 2}-hr-atm{sup 0.5} at 350 C was assumed in these simulations. In the laboratory we achieved a H{sub 2} permeance of 50 m{sup 3}/(m{sup 2}-hr-atm{sup 0.5}) with a H{sub 2}/N{sub 2} selectivity of 110 at 350 C with pure Pd. We also demonstrated that we could produce Pd-Ag membranes. Such alloy membranes are necessary because they aren't prone to the Pd-hydride {alpha}-{beta} phase transition that is known to cause membrane failure in cyclic operation. When funding was terminated we were on track to demonstrated Pd-Ag alloy deposition on a nano-porous ({approx}80 nm) oxide layer supported on porous stainless steel tubing using a process designed for scale-up.

  11. Heavy Water Components Test Reactor Decommissioning - Major Component Removal

    SciTech Connect (OSTI)

    Austin, W.; Brinkley, D.

    2010-05-05T23:59:59.000Z

    The Heavy Water Components Test Reactor (HWCTR) facility (Figure 1) was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR facility is on high, well-drained ground, about 30 meters above the water table. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. It was not a defense-related facility like the materials production reactors at SRS. The reactor was moderated with heavy water and was rated at 50 megawatts thermal power. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In 1965, fuel assemblies were removed, systems that contained heavy water were drained, fluid piping systems were drained, deenergized and disconnected and the spent fuel basin was drained and dried. The doors of the reactor facility were shut and it wasn't until 10 years later that decommissioning plans were considered and ultimately postponed due to budget constraints. In the early 1990s, DOE began planning to decommission HWCTR again. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. The $1.6 billion allocation from the American Recovery and Reinvestment Act to SRS for site clean up at SRS has opened the doors to the HWCTR again - this time for final decommissioning. During the lifetime of HWCTR, 36 different fuel assemblies were tested in the facility. Ten of these experienced cladding failures as operational capabilities of the different designs were being established. In addition, numerous spills of heavy water occurred within the facility. Currently, radiation and radioactive contamination levels are low within HWCTR with most of the radioactivity contained within the reactor vessel. There are no known insults to the environment, however with the increasing deterioration of the facility, the possibility exists that contamination could spread outside the facility if it is not decommissioned. An interior panoramic view of the ground floor elevation taken in August 2009 is shown in Figure 2. The foreground shows the transfer coffin followed by the reactor vessel and control rod drive platform in the center. Behind the reactor vessel is the fuel pool. Above the ground level are the polar crane and the emergency deluge tank at the top of the dome. Note the considerable rust and degradation of the components and the interior of the containment building. Alternative studies have concluded that the most environmentally safe, cost effective option for final decommissioning is to remove the reactor vessel, steam generators, and all equipment above grade including the dome. Characterization studies along with transport models have concluded that the remaining below grade equipment that is left in place including the transfer coffin will not contribute any significant contamination to the environment in the future. The below grade space will be grouted in place. A concrete cover will be placed over the remaining footprint and the groundwater will be monitored for an indefinite period to ensure compliance with environmental regulations. The schedule for completion of decommissioning is late FY2011. This paper describes the concepts planned in order to remove the major components including the dome, the reactor vessel (RV), the two steam generators (SG), and relocating the transfer coffin (TC).

  12. Evolution integrals

    E-Print Network [OSTI]

    Rocco Duvenhage

    2006-05-24T23:59:59.000Z

    A framework analogous to path integrals in quantum physics is set up for abstract dynamical systems in a W*-algebraic setting. We consider spaces of evolutions, defined in a specific way, of a W*-algebra A as an analogue of spaces of classical paths, and show how integrals over such spaces, which we call ``evolution integrals'', lead to dynamics in a Hilbert space on a ``higher level'' which is viewed as an analogue of quantum dynamics obtained from path integrals. The measures with respect to which these integrals are performed are projection valued.

  13. ITERATED INTEGRALS OF SUPERCONNECTIONS 0902a KIYOSHI IGUSA

    E-Print Network [OSTI]

    Igusa, Kiyoshi

    ITERATED INTEGRALS OF SUPERCONNECTIONS 0902a KIYOSHI IGUSA Abstract. Iterated integration of a superconnection gives something else which we call a "superconnection parallel transport." We ask under what the superconnection parallel transports give homotopies and higher homotopies of these chain maps. This happens

  14. Gaseous reactor control system

    SciTech Connect (OSTI)

    Abdel-Khalik, S.

    1991-09-03T23:59:59.000Z

    This paper describes a nuclear reactor control system for controlling the reactivity of the core of a nuclear reactor. It includes a control gas having a high neutron cross-section; a first tank containing a first supply of the control gas; a first conduit providing a first fluid passage extending into the core, the first conduit being operatively connected to communicate with the first tank; a first valve operatively connected to regulate the flow of the control gas between the first tank and the first conduit; a second conduit concentrically disposed around the first conduit such that a second fluid passage is defined between the outer surface of the first conduit and the inner surface of the second conduit; a second tank containing a second supply of the control gas, the second tank being operatively connected to communicate with the second fluid passage; a second supply valve operatively connected to regulate the flow of the control gas between the second tank and the second fluid passage.

  15. Materials Degradation in Light Water Reactors: Life After 60,???

    SciTech Connect (OSTI)

    Busby, Jeremy T [ORNL; Nanstad, Randy K [ORNL; Stoller, Roger E [ORNL; Feng, Zhili [ORNL; Naus, Dan J [ORNL

    2008-04-01T23:59:59.000Z

    Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high temperature water, stress, vibration, and an intense neutron field. Degradation of materials in this environment can lead to reduced performance, and in some cases, sudden failure. A recent EPRI-led study interviewed 47 US nuclear utility executives to gauge perspectives on long-term operation of nuclear reactors. Nearly 90% indicated that extensions of reactor lifetimes to beyond 60 years were likely. When polled on the most challenging issues facing further life extension, two-thirds cited plant reliability as the key issue with materials aging and cable/piping as the top concerns for plant reliability. Materials degradation within a nuclear power plant is very complex. There are many different types of materials within the reactor itself: over 25 different metal alloys can be found with can be found within the primary and secondary systems, not to mention the concrete containment vessel, instrumentation and control, and other support facilities. When this diverse set of materials is placed in the complex and harsh environment coupled with load, degradation over an extended life is indeed quite complicated. To address this issue, the USNRC has developed a Progressive Materials Degradation Approach (NUREG/CR-6923). This approach is intended to develop a foundation for appropriate actions to keep materials degradation from adversely impacting component integrity and safety and identify materials and locations where degradation can reasonably be expected in the future. Clearly, materials degradation will impact reactor reliability, availability, and potentially, safe operation. Routine surveillance and component replacement can mitigate these factors, although failures still occur. With reactor life extensions to 60 years or beyond or power uprates, many components must tolerate the reactor environment for even longer times. This may increase susceptibility for most components and may introduce new degradation modes. While all components (except perhaps the reactor vessel) can be replaced, it may not be economically favorable. Therefore, understanding, controlling, and mitigating materials degradation processes are key priorities for reactor operation, power uprate considerations, and life extensions. This document is written to give an overview of some of the materials degradation issues that may be key for extend reactor service life. A detailed description of all the possible forms of degradation is beyond the scope of this short paper and has already been described in other documents (for example, the NUREG/CR-6923). The intent of this document is to present an overview of current materials issues in the existing reactor fleet and a brief analysis of the potential impact of extending life beyond 60 years. Discussion is presented in six distinct areas: (1) Reactor pressure vessel; (2) Reactor core and primary systems; (3) Reactor secondary systems; (4) Weldments; (5) Concrete; and (6) Modeling and simulations. Following each of these areas, some research thrust directions to help identify and mitigate lifetime extension issues are proposed. Note that while piping and cabling are important for extended service, these components are discussed in more depth in a separate paper. Further, the materials degradation issues associated with fuel cladding and fuel assemblies are not discussed in this section as these components are replaced periodically and will not influence the overall lifetime of the reactor.

  16. Analysis of Rod Removal Transient Experiments in VVER Reactors at Zero Power

    SciTech Connect (OSTI)

    Difillippo, F.C.

    2000-05-07T23:59:59.000Z

    Within the context of the Fissile Materials Disposition Program of the U.S. Department of Energy we analyzed rod removal transient experiments performed at the Kurchatov Institute in a full-scale mockup of VVER reactors, The transients were started (via water inlet) in slightly (few cents) supercritical configurations with all the control rods withdrawn. After a few minutes, control rods banks or individual control rods were f and t inserted and later withdrawn (returning to the initial state). Available experimental data include the relative time profiles of nine incore and excore detectors. Because of the mild nature of the transients (very low power and no more than 2 $ reactivities) we decided to use a quasistatic approach. The time-dependent flux is factorized into two terms: a function of phase space, given by the solution of the static equation with parametric excitation; and a function of time, given by the solution of the point kinetic equations with time-dependent kinetics para meters. Due to the nature of the experiment, cold conditions, control rods withdrawn and critical state with water level, the power distributions, measured and calculated, are quite unusual, with the inner part of the core heavily shielded. Measured power levels at the center of the reactor are almost 20 times smaller than similar regions at the periphery. Transport and diffusion calculations of the power distributions are in reasonable agreement, so the division code BOLD-VENTURE was used to calculate the kinetics parameters and the relative changes of the detector field of view. The numerical integration of the time-dependent part of the solution was made with the LSODE package using ENDFIB-V and VI delayed neutron data. Very good results were obtained for the nine lime profiles.

  17. Solvent refined coal reactor quench system

    DOE Patents [OSTI]

    Thorogood, R.M.

    1983-11-08T23:59:59.000Z

    There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream. 1 fig.

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

    SciTech Connect (OSTI)

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

    2014-04-01T23:59:59.000Z

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

  19. Benchmark for evaluation and validation of reactor simulations (BEAVRS)

    SciTech Connect (OSTI)

    Horelik, N.; Herman, B.; Forget, B.; Smith, K. [Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 77 Massachusetts Avenue, Cambridge, MA 02139 (United States)

    2013-07-01T23:59:59.000Z

    Advances in parallel computing have made possible the development of high-fidelity tools for the design and analysis of nuclear reactor cores, and such tools require extensive verification and validation. This paper introduces BEAVRS, a new multi-cycle full-core Pressurized Water Reactor (PWR) depletion benchmark based on two operational cycles of a commercial nuclear power plant that provides a detailed description of fuel assemblies, burnable absorbers, in-core fission detectors, core loading patterns, and numerous in-vessel components. This benchmark enables analysts to develop extremely detailed reactor core models that can be used for testing and validation of coupled neutron transport, thermal-hydraulics, and fuel isotopic depletion. The benchmark also provides measured reactor data for Hot Zero Power (HZP) physics tests, boron letdown curves, and three-dimensional in-core flux maps from fifty-eight instrumented assemblies. Initial comparisons between calculations performed with MIT's OpenMC Monte Carlo neutron transport code and measured cycle 1 HZP test data are presented, and these results display an average deviation of approximately 100 pcm for the various critical configurations and control rod worth measurements. Computed HZP radial fission detector flux maps also agree reasonably well with the available measured data. All results indicate that this benchmark will be extremely useful in validation of coupled-physics codes and uncertainty quantification of in-core physics computational predictions. The detailed BEAVRS specification and its associated data package is hosted online at the MIT Computational Reactor Physics Group web site (http://crpg.mit.edu/), where future revisions and refinements to the benchmark specification will be made publicly available. (authors)

  20. IMPACT OF BEAM TRANSPORT METHOD ON CHAMBER AND DRIVER DESIGN FOR

    E-Print Network [OSTI]

    IMPACT OF BEAM TRANSPORT METHOD ON CHAMBER AND DRIVER DESIGN FOR HEAVY ION INERTIAL FUSION ENERGY D propagate in thick-liquid-wall, wetted-wall, and dry-wall chambers. KEYWORDS: heavy ion fusion, ion beam transport, reactor chamber design I. INTRODUCTION The U.S. heavy ion fusion ~HIF! program is working toward

  1. Reactor Neutrino Physics -- An Update

    E-Print Network [OSTI]

    Felix Boehm

    1999-06-18T23:59:59.000Z

    We review the status and the results of reactor neutrino experiments. Long baseline oscillation experiments at Palo Verde and Chooz have provided limits for the oscillation parameters while the recently proposed Kamland experiment at a baseline of more than 100km is now in the planning stage. We also describe the status of neutrino magnetic moment experiments at reactors.

  2. PROCESS SYSTEMS ENGINEERING Discovery of Transport and Reaction

    E-Print Network [OSTI]

    Linninger, Andreas A.

    .interscience.wiley.com). In distributed systems, transport phenomena coupled with chemical or metabolic reac- tions are functions of space, which can often not be achieved in the real processes. Moreover, in many biological systems and safety, their extrapolation to larger reactor dimen- sions or different configuration can become unreliable. Parameter

  3. Study of Fuel Cell Water Transport With Neutron Imaging

    E-Print Network [OSTI]

    detector system, constructing computer controlled fuel cell handling mechanism and optimizing dataStudy of Fuel Cell Water Transport With Neutron Imaging David Jacobson (NIST) Paul Huffman (NIST in fully assembled operating fuel cells. Develop a nuclear reactor based state of the art neutron imaging

  4. Granular Dynamics in Pebble Bed Reactor Cores

    E-Print Network [OSTI]

    Laufer, Michael Robert

    2013-01-01T23:59:59.000Z

    gas reactors with the effective heat transfer of a molten salt coolant and the passive natural circulation safety systems of sodium fast reactors.

  5. MOOSE simulating nuclear reactor CRUD buildup

    ScienceCinema (OSTI)

    None

    2014-07-21T23:59:59.000Z

    This simulation uses multiple physical models to show how the buildup of boron deposits on reactor fuel can affect performance and the reactor's power profile.

  6. MOOSE simulating nuclear reactor CRUD buildup

    SciTech Connect (OSTI)

    None

    2014-02-06T23:59:59.000Z

    This simulation uses multiple physical models to show how the buildup of boron deposits on reactor fuel can affect performance and the reactor's power profile.

  7. The Endurance Bioenergy Reactor | Argonne National Laboratory

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

    The Endurance Bioenergy Reactor Share Description Argonne biophysicist Dr. Philip Laible and Air Force Major Matt Michaud talks about he endurance bioenergy reactor-a device that...

  8. Fast reactors and nuclear nonproliferation

    SciTech Connect (OSTI)

    Avrorin, E.N. [Russian Federal Nuclear Center - Zababakhin Institute of Applied Physics, Snezhinsk (Russian Federation); Rachkov, V.I.; Chebeskov, A.N. [State Scientific Center of the Russian Federation - Institute for Physics and Power Engineering, Bondarenko Square, 1, Obninsk, Kaluga region, 249033 (Russian Federation)

    2013-07-01T23:59:59.000Z

    Problems are discussed with regard to nuclear fuel cycle resistance in fast reactors to nuclear proliferation risk due to the potential for use in military programs of the knowledge, technologies and materials gained from peaceful nuclear power applications. Advantages are addressed for fast reactors in the creation of a more reliable mode of nonproliferation in the closed nuclear fuel cycle in comparison with the existing fully open and partially closed fuel cycles of thermal reactors. Advantages and shortcomings are also discussed from the point of view of nonproliferation from the start with fast reactors using plutonium of thermal reactor spent fuel and enriched uranium fuel to the gradual transition using their own plutonium as fuel. (authors)

  9. Transporting particulate material

    DOE Patents [OSTI]

    Aldred, Derek Leslie (North Hollywood, CA); Rader, Jeffrey A. (North Hollywood, CA); Saunders, Timothy W. (North Hollywood, CA)

    2011-08-30T23:59:59.000Z

    A material transporting system comprises a material transporting apparatus (100) including a material transporting apparatus hopper structure (200, 202), which comprises at least one rotary transporting apparatus; a stationary hub structure (900) constraining and assisting the at least one rotary transporting apparatus; an outlet duct configuration (700) configured to permit material to exit therefrom and comprising at least one diverging portion (702, 702'); an outlet abutment configuration (800) configured to direct material to the outlet duct configuration; an outlet valve assembly from the material transporting system venting the material transporting system; and a moving wall configuration in the material transporting apparatus capable of assisting the material transporting apparatus in transporting material in the material transporting system. Material can be moved from the material transporting apparatus hopper structure to the outlet duct configuration through the at least one rotary transporting apparatus, the outlet abutment configuration, and the outlet valve assembly.

  10. Transportation Electrification

    SciTech Connect (OSTI)

    Schwendeman, Lawrence; Crouch, Alan

    2013-12-17T23:59:59.000Z

    This project has accomplished the following objectives: to address the critical need for technician training in new and emerging propulsion technologies by developing new courses, including information and training on electric vehicles, plug-in hybrid electric vehicles, and fuel cell vehicles; to integrate the new certificate with the existing Associate of Applied Science Degree and Certificate automotive degrees; to disseminate these leading edge courses throughout the Commonwealth of Virginia and neighboring Mid-Atlantic States; and to provide training opportunities for displaced workers and underrepresented populations seeking careers in the automotive industry.

  11. Compact Reversed-Field Pinch Reactors (CRFPR): preliminary engineering considerations

    SciTech Connect (OSTI)

    Hagenson, R.L.; Krakowski, R.A.; Bathke, C.G.; Miller, R.L.; Embrechts, M.J.; Schnurr, N.M.; Battat, M.E.; LaBauve, R.J.; Davidson, J.W.

    1984-08-01T23:59:59.000Z

    The unique confinement physics of the Reversed-Field Pinch (RFP) projects to a compact, high-power-density fusion reactor that promises a significant reduction in the cost of electricity. The compact reactor also promises a factor-of-two reduction in the fraction of total cost devoted to the reactor plant equipment (i.e., fusion power core (FPC) plus support systems). In addition to operational and developmental benefits, these physically smaller systems can operate economically over a range of total power output. After giving an extended background and rationale for the compact fusion approaches, key FPC subsystems for the Compact RFP Reactor (CRFPR) are developed, designed, and integrated for a minimum-cost, 1000-MWe(net) system. Both the problems and promise of the compact, high-power-density fusion reactor are quantitatively evaluated on the basis of this conceptual design. The material presented in this report both forms a framework for a broader, more expanded conceptual design as well as suggests directions and emphases for related research and development.

  12. Ion transport membrane module and vessel system

    DOE Patents [OSTI]

    Stein, VanEric Edward (Allentown, PA); Carolan, Michael Francis (Allentown, PA); Chen, Christopher M. (Allentown, PA); Armstrong, Phillip Andrew (Orefield, PA); Wahle, Harold W. (North Canton, OH); Ohrn, Theodore R. (Alliance, OH); Kneidel, Kurt E. (Alliance, OH); Rackers, Keith Gerard (Louisville, OH); Blake, James Erik (Uniontown, OH); Nataraj, Shankar (Allentown, PA); van Doorn, Rene Hendrik Elias (Obersulm-Willsbach, DE); Wilson, Merrill Anderson (West Jordan, UT)

    2008-02-26T23:59:59.000Z

    An ion transport membrane system comprising (a) a pressure vessel having an interior, an exterior, an inlet, and an outlet; (b) a plurality of planar ion transport membrane modules disposed in the interior of the pressure vessel and arranged in series, each membrane module comprising mixed metal oxide ceramic material and having an interior region and an exterior region, wherein any inlet and any outlet of the pressure vessel are in flow communication with exterior regions of the membrane modules; and (c) one or more gas manifolds in flow communication with interior regions of the membrane modules and with the exterior of the pressure vessel.The ion transport membrane system may be utilized in a gas separation device to recover oxygen from an oxygen-containing gas or as an oxidation reactor to oxidize compounds in a feed gas stream by oxygen permeated through the mixed metal oxide ceramic material of the membrane modules.

  13. Ion transport membrane module and vessel system

    DOE Patents [OSTI]

    Stein, VanEric Edward (Allentown, PA); Carolan, Michael Francis (Allentown, PA); Chen, Christopher M. (Allentown, PA); Armstrong, Phillip Andrew (Orefield, PA); Wahle, Harold W. (North Canton, OH); Ohrn, Theodore R. (Alliance, OH); Kneidel, Kurt E. (Alliance, OH); Rackers, Keith Gerard (Louisville, OH); Blake, James Erik (Uniontown, OH); Nataraj, Shankar (Allentown, PA); Van Doorn, Rene Hendrik Elias (Obersulm-Willsbach, DE); Wilson, Merrill Anderson (West Jordan, UT)

    2012-02-14T23:59:59.000Z

    An ion transport membrane system comprising (a) a pressure vessel having an interior, an exterior, an inlet, and an outlet; (b) a plurality of planar ion transport membrane modules disposed in the interior of the pressure vessel and arranged in series, each membrane module comprising mixed metal oxide ceramic material and having an interior region and an exterior region, wherein any inlet and any outlet of the pressure vessel are in flow communication with exterior regions of the membrane modules; and (c) one or more gas manifolds in flow communication with interior regions of the membrane modules and with the exterior of the pressure vessel. The ion transport membrane system may be utilized in a gas separation device to recover oxygen from an oxygen-containing gas or as an oxidation reactor to oxidize compounds in a feed gas stream by oxygen permeated through the mixed metal oxide ceramic material of the membrane modules.

  14. TXSAMC (transport cross sections from applied Monte Carlo): a new tool for generating shielded multigroup cross sections 

    E-Print Network [OSTI]

    Hiatt, Matthew Torgerson

    2009-06-02T23:59:59.000Z

    This thesis describes a tool called TXSAMC (Transport Cross Sections from Applied Monte Carlo) that produces shielded and homogenized multigroup cross sections for small fast reactor systems. The motivation for this tool comes from a desire...

  15. Automatic reactor power control for a pressurized water reactor

    SciTech Connect (OSTI)

    Jungin Choi (Kyungwon Univ. (Korea, Republic of)); Yungjoon Hah (Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of)); Unchul Lee (Seoul National Univ. (Korea, Republic of))

    1993-05-01T23:59:59.000Z

    An automatic reactor power control system is presented for a pressurized water reactor (PWR). The associated reactor control strategy is called mode K.' The new system implements a heavy-worth bank dedicated to axial shape control, independent of the existing regulating banks. The heavy bank provides a monotonic relationship between its motion and the axial shape change, which allows automatic control of the axial power distribution. Thus, the mode K enables precise regulation of both the reactivity and the power distribution, by using double closed-loop control of the reactor coolant temperature and the axial power difference. Automatic reactor power control permits the nuclear power plant to accommodate the load-follow operations, including frequency control, to respond to the grid requirements. The mode K reactor control concepts were tested using simulation responses of a Korean standardized 1,000-MW (electric) PWR. The simulation results illustrate that the mode K would be a practical reactor power control strategy for the increased automation of nuclear plants.

  16. RELAP5/MOD2 split reactor vessel model and steamline break analysis

    SciTech Connect (OSTI)

    Petelin, S.; Mavko, B.; Gortnar, O. (Univ. of Ljubljana, (Slovenia))

    1993-04-01T23:59:59.000Z

    A split reactor vessel model for the RELAP5/ MOD2 computer code is developed in an attempt to realize more realistic predictions of asymmetrical transients in a two-loop nuclear power plant. Based on this split reactor model, coolant mixing processes within the reactor vessel are examined. This study evaluates the model improvements in terms of thermal-hydraulic simulations of the reactor core inlet fluid condition and the consequent core behavior. Furthermore, the split reactor vessel model is introduced into an integral RELAP5/MOD2 power plant model, and a steamline break analysis is performed to determine the influence of the boron concentration in the boron injection tank on accident consequences.

  17. Passive air cooling of liquid metal-cooled reactor with double vessel leak accommodation capability

    DOE Patents [OSTI]

    Hunsbedt, A.; Boardman, C.E.

    1995-04-11T23:59:59.000Z

    A passive and inherent shutdown heat removal method with a backup air flow path which allows decay heat removal following a postulated double vessel leak event in a liquid metal-cooled nuclear reactor is disclosed. The improved reactor design incorporates the following features: (1) isolation capability of the reactor cavity environment in the event that simultaneous leaks develop in both the reactor and containment vessels; (2) a reactor silo liner tank which insulates the concrete silo from the leaked sodium, thereby preserving the silo`s structural integrity; and (3) a second, independent air cooling flow path via tubes submerged in the leaked sodium which will maintain shutdown heat removal after the normal flow path has been isolated. 5 figures.

  18. Passive air cooling of liquid metal-cooled reactor with double vessel leak accommodation capability

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA)

    1995-01-01T23:59:59.000Z

    A passive and inherent shutdown heat removal method with a backup air flow path which allows decay heat removal following a postulated double vessel leak event in a liquid metal-cooled nuclear reactor. The improved reactor design incorporates the following features: (1) isolation capability of the reactor cavity environment in the event that simultaneous leaks develop in both the reactor and containment vessels; (2) a reactor silo liner tank which insulates the concrete silo from the leaked sodium, thereby preserving the silo's structural integrity; and (3) a second, independent air cooling flow path via tubes submerged in the leaked sodium which will maintain shutdown heat removal after the normal flow path has been isolated.

  19. A method for measurement of delayed neutron parameters for liquid-metal-cooled power reactors

    SciTech Connect (OSTI)

    Vilim, R.B. [Argonne National Lab., IL (United States); Brock, R.W. [Babcock and Wilcox, Lynchburg, VA (United States)

    1996-06-01T23:59:59.000Z

    The trend toward increased reliance on passive features for power reactor safety makes it important to obtain the characteristics of the reactor system from measurements on the system. A method is described for solving for the delayed neutron parameters in a liquid-metal power reactor by fitting an analytic solution of the point-kinetics equations to the flux die-away from a dropped rod in an initially critical core. The method includes treatment of those conditions found in a power reactor that depart from those in a critical assembly experiment. These include a comparatively long rod drop time and a detector signal that instead of providing an integrated count rate is a sampled data signal proportional to the instantaneous fission power. The delayed neutron parameter values calculated from a rod drop experiment in the Experimental Breeder Reactor II are in agreement with values calculated using first principles and knowledge of core material composition and nuclear cross sections.

  20. Development of ceramic membrane reactors for high temperature gas cleanup. Final report

    SciTech Connect (OSTI)

    Roberts, D.L.; Abraham, I.C.; Blum, Y.; Gottschlich, D.E.; Hirschon, A.; Way, J.D.; Collins, J.

    1993-06-01T23:59:59.000Z

    The objective of this project was to develop high temperature, high pressure catalytic ceramic membrane reactors and to demonstrate the feasibility of using these membrane reactors to control gaseous contaminants (hydrogen sulfide and ammonia) in integrated gasification combined cycle (IGCC) systems. Our strategy was to first develop catalysts and membranes suitable for the IGCC application and then combine these two components as a complete membrane reactor system. We also developed a computer model of the membrane reactor and used it, along with experimental data, to perform an economic analysis of the IGCC application. Our results have demonstrated the concept of using a membrane reactor to remove trace contaminants from an IGCC process. Experiments showed that NH{sub 3} decomposition efficiencies of 95% can be achieved. Our economic evaluation predicts ammonia decomposition costs of less than 1% of the total cost of electricity; improved membranes would give even higher conversions and lower costs.