Sample records for iris reactor core

  1. Determination of a test section parameters for IRIS nuclear reactor pressurizer

    Microsoft Academic Search

    Mário Augusto Bezerra da Silva; Carlos Alberto Brayner de Oliveira Lira; Antonio Carlos de Oliveira Barroso

    2009-01-01

    An integral, modular and medium size nuclear reactor, known as IRIS, is being developed by Westinghouse and by research centers. IRIS is characterized by having most of its components inside the pressure vessel, eliminating the probability of accidents. Due to its integral configuration, there is no spray system for boron homogenization, which may cause power transients. Thus, boron mixing must

  2. Fissioning Plasma Core Reactor

    NASA Technical Reports Server (NTRS)

    Albright, Dennis; Butler, Carey; West, Nicole; Cole, John W. (Technical Monitor)

    2002-01-01

    Institute for Scientific Research, Inc. (ISR) research program consist of: 1.Study core physics by adapting existing codes: MCNP4C - Monte Carlo code; COMBINE/VENTURE - diffusion theory; SCALE4 - Monte Carlo, with many utility codes. 2. Determine feasibility and study major design parameters: fuel selection, temperature and reflector sizing. 3. Study reactor kinetics: develop QCALC1 to model point kinetics; study dynamic behavior of the power release.

  3. Plasma core reactor applications

    NASA Technical Reports Server (NTRS)

    Latham, T. S.; Rodgers, R. J.

    1976-01-01

    Analytical and experimental investigations were conducted to demonstrate the feasibility of fissioning uranium plasma core reactors and to characterize space and terrestrial applications for such reactors. Uranium hexafluoride fuel is injected into core cavities and confined away from the surface by argon buffer gas injected tangentially from the peripheral walls. Radiant heat transfer calculations were performed for a six-cavity reactor configuration. Axial working fluid channels are located along a fraction of each cavity peripheral wall. Results of calculations for outward-directed radiant energy fluxes corresponding to radiating temperatures of 2000 to 5000 K indicate total operating pressures from 80 to 650 atm, centerline temperatures from 6900 to 30,000 K, and total radiated powers from 25 to 2500 MW, respectively. Applications are described for this type of reactor such as (1) high-thrust, high specific impulse space propulsion, (2) highly efficient systems for generation of electricity, and (3) hydrogen or synthetic fuel production systems using the intense radiant energy fluxes.

  4. Reactor core isolation cooling system

    DOEpatents

    Cooke, F.E.

    1992-12-08

    A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom. 1 figure.

  5. Cogeneration of Electricity and Potable Water Using The International Reactor Innovative And Secure (IRIS) Design

    SciTech Connect

    Ingersoll, D.T.; Binder, J.L.; Kostin, V.I.; Panov, Y.K.; Polunichev, V.; Ricotti, M.E.; Conti, D.; Alonso, G.

    2004-10-06

    The worldwide demand for potable water has been steadily growing and is projected to accelerate, driven by a continued population growth and industrialization of emerging countries. This growth is reflected in a recent market survey by the World Resources Institute, which shows a doubling in the installed capacity of seawater desalination plants every ten years. The production of desalinated water is energy intensive, requiring approximately 3-6 kWh/m3 of produced desalted water. At current U.S. water use rates, a dedicated 1000 MW power plant for every one million people would be required to meet our water needs with desalted water. Nuclear energy plants are attractive for large scale desalination application. The thermal energy produced in a nuclear plant can provide both electricity and desalted water without the production of greenhouse gases. A particularly attractive option for nuclear desalination is to couple a desalination plant with an advanced, modular, passively safe reactor design. The use of small-to-medium sized nuclear power plants allows for countries with smaller electrical grid needs and infrastructure to add new electrical and water capacity in more appropriate increments and allows countries to consider siting plants at a broader number of distributed locations. To meet these needs, a modified version of the International Reactor Innovative and Secure (IRIS) nuclear power plant design has been developed for the cogeneration of electricity and desalted water. The modular, passively safe features of IRIS make it especially well adapted for this application. Furthermore, several design features of the IRIS reactor will ensure a safe and reliable source of energy and water even for countries with limited nuclear power experience and infrastructure. The IRIS-D design utilizes low-quality steam extracted from the low-pressure turbine to boil seawater in a multi-effect distillation desalination plant. The desalination plant is based on the horizontal tube film evaporation design used successfully with the BN-350 nuclear plant in Aktau, Kazakhstan. Parametric studies have been performed to optimize the balance of plant design. Also, an economic analysis has been performed, which shows that IRIS-D should be able to provide electricity and clean water at highly competitive costs.

  6. Research on plasma core reactors

    NASA Technical Reports Server (NTRS)

    Jarvis, G. A.; Barton, D. M.; Helmick, H. H.; Bernard, W.; White, R. H.

    1976-01-01

    Experiments and theoretical studies are being conducted for NASA on critical assemblies with one-meter diameter by one-meter long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17 cm thick by 89 cm diameter beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF6 container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials were measured. Finally, an 85,000 cu cm aluminum canister in the central region was fueled with UF6 gas and fission density distributions determined. These results are to be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation.

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

    SciTech Connect

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

    2011-05-31

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

  8. Modeling and Experimental Tests on the Hydraulically Driven Control Rod option for IRIS Reactor

    SciTech Connect

    Cammi, Antonio; Ricotti, Marco E.; Vitulo, Alessia [Department of Nuclear Engineering, Politecnico di Milano, Via Ponzio, 34/3, 20133 Milano (Italy)

    2004-07-01

    The adoption of Internal Control Rod Drive Mechanisms (ICRDMs) represents a valuable alternative to classical, external CRDMs based on electro-magnetic devices, as adopted in current PWRs. The advantages on the safety features of the reactor are apparent: inherent elimination of the Rod Ejection accidents and of possible concerns about the vessel head penetrations. A further positive feedback on the design is the reduction of the primary system overall dimensions. Within the frame of the ICRDM concepts, the Hydraulically Driven Control Rod solution is investigated as a possible option for the IRIS integral reactor. After a brief comparison of the solutions currently proposed for integral reactors, the configuration of the Hydraulic Control Rod device for IRIS, made up by an external movable piston and an internal fixed cylinder, is described. A description of the whole control system is reported as well. Particular attention is devoted to the Control Rod profile characterization, performed by means of a Computational Fluid Dynamics (CFD) analysis. The investigation of the system behavior has been carried out, including the dynamic equilibrium and its stability properties, the withdrawal and insertion step movement and the sensitivity study on command time periods. A suitable dynamic model has been set up for the mentioned purposes: the models corresponding to the various Control Rod system devices have been written in an Object-Oriented language (Modelica), thus allowing an easy implementation of such a system into the simulator for the whole reactor. Finally, a preliminary low pressure, low temperature, reduced length experimental facility has been built. Tests on HDCR stability and operational transients have been performed. The results are compared with the dynamic system model and CFD simulation model, showing good agreement between simulations and experimental data. During these preliminary tests, the control system performed correctly, allowing stable dynamic equilibrium positions for the Control Rod and stable behavior during withdrawal and insertion steps. (authors)

  9. The solar core rotation from LOWL and IRIS or BiSON data T. Corbard 1 , G. Berthomieu 1 , J. Provost 1 , E. Fossat 2

    E-print Network

    Corbard, Thierry

    The solar core rotation from LOWL and IRIS or BiSON data T. Corbard 1 , G. Berthomieu 1 , J published splittings of the IRIS and the BiSON network groups. These two sets have been known by different methods applied on the same averaged spectra (Lazrek et al. 1996). The BiSON data have been

  10. Lateral restraint assembly for reactor core

    DOEpatents

    Gorholt, Wilhelm (San Diego, CA); Luci, Raymond K. (Del Mar, CA)

    1986-01-01

    A restraint assembly for use in restraining lateral movement of a reactor core relative to a reactor vessel wherein a plurality of restraint assemblies are interposed between the reactor core and the reactor vessel in circumferentially spaced relation about the core. Each lateral restraint assembly includes a face plate urged against the outer periphery of the core by a plurality of compression springs which enable radial preloading of outer reflector blocks about the core and resist low-level lateral motion of the core. A fixed radial key member cooperates with each face plate in a manner enabling vertical movement of the face plate relative to the key member but restraining movement of the face plate transverse to the key member in a plane transverse to the center axis of the core. In this manner, the key members which have their axes transverse to or subtending acute angles with the direction of a high energy force tending to move the core laterally relative to the reactor vessel restrain such lateral movement.

  11. ANALYSIS OF SEPCTRUM CHOICES FOR SMALL MODULAR REACTORS-PERFORMANCE AND DEVELOPMENT

    E-print Network

    Kafle, Nischal

    2011-04-26

    and for keeping me motivated. v NOMENCLATURE BWR Boiling Water Reactor FP Fission Products LWR Light Water Reactor IRIS International Reactor Innovative and Secure MA Minor actinides MHR Modular High Temperature Reactor MWe Megawatt Electric 239Pu... ........................................................................................... 30 vii LIST OF FIGURES FIGURE Page 1 The reactor schematic of International reactor innovative and secure (IRIS) ........ 4 2 The reactor schematic of Pebble bed modular reactor (PBMR) ............................ 6 3 Reactor core...

  12. Gas core reactors for coal gasification

    NASA Technical Reports Server (NTRS)

    Weinstein, H.

    1976-01-01

    The concept of using a gas core reactor to produce hydrogen directly from coal and water is presented. It is shown that the chemical equilibrium of the process is strongly in favor of the production of H2 and CO in the reactor cavity, indicating a 98% conversion of water and coal at only 1500 K. At lower temperatures in the moderator-reflector cooling channels the equilibrium strongly favors the conversion of CO and additional H2O to CO2 and H2. Furthermore, it is shown the H2 obtained per pound of carbon has 23% greater heating value than the carbon so that some nuclear energy is also fixed. Finally, a gas core reactor plant floating in the ocean is conceptualized which produces H2, fresh water and sea salts from coal.

  13. Gas-core reactor power transient analysis.

    NASA Technical Reports Server (NTRS)

    Kascak, A. F.

    1972-01-01

    The nuclear fuel in the gas-core reactor concept is a ball of uranium plasma radiating thermal photons. The photons are met by an inflowing hydrogen stream, which is seeded with submicron size, depleted uranium particles. A 'wall-burnout' condition exists if the thermal photons can reach the cavity liner because of insufficient absorption by the hydrogen. An analysis was conducted in order to determine the time for which the maximum steady state reactor power could be exceeded without damage to the cavity liner due to burnout. Wall-burnout time as a function of the power increase above the initial steady state condition is shown in a graph.

  14. Conceptual Design of a Modular Island Core Fast Breeder Reactor \\

    Microsoft Academic Search

    Mitsuru KAMBE

    2002-01-01

    A metal fueled modular island core sodium cooled fast breeder reactor concept RAPID-M to improve reactor per- formance and proliferation resistance and to accommodate various power requirements has been demonstrated. The essential feature of the RAPID-M concept is that the reactor core consists of integrated fuel assemblies (IFAs) instead of conventional fuel subassemblies. The RAPID concept enables quick and simplified

  15. Sodium fast reactor evaluation: Core materials

    NASA Astrophysics Data System (ADS)

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

    2009-07-01

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

  16. Gas-core reactor power transient analysis

    NASA Technical Reports Server (NTRS)

    Kascak, A. F.

    1972-01-01

    The gas core reactor is a proposed device which features high temperatures. It has applications in high specific impulse space missions, and possibly in low thermal pollution MHD power plants. The nuclear fuel is a ball of uranium plasma radiating thermal photons as opposed to gamma rays. This thermal energy is picked up before it reaches the solid cavity liner by an inflowing seeded propellant stream and convected out through a rocket nozzle. A wall-burnout condition will exist if there is not enough flow of propellant to convect the energy back into the cavity. A reactor must therefore operate with a certain amount of excess propellant flow. Due to the thermal inertia of the flowing propellant, the reactor can undergo power transients in excess of the steady-state wall burnout power for short periods of time. The objective of this study was to determine how long the wall burnout power could be exceeded without burning out the cavity liner. The model used in the heat-transfer calculation was one-dimensional, and thermal radiation was assumed to be a diffusion process.

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

    Microsoft Academic Search

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

    2011-01-01

    Development and deployment of small-scale nuclear power reactors and their maintenance, monitoring, and control are part of the mission under the Small Modular Reactor (SMR) program. The objectives of this NERI-consortium research project are to investigate, develop, and validate advanced methods for sensing, controlling, monitoring, diagnosis, and prognosis of these reactors, and to demonstrate the methods with application to one

  18. Applications of plasma core reactors to terrestrial energy systems

    NASA Technical Reports Server (NTRS)

    Latham, T. S.; Biancardi, F. R.; Rodgers, R. J.

    1974-01-01

    Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrial applications are described. Closed-cycle gas turbines, MHD systems, photo- and thermo-chemical hydrogen production processes, and laser systems using plasma core reactors as prime energy sources are considered. Cycle efficiencies in the range of 50 to 65 percent are calculated for closed-cycle gas turbine and MHD electrical generators. Reactor advantages include continuous fuel reprocessing which limits inventory of radioactive by-products and thorium-U-233 breeder configurations with about 5-year doubling times.-

  19. REACTOR CORE SURROUNDED BY BERYLLIUM MODERATOR. CAMERA LOOKS DOWN AND ...

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

    REACTOR CORE SURROUNDED BY BERYLLIUM MODERATOR. CAMERA LOOKS DOWN AND TOWARD NORTH INTO LOWER GRID CASTING. HOLES OF VARIOUS SIZES ACCOMMODATE COOLANT WATER AND EXPERIMENTAL POSITIONS. INL NEGATIVE NO. 4197. Unknown Photographer, 2/11/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  20. Hanging core support system for a nuclear reactor. [LMFBR

    DOEpatents

    Burelbach, J.P.; Kann, W.J.; Pan, Y.C.; Saiveau, J.G.; Seidensticker, R.W.

    1984-04-26

    For holding the reactor core in the confining reactor vessel, a support is disclosed that is structurally independent of the vessel, that is dimensionally accurate and stable, and that comprises tandem tension linkages that act redundantly of one another to maintain stabilized core support even in the unlikely event of the complete failure of one of the linkages. The core support has a mounting platform for the reactor core, and unitary structure including a flange overlying the top edge of the reactor vessels, and a skirt and box beams between the flange and platform for establishing one of the linkages. A plurality of tension rods connect between the deck closing the reactor vessel and the platform for establishing the redundant linkage. Loaded Belleville springs flexibly hold the tension rods at the deck and separable bayonet-type connections hold the tension rods at the platform.

  1. Nuclear waste disposal utilizing a gaseous core reactor

    NASA Technical Reports Server (NTRS)

    Paternoster, R. R.

    1975-01-01

    The feasibility of a gaseous core nuclear reactor designed to produce power to also reduce the national inventories of long-lived reactor waste products through nuclear transmutation was examined. Neutron-induced transmutation of radioactive wastes is shown to be an effective means of shortening the apparent half life.

  2. Core design of the upgraded TREAT reactor

    Microsoft Academic Search

    D. C. Wade; S. K. Bhattacharyya; W. C. Lipinski; C. C. Stone

    1982-01-01

    The upgrading of the TREAT reactor involves the replacement of the central 11 x 11 subzone of the 19 x 19 fuel assembly array by new, Inconel-clad, high-temperature fuel assemblies, and the additions of a new reactor control system, a safety-grade plant protection system, and an enhanced reactor filtration\\/coolant system. The final design of these modifications will be completed in

  3. Identifying test methods for breeder reactor core materials

    SciTech Connect

    Not Available

    1981-01-01

    The practice covers the test methods that may be used for measuring physical and mechanical properties for solid metallic materials for use in the design and evaluation of fast breeder reactor core materials. The test methods referenced are applicable to the in-core portion of ducts, cladding (both end caps and tubing), wire wrap, spacer grids, wear pads, and welded structures. (JMT)

  4. DEVELOPMENT OF CORE ELEMENTS FOR THE ENRICO FERMI POWER REACTOR

    Microsoft Academic Search

    W. N. McDaniel; O. E. Homeister; D. O. Leeser

    1958-01-01

    Various core element binary alloys and configurations have been covered ; in the core element fabrication development program for the Enrico Fermi Reactor. ; Atloys of U-Cr, U-Zr and U-Mo were considered. These with good casting qualities ; were cast into flat plates, corrugated plates, and even large castings with ; integral coolant channels. The U-3.5 wt. % Mo alloy

  5. Shield Design for a Space Based Vapor Core Reactor

    SciTech Connect

    Knight, Travis; Anghaie, Samim [Innovative Nuclear Space Power and Propulsion Institute (INSPI), PO Box 116502, University of Florida, Gainesville, FL 32611-6502 (United States)

    2002-07-01

    Innovative shielding strategies were sought to reduce the mass of the required shielding for a space based vapor core reactor system with magnetohydrodynamic energy conversion. Gamma-rays directly resultant from fission were found to play no role in the dose rate, while secondary gamma-rays from fission neutron interactions were the dominant contributor to the dose rate. Hydrogen containing materials such as polyethylene were utilized to provide shielding of both radiation from the reactor complex and also solar and galactic cosmic radiation. This shield design was found to contribute 0.125 kg/kWe to the baseline vapor core reactor system specific mass. (authors)

  6. Identification and control of a nuclear reactor core (VVER) using recurrent neural networks and fuzzy systems

    Microsoft Academic Search

    Mehrdad Boroushaki; Mohammad B. Ghofrani; Caro Lucas; Mohammad J. Yazdanpanah

    2003-01-01

    Improving the methods of identification and control of nuclear power reactors core is an important area in nuclear engineering. Controlling the nuclear reactor core during load following operation encounters some difficulties in control of core thermal power while considering the core limitations in local power peaking and safety margins. In this paper, a nuclear power reactor core (VVER) is identified

  7. Core design of the upgraded TREAT reactor

    SciTech Connect

    Wade, D.C.; Bhattacharyya, S.K.; Lipinski, W.C.; Stone, C.C.

    1982-01-01

    The upgrading of the TREAT reactor involves the replacement of the central 11 x 11 subzone of the 19 x 19 fuel assembly array by new, Inconel-clad, high-temperature fuel assemblies, and the additions of a new reactor control system, a safety-grade plant protection system, and an enhanced reactor filtration/coolant system. The final design of these modifications will be completed in early 1983. The TREAT facility is scheduled to be shut down for modification in mid-1984, and should resume the safety test program in mid-1985. The upgrading will provide a capability to conduct fast reactor safety tests on clusters of up to 37 prototypic LMFBR pins.

  8. Automated Design and Optimization of Pebble-bed Reactor Cores

    SciTech Connect

    Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

    2010-07-01

    We present a conceptual design approach for high-temperature gas-cooled reactors using recirculating pebble-bed cores. The design approach employs PEBBED, a reactor physics code specifically designed to solve for and analyze the asymptotic burnup state of pebble-bed reactors, in conjunction with a genetic algorithm to obtain a core that maximizes a fitness value that is a function of user-specified parameters. The uniqueness of the asymptotic core state and the small number of independent parameters that define it suggest that core geometry and fuel cycle can be efficiently optimized toward a specified objective. PEBBED exploits a novel representation of the distribution of pebbles that enables efficient coupling of the burnup and neutron diffusion solvers. With this method, even complex pebble recirculation schemes can be expressed in terms of a few parameters that are amenable to modern optimization techniques. With PEBBED, the user chooses the type and range of core physics parameters that represent the design space. A set of traits, each with acceptable and preferred values expressed by a simple fitness function, is used to evaluate the candidate reactor cores. The stochastic search algorithm automatically drives the generation of core parameters toward the optimal core as defined by the user. The optimized design can then be modeled and analyzed in greater detail using higher resolution and more computationally demanding tools to confirm the desired characteristics. For this study, the design of pebble-bed high temperature reactor concepts subjected to demanding physical constraints demonstrated the efficacy of the PEBBED algorithm.

  9. Assessment of HCDA energetics in the CRBRP heterogeneous reactor core

    SciTech Connect

    Rhow, S K; Switick, D M; McElroy, J L; Joe, B W; Elawar, Z J

    1981-03-27

    The results of hypothetical core disruptive event analyses for the CRBRP heterogeneous reactor core are reported. The analytical results cover a large number of parametric cases including variations in design parameters and phenomenological assumptions. Reactor core configurations at the beginning of cycle one and end of cycle four are evaluated. The energetic consequences are evaluated based upon both fuel expansion thermodynamic work potential and a relative probability assignment. It is concluded that the structural loads, which result from 101 megajoules of available expansion work at sodium slug impact on the reactor closure head (equivalent to 661 megajoules of fuel expansion work to one atmosphere), is an adequate energetic consequence envelope for use in specifying the Structural Margins Beyond the Design Basis.

  10. Technical description of the heterogeneous gas core reactor

    SciTech Connect

    Diaz, N.J.; Dugan, E.T.; Han, K.I.

    1986-01-01

    Solid-fueled and liquid-fueled heterogeneous nuclear reactors have already undergone extensive analysis and development. Gas-fueled nuclear reactor studies, however, have been restricted to either (1) homogeneous core systems, (2) two large separate regions of fuel and moderator; and (3) concentric rings of fuel-moderator arrangements. The gas-fueled heterogeneous core reactors consist of an array of coolant/moderator channels in a vessel containing a fissionable gas or a mixture of fissionable and non-fissionable gases. The moderator/coolant is, therefore, distributed in the core but physically separated from the gas. It is this moderator/coolant distribution in a variable lattice arrangement surrounded by the gas which constitutes the bases of this concept. The results of this arrangement on the neutron and fuel economy, power distribution, power density, and heat transfer characteristics have been found to be extremely advantageous. 13 figs.

  11. Core Physics and Kinetics Calculations for the Fissioning Plasma Core Reactor

    NASA Technical Reports Server (NTRS)

    Butler, C.; Albright, D.

    2007-01-01

    Highly efficient, compact nuclear reactors would provide high specific impulse spacecraft propulsion. This analysis and numerical simulation effort has focused on the technical feasibility issues related to the nuclear design characteristics of a novel reactor design. The Fissioning Plasma Core Reactor (FPCR) is a shockwave-driven gaseous-core nuclear reactor, which uses Magneto Hydrodynamic effects to generate electric power to be used for propulsion. The nuclear design of the system depends on two major calculations: core physics calculations and kinetics calculations. Presently, core physics calculations have concentrated on the use of the MCNP4C code. However, initial results from other codes such as COMBINE/VENTURE and SCALE4a. are also shown. Several significant modifications were made to the ISR-developed QCALC1 kinetics analysis code. These modifications include testing the state of the core materials, an improvement to the calculation of the material properties of the core, the addition of an adiabatic core temperature model and improvement of the first order reactivity correction model. The accuracy of these modifications has been verified, and the accuracy of the point-core kinetics model used by the QCALC1 code has also been validated. Previously calculated kinetics results for the FPCR were described in the ISR report, "QCALC1: A code for FPCR Kinetics Model Feasibility Analysis" dated June 1, 2002.

  12. Fission rate measurements in fuel plate type assembly reactor cores

    SciTech Connect

    Rogers, J.W.

    1988-01-01

    The methods, materials and equipment have been developed to allow extensive and precise measurement of fission rate distributions in water moderated, U-Al fuel plate assembly type reactor cores. Fission rate monitors are accurately positioned in the reactor core, the reactor is operated at a low power for a short time, the fission rate monitors are counted with detectors incorporating automated sample changers and the measurements are converted to fission rate distributions. These measured fission rate distributions have been successfully used as baseline information related to the operation of test and experimental reactors with respect to fission power and distribution, fuel loading and fission experiments for approximately twenty years at the Idaho National Engineering Laboratory (INEL). 7 refs., 8 figs.

  13. Modification of the Core Cooling System of TRIGA 2000 Reactor

    SciTech Connect

    Umar, Efrizon; Fiantini, Rosalina [National Nuclear Energy Agency of Indonesia, Jalan Tamansari 71, Bandung, 40132 (Indonesia)

    2010-06-22

    To accomplish safety requirements, a set of actions has to be performed following the recommendations of the IAEA safety series 35 applied to research reactor. Such actions are considered in modernization of the old system, improving the core cooling system and safety evaluations. Due to the complexity of the process and the difficulty in putting the apparatus in the reactor core, analytical and experimental study on the determination of flow and temperature distribution in the whole coolant channel are difficult to be done. In the present work, a numerical study of flow and temperature distribution in the coolant channel of TRIGA 2000 has been carried out using CFD package. For this study, simulations were carried out on 3-D tested model. The model consists of the reactor tank, thermal and thermalizing column, reflector, rotary specimen rack, chimney, fuel element, primary pipe, diffuser, beam tube and a part of the core are constructed by 1.50 million unstructured tetrahedral cell elements. The results show that for the initial condition (116 fuel elements in the core) and for the inlet temperature of 24 deg. C and the primary velocity of 5.6 m/s, there no boiling phenomena occur in the coolant channel. Due to this result, it is now possible to improve the core cooling system of TRIGA 2000 reactor. Meanwhile, forced flow from the diffuser system only affected the flow pattern in the outside of chimney and put on a small effect to the fluid flow's velocity in the inside of chimney.

  14. Modification of the Core Cooling System of TRIGA 2000 Reactor

    NASA Astrophysics Data System (ADS)

    Umar, Efrizon; Fiantini, Rosalina

    2010-06-01

    To accomplish safety requirements, a set of actions has to be performed following the recommendations of the IAEA safety series 35 applied to research reactor. Such actions are considered in modernization of the old system, improving the core cooling system and safety evaluations. Due to the complexity of the process and the difficulty in putting the apparatus in the reactor core, analytical and experimental study on the determination of flow and temperature distribution in the whole coolant channel are difficult to be done. In the present work, a numerical study of flow and temperature distribution in the coolant channel of TRIGA 2000 has been carried out using CFD package. For this study, simulations were carried out on 3-D tested model. The model consists of the reactor tank, thermal and thermalizing column, reflector, rotary specimen rack, chimney, fuel element, primary pipe, diffuser, beam tube and a part of the core are constructed by 1.50 million unstructured tetrahedral cell elements. The results show that for the initial condition (116 fuel elements in the core) and for the inlet temperature of 24°C and the primary velocity of 5.6 m/s, there no boiling phenomena occur in the coolant channel. Due to this result, it is now possible to improve the core cooling system of TRIGA 2000 reactor. Meanwhile, forced flow from the diffuser system only affected the flow pattern in the outside of chimney and put on a small effect to the fluid flow's velocity in the inside of chimney.

  15. Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

    NASA Technical Reports Server (NTRS)

    Bragg-Sitton, Shannon M.

    2005-01-01

    The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

  16. Feasibility study of full-reactor gas core demonstration test

    NASA Technical Reports Server (NTRS)

    Kunze, J. F.; Lofthouse, J. H.; Shaffer, C. J.; Macbeth, P. J.

    1973-01-01

    Separate studies of nuclear criticality, flow patterns, and thermodynamics for the gas core reactor concept have all given positive indications of its feasibility. However, before serious design for a full scale gas core application can be made, feasibility must be shown for operation with full interaction of the nuclear, thermal, and hydraulic effects. A minimum sized, and hence minimum expense, test arrangement is considered for a full gas core configuration. It is shown that the hydrogen coolant scattering effects dominate the nuclear considerations at elevated temperatures. A cavity diameter of somewhat larger than 4 ft (122 cm) will be needed if temperatures high enough to vaporize uranium are to be achieved.

  17. Two stochastic optimization algorithms applied to nuclear reactor core design

    Microsoft Academic Search

    Wagner F. Sacco; Cassiano R. E. de oliveira; Cláudio M. N. A. Pereira

    2006-01-01

    Two stochastic optimization algorithms conceptually similar to Simulated Annealing are presented and applied to a core design optimization problem previously solved with Genetic Algorithms. The two algorithms are the novel Particle Collision Algorithm (PCA), which is introduced in detail, and Dueck's Great Deluge Algorithm (GDA). The optimization problem consists in adjusting several reactor cell parameters, such as dimensions, enrichment and

  18. Structural homogenized analysis for a nuclear reactor core

    Microsoft Academic Search

    R. J. Zhang

    1998-01-01

    A nuclear reactor core is composed of a great number of tubular beams with periodic structure, which are immersed in an acoustic fluid. In the present paper, a 3-D homogenization model is developed to predict its overall dynamic behavior. An approximate solution to the local problem is given. The application to an 1-D example shows that approximate expressions of the

  19. Benchmark Evaluation of the NRAD Reactor LEU Core Startup Measurements

    SciTech Connect

    J. D. Bess; T. L. Maddock; M. A. Marshall

    2011-09-01

    The Neutron Radiography (NRAD) reactor is a 250-kW TRIGA-(Training, Research, Isotope Production, General Atomics)-conversion-type reactor at the Idaho National Laboratory; it is primarily used for neutron radiography analysis of irradiated and unirradiated fuels and materials. The NRAD reactor was converted from HEU to LEU fuel with 60 fuel elements and brought critical on March 31, 2010. This configuration of the NRAD reactor has been evaluated as an acceptable benchmark experiment and is available in the 2011 editions of the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) and the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Significant effort went into precisely characterizing all aspects of the reactor core dimensions and material properties; detailed analyses of reactor parameters minimized experimental uncertainties. The largest contributors to the total benchmark uncertainty were the 234U, 236U, Er, and Hf content in the fuel; the manganese content in the stainless steel cladding; and the unknown level of water saturation in the graphite reflector blocks. A simplified benchmark model of the NRAD reactor was prepared with a keff of 1.0012 {+-} 0.0029 (1s). Monte Carlo calculations with MCNP5 and KENO-VI and various neutron cross section libraries were performed and compared with the benchmark eigenvalue for the 60-fuel-element core configuration; all calculated eigenvalues are between 0.3 and 0.8% greater than the benchmark value. Benchmark evaluations of the NRAD reactor are beneficial in understanding biases and uncertainties affecting criticality safety analyses of storage, handling, or transportation applications with LEU-Er-Zr-H fuel.

  20. Support arrangements for core modules of nuclear reactors. [PWR

    DOEpatents

    Bollinger, L.R.

    1983-11-03

    A support arrangement is provided for the core modules of a nuclear reactor which provides support access through the control drive mechanisms of the reactor. This arrangement provides axial support of individual reactor core modules from the pressure vessel head in a manner which permits attachment and detachment of the modules from the head to be accomplished through the control drive mechanisms after their leadscrews have been removed. The arrangement includes a module support nut which is suspended from the pressure vessel head and screw threaded to the shroud housing for the module. A spline lock prevents loosening of the screw connection. An installation tool assembly, including a cell lifting and preloading tool and a torquing tool, fits through the control drive mechanism and provides lifting of the shroud housing while disconnecting the spline lock, as well as application of torque to the module support nut.

  1. NEUTRON RADIOGRAPHY (NRAD) REACTOR 64-ELEMENT CORE UPGRADE

    SciTech Connect

    John D. Bess

    2014-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA (registered) (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The interim critical configuration developed during the core upgrade, which contains only 62 fuel elements, has been evaluated as an acceptable benchmark experiment. The final 64-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (approximately +/-1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  2. Support arrangement for core modules of nuclear reactors

    DOEpatents

    Bollinger, Lawrence R. (Schenectady, NY)

    1987-01-01

    A support arrangement is provided for the core modules of a nuclear reactor which provides support access through the control drive mechanisms of the reactor. This arrangement provides axial support of individual reactor core modules from the pressure vessel head in a manner which permits attachment and detachment of the modules from the head to be accomplished through the control drive mechanisms after their leadscrews have been removed. The arrangement includes a module support nut which is suspended from the pressure vessel head and screw threaded to the shroud housing for the module. A spline lock prevents loosening of the screw connection. An installation tool assembly, including a cell lifting and preloading tool and a torquing tool, fits through the control drive mechanism and provides lifting of the shroud housing while disconnecting the spline lock, as well as application of torque to the module support nut.

  3. Core Optimization of a Deep-Burn Pebble Bed Reactor

    SciTech Connect

    Brian Boer; Abderrafi M. Ougouag

    2010-06-01

    Achieving a high fuel burnup in the Deep-Burn (DB) pebble bed reactor design, while remaining within the limits for fuel temperature, power peaking and temperature reactivity feedback, is challenging. The high content of Pu and Minor Actinides in the Deep-Burn fuel significantly impacts the thermal neutron energy spectrum as compared to a ’standard’ UO2 fueled core. This can result in power and temperature peaking in the pebble bed core in locally thermalized regions near the graphite reflectors. Furthermore, the interplay of the Pu resonances of the neutron absorption cross sections at low-lying energies can lead to a positive temperature reactivity coefficient for the graphite moderator at certain operating conditions. The DB concept focuses on the destruction of spent fuel transuranics in TRISO coated particle fueled gas-cooled reactors with the aim of a fractional fuel burnup of 60-70% in fissions per initial metal atom (FIMA), using a single-pass, multi in-core fuel (re)cycling scheme. In principle, the DB pebble bed concept employs the same reactor designs as the present low enriched uranium core designs, i.e. the 400 MWth Pebble Bed Modular Reactor (PBMR-400). A Pu and Minor Actinide fueled PBMR-400 design serves as the starting point for a core optimization study. The fuel temperature, power peak, temperature reactivity coefficients, and burnup capabilities of the modified designs are analyzed with the PEBBED code. A code-to-code coupling with the PASTA code allows for the analysis of the TRISO fuel performance for both normal and Loss Of Forced Cooling conditions. An improved core design is sought, maximizing the fuel discharge burnup, while retaining negative temperature reactivity feedback coefficients for the entire temperature range and avoiding high fuel temperatures (fuel failure probabilities).

  4. Development and Assessment of Advanced Reactor Core Protection System

    NASA Astrophysics Data System (ADS)

    in, Wang-Kee; Park, Young-Ho; Baeg, Seung-Yeob

    An advanced core protection system for a pressurized water reactor, Reactor Core Protection System(RCOPS), was developed by adopting a high performance hardware platform and optimal system configuration. The functional algorithms of the core protection system were also improved to enhance the plant availability by reducing unnecessary reactor trips and increasing operational margin. The RCOPS consists of four independent safety channels providing a two-out-of-four trip logic. The reliability analysis using the reliability block diagram method showed the unavailability of the RCOPS to be lower than the conventional system. The failure mode and effects analysis demonstrated that the RCOPS does not lose its intended safety functions for most failures. New algorithms for the RCOPS functional design were implemented in order to avoid unnecessary reactor trips by providing auxiliary pre-trip alarms and signal validation logic for the control rod position. The new algorithms in the RCOPS were verified by comparing the RCOPS calculations with reference results. The new thermal margin algorithm for the RCOPS was expected to increase the operational margin to the limit for Departure from Nucleate Boiling Ratio (DNBR) by approximately 1%.

  5. CRITICAL STUDIES WITH ZPRIII FOR THE ENRICO FERMI FAST REACTOR CORE B

    Microsoft Academic Search

    T. A. Doyle; A. L. Hess

    1962-01-01

    Experiments in the Zero Power Reactor with a mockup of the second core ; loading, Core B, of the Fermi Fast Reactor are described. These experiments were ; conducted to provide basic critical data for the Core B reactor. Measurements ; were made on the reaction rates and fission ratios, and information was obtained ; concerning reactivity effects of various

  6. System Study: Reactor Core Isolation Cooling 1998–2012

    SciTech Connect

    T. E. Wierman

    2013-10-01

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2012 for selected components were obtained from the Equipment Performance and Information Exchange (EPIX). The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing trend was identified in the HPCI results. Statistically significant decreasing trends were identified for RCIC start-only and 8-hour trends.

  7. Gas core reactor concepts and technology - Issues and baseline strategy

    NASA Technical Reports Server (NTRS)

    Diaz, Nils J.; Dugan, Edward T.; Kahook, Samer; Maya, Isaac

    1991-01-01

    Results of a research program including phenomenological studies, conceptual design, and systems analysis of a series of gaseous/vapor fissile fuel driven engines for space power platforms and for thermal and electric propulsion are reviewed. It is noted that gas and vapor phase reactors provide the path for minimum mass in orbit and trip times, with a specific impulse from 1020 sec at the lowest technololgical risk to 5200 sec at the highest technological risk. The discussion covers various configurations of gas core reactors and critical technologies and the nuclear vapor thermal rocket engine.

  8. Hyper-heuristic applied to nuclear reactor core design

    NASA Astrophysics Data System (ADS)

    Domingos, R. P.; Platt, G. M.

    2013-02-01

    The design of nuclear reactors gives rises to a series of optimization problems because of the need for high efficiency, availability and maintenance of security levels. Gradient-based techniques and linear programming have been applied, as well as genetic algorithms and particle swarm optimization. The nonlinearity, multimodality and lack of knowledge about the problem domain makes de choice of suitable meta-heuristic models particularly challenging. In this work we solve the optimization problem of a nuclear reactor core design through the application of an optimal sequence of meta-heuritics created automatically. This combinatorial optimization model is known as hyper-heuristic.

  9. Post impact behavior of mobile reactor core containment systems

    NASA Technical Reports Server (NTRS)

    Puthoff, R. L.; Parker, W. G.; Vanbibber, L. E.

    1972-01-01

    The reactor core containment vessel temperatures after impact, and the design variables that affect the post impact survival of the system are analyzed. The heat transfer analysis includes conduction, radiation, and convection in addition to the core material heats of fusion and vaporization under partially burial conditions. Also, included is the fact that fission products vaporize and transport radially outward and condense outward and condense on cooler surfaces, resulting in a moving heat source. A computer program entitled Executive Subroutines for Afterheat Temperature Analysis (ESATA) was written to consider this complex heat transfer analysis. Seven cases were calculated of a reactor power system capable of delivering up to 300 MW of thermal power to a nuclear airplane.

  10. Fuel Breeding and Core Behavior Analyses on In Core Fuel Management of Water Cooled Thorium Reactors

    NASA Astrophysics Data System (ADS)

    Permana, Sidik; Sekimoto, Hiroshi; Waris, Abdul; Subhki, Muhamad Nurul; Ismail

    2010-12-01

    Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period has been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this evaluation has confirmed that breeding condition and negative coefficient can be obtained simultaneously for water-cooled thorium reactor obtains based on the whole core fuel arrangement.

  11. Fuel Breeding and Core Behavior Analyses on In Core Fuel Management of Water Cooled Thorium Reactors

    SciTech Connect

    Permana, Sidik [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, 2-12-1-N1-17, O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132 (Indonesia); Sekimoto, Hiroshi [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, 2-12-1-N1-17, O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Waris, Abdul; Subhki, Muhamad Nurul [Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132 (Indonesia); Ismail, [BAPETEN (Indonesia)

    2010-12-23

    Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period has been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this evaluation has confirmed that breeding condition and negative coefficient can be obtained simultaneously for water-cooled thorium reactor obtains based on the whole core fuel arrangement.

  12. 100KE REACTOR CORE REMOVAL PROJECT ALTERNATIVE ANALYSIS WORKSHOP REPORT

    Microsoft Academic Search

    HARRINGTON RA

    2010-01-01

    On December 15-16, 2009, a 100-KE Reactor Core Removal Project Alternative Analysis Workshop was conducted at the Washington State University Consolidated Information Center, Room 214. Colburn Kennedy, Project Director, CH2M HILL Plateau Remediation Company (CHPRC) requested the workshop and Richard Harrington provided facilitation. The purpose of the session was to select the preferred Bio Shield Alternative, for integration with the

  13. Sequential Reactions Directed by Core/Shell Catalytic Reactors

    SciTech Connect

    Wei, Yanhu [Northwestern Univ., Evanston, IL (United States); Soh, Siowling [Northwestern Univ., Evanston, IL (United States); Apodaca, Mario M. [Northwestern Univ., Evanston, IL (United States); Kim, Jiwon [Northwestern Univ., Evanston, IL (United States); Grzybowski, Bartosz A. [Northwestern Univ., Evanston, IL (United States)

    2010-01-01

    Millimeter-sized reactor particles made of permeable polymer doped with catalysts arranged in a core/shell fashion direct sequences of chemical reactions (e.g., alkyne coupling followed by hydrogenation or hydrosilylation followed by hydrogenation). Spatial compartmentalization of catalysts coupled with the diffusion of substrates controls reaction order and avoids formation of byproducts. The experimentally observed yields of reaction sequences are reproduced by a theoretical model, which accounts for the reaction kinetics and the diffusion of the species involved.

  14. Role of Minor Actinides for Long-Life Reactor Cores

    SciTech Connect

    Saito, M.; Artisyuk, V. [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, 2-12-1 O-okayama, Meguro-ku, Tokyo 152 (Japan); Shmelev, A. [Moscow Engineering and Physics Institute, 31, Kashirskoe Shosse, Moscow (Russian Federation); Nikitin, K.; Peryoga, Y

    2002-07-01

    The paper addresses the study on advanced fuel cycles for LWR oriented to high burnup values that exceed 100 GWd/tHM, thus giving the chance to establish the long-life reactor cores without fuel reloading on site. The key element of this approach is a broad involvement of Minor Actinides whose admixture to 20% enriched uranium fuel provides safe release of initial reactivity excess and improved proliferation resistance properties. (authors)

  15. Piezoelectric material for use in a nuclear reactor core

    NASA Astrophysics Data System (ADS)

    Parks, D. A.; Reinhardt, Brian; Tittmann, B. R.

    2012-05-01

    In radiation environments ultrasonic nondestructive evaluation has great potential for improving reactor safety and furthering the understanding of radiation effects and materials. In both nuclear power plants and materials test reactors, elevated temperatures and high levels of radiation present challenges to ultrasonic NDE methodologies. The challenges are primarily due to the degradation of the ultrasonic sensors utilized. We present results from the operation of a ultrasonic piezoelectric transducer, composed of bulk single crystal AlN, in a nuclear reactor core for over 120 MWHrs. The transducer was coupled to an aluminum cylinder and operated in pulse echo mode throughout the irradiation. In addition to the pulse echo testing impedance data were obtained. Further, the piezoelectric coefficient d33 was measured prior to irradiation and found to be 5.5 pC/N which is unchanged from as-grown samples, and in fact higher than the measured d33 for many as-grown samples.

  16. Piezoelectric material for use in a nuclear reactor core

    SciTech Connect

    Parks, D. A.; Reinhardt, Brian; Tittmann, B. R. [EES Department, Penn State University, University Park, PA 16802 (United States)

    2012-05-17

    In radiation environments ultrasonic nondestructive evaluation has great potential for improving reactor safety and furthering the understanding of radiation effects and materials. In both nuclear power plants and materials test reactors, elevated temperatures and high levels of radiation present challenges to ultrasonic NDE methodologies. The challenges are primarily due to the degradation of the ultrasonic sensors utilized. We present results from the operation of a ultrasonic piezoelectric transducer, composed of bulk single crystal AlN, in a nuclear reactor core for over 120 MWHrs. The transducer was coupled to an aluminum cylinder and operated in pulse echo mode throughout the irradiation. In addition to the pulse echo testing impedance data were obtained. Further, the piezoelectric coefficient d{sub 33} was measured prior to irradiation and found to be 5.5 pC/N which is unchanged from as-grown samples, and in fact higher than the measured d{sub 33} for many as-grown samples.

  17. RMC - A Monte Carlo Code for Reactor Core Analysis

    NASA Astrophysics Data System (ADS)

    Wang, Kan; Li, Zeguang; She, Ding; Liang, Jin'gang; Xu, Qi; Qiu, Yishu; Yu, Jiankai; Sun, Jialong; Fan, Xiao; Yu, Ganglin

    2014-06-01

    A new Monte Carlo transport code RMC has been being developed by Department of Engineering Physics, Tsinghua University, Beijing as a tool for reactor core 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 effciency. 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.

  18. Coupled simulation of the reactor core using CUPID/MASTER

    SciTech Connect

    Lee, J. R.; Cho, H. K.; Yoon, H. Y.; Jeong, J. J. [Korea Atomic Energy Research Institue, Daedeok-daero 989-111, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

    2012-07-01

    The CUPID is a component-scale thermal hydraulics code which is aimed for the analysis of transient two-phase flows in nuclear reactor components such as the reactor vessel, steam generator, containment. This code adopts a three-dimensional, transient, two-phase and three-field model, and includes physical models and correlations of the interfacial mass, momentum, and energy transfer for the closure. In the present paper, a multi-physics simulation was performed by coupling CUPID with a three dimensional neutron kinetics code, MASTER. MASTER is merged into CUPID as a dynamic link library (DLL). The APR1400 reactor core during a control rod drop/ejection accident was simulated as an example by adopting a porous media approach to employ a fuel assembly. The following sections present the numerical modeling for the reactor core, coupling of the kinetics code, and the simulation results. And also, a preliminary study for multi-scale simulation between CUPID and system-scaled thermal hydraulics code, MARS will be introduced as well. (authors)

  19. MCNP/MCNPX model of the annular core research reactor.

    SciTech Connect

    DePriest, Kendall Russell; Cooper, Philip J.; Parma, Edward J., Jr. (.,; .)

    2006-10-01

    Many experimenters at the Annular Core Research Reactor (ACRR) have a need to predict the neutron/gamma environment prior to testing. In some cases, the neutron/gamma environment is needed to understand the test results after the completion of an experiment. In an effort to satisfy the needs of experimenters, a model of the ACRR was developed for use with the Monte Carlo N-Particle transport codes MCNP [Br03] and MCNPX [Wa02]. The model contains adjustable safety, transient, and control rods, several of the available spectrum-modifying cavity inserts, and placeholders for experiment packages. The ACRR model was constructed such that experiment package models can be easily placed in the reactor after being developed as stand-alone units. An addition to the 'standard' model allows the FREC-II cavity to be included in the calculations. This report presents the MCNP/MCNPX model of the ACRR. Comparisons are made between the model and the reactor for various configurations. Reactivity worth curves for the various reactor configurations are presented. Examples of reactivity worth calculations for a few experiment packages are presented along with the measured reactivity worth from the reactor test of the experiment packages. Finally, calculated neutron/gamma spectra are presented.

  20. Reactor dynamics and stability analysis for two gaseous core reactor space power systems

    NASA Astrophysics Data System (ADS)

    Dugan, Edward T.; Kahook, Samer D.

    1992-01-01

    Reactor dynamics and system stability studies are performed for two conceptual gaseous core reactor space nuclear power systems. The analysis is conducted using non-linear models which include circulating fuel, point reactor kinetics equations and appropriate thermodynamic, heat transfer and one-dimensional isentropic flow equations. The studies reveal the existence of some unique and very effective inherent reactivity feedback effects such as the vapor fuel density power coefficient that are capable of stabilizing these systems safely and quickly, within a few seconds, even when large positive reactivity insertions are imposed. However, due to the strength of these feedbacks, it is found that external reactivity insertions alone are inadequate for bringing about significant power level changes during normal operations. Additional methods of reactivity control such as changes in the gaseous fuel mass flow rate, or gaseous fuel core inlet pressure are needed to achieve the desired power level control.

  1. Development of Mold-Core Type Gapped Iron-Core Reactor using Adhesive Coated Electromagnetic Sheet

    NASA Astrophysics Data System (ADS)

    Kuwata, Minoru; Nogawa, Shuichi; Takahashi, Norio; Miyagi, Daisuke; Takeda, Kazutoshi

    The adhesive coated non-oriented electromagnetic steel sheet is well known to be effective in realizing high efficiency, high power motors with compact size and low noise. The authors developed newly adhesive coated grain oriented electromagnetic steel sheets. Then, a gapped iron-core type reactor having a new iron-core structure is developed using the adhesive coated steel sheets. By adopting the adhesive coated grain oriented electromagnetic steel sheets to the leg of the reactor, the fastening studs of the laminated electromagnetic steel sheets, which are required in conventional reactor, and the through hole for the fastening studs could be omitted. This enabled us to simplify the structure and reduce the core diameter. On the other hand, we examined the magnetic flux distributions and local loss distributions in the yoke by the detailed magnetic field analysis to utilize the grain oriented electromagnetic steel sheet with bolt-less construction, and realized the reduction of dimensions and weights due to the increase of magnetic flux density. The reactor developed has such features of smaller size, lighter weight and lower noise level.

  2. Seismic responses of a pool-type fast reactor with different core support designs

    SciTech Connect

    Wu, Ting-shu; Seidensticker, R.W. (Argonne National Lab., IL (USA))

    1989-01-01

    In designing the core support system for a pool-type fast reactor, there are many issues which must be considered in order to achieve an optimum and balanced design. These issues include safety, reliability, as well as costs. Several design options are possible to support the reactor core. Different core support options yield different frequency ranges and responses. Seismic responses of a large pool-type fast reactor incorporated with different core support designs have been investigated. 4 refs., 3 figs.

  3. Conceptual Design of a Modular Island Core Fast Breeder Reactor “RAPID-M”

    Microsoft Academic Search

    Mitsuru KAMBE

    2002-01-01

    A metal fueled modular island core sodium cooled fast breeder reactor concept RAPID-M to improve reactor performance and proliferation resistance and to accommodate various power requirements has been demonstrated. The essential feature of the RAPID-M concept is that the reactor core consists of integrated fuel assemblies (IFAs) instead of conventional fuel subassemblies. The RAPID concept enables quick and simplified refueling

  4. Piccolo Micromegas: first in-core measurements in a nuclear reactor

    E-print Network

    Paris-Sud XI, Université de

    Piccolo Micromegas: first in-core measurements in a nuclear reactor J. Pancina , S. Andriamonjea in the coupling of an accelerator with a nuclear reactor. Such systems will need neutron detectors working domains. For the first time, Piccolo Micromegas has been placed in the core of a nuclear reactor

  5. Corium retention for high power reactors by an in-vessel core catcher in combination with External Reactor Vessel Cooling

    Microsoft Academic Search

    J. L. Rempe; D. L. Knudson; K. G. Condie; K. Y. Suh; F.-B. Cheung; S.-B. Kim

    2004-01-01

    If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel lower head remains intact so that relocated core materials are retained within

  6. Heat transfer evaluation in a plasma core reactor

    NASA Technical Reports Server (NTRS)

    Smith, D. E.; Smith, T. M.; Stoenescu, M. L.

    1976-01-01

    Numerical evaluations of heat transfer in a fissioning uranium plasma core reactor cavity, operating with seeded hydrogen propellant, was performed. A two-dimensional analysis is based on an assumed flow pattern and cavity wall heat exchange rate. Various iterative schemes were required by the nature of the radiative field and by the solid seed vaporization. Approximate formulations of the radiative heat flux are generally used, due to the complexity of the solution of a rigorously formulated problem. The present work analyzes the sensitivity of the results with respect to approximations of the radiative field, geometry, seed vaporization coefficients and flow pattern. The results present temperature, heat flux, density and optical depth distributions in the reactor cavity, acceptable simplifying assumptions, and iterative schemes. The present calculations, performed in cartesian and spherical coordinates, are applicable to any most general heat transfer problem.

  7. VIPRE-01: A thermal-hydraulic code for reactor cores:

    SciTech Connect

    Stewart, C.W.; Cuta, J.M.

    1988-03-01

    VIPRE (Versatile Internals and Component Program for Reactors;EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear reactor core safety limits including minimum departure from nucleate boiling ratio (NDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume discusses general and specific considerations in using VIPRE as a thermal-hydraulic analysis tool. Volume 1: Mathematical Modeling, explains the major thermal-hydraulic models and supporting mathematial correlations in detail. Volume 2: Users's Manual, describes the input requirements of the codes in the VIPRE code package. Volume 3: Programmer's Manual, explains the code structure and computer interface. Experimence in running VIPRE is documented in Volume 4: Applications. 25 refs., 31 figs., 7 tabs.

  8. Construction of linear empirical core models for pressurized water reactor in-core fuel management

    SciTech Connect

    Okafor, K.C.; Aldemir, T. (The Ohio State Univ., Dept. of Mechanical Engineering, Nuclear Engineering Program, 206 West 18th Ave., Columbus, OH (US))

    1988-06-01

    An empirical core model construction procedure for pressurized water reactor (PWR) in-core fuel management problems is presented that (a) incorporates the effect of composition changes in all the control zones in the core of a given fuel assembly, (b) is valid at all times during the cycle for a given range of control variables, (c) allows determining the optimal beginning of cycle (BOC) kappainfinity distribution as a single linear programming problem,and (d) provides flexibility in the choice of the material zones to describe core composition. Although the modeling procedure assumes zero BOC burnup, the predicted optimal kappainfinity profiles are also applicable to reload cores. In model construction, assembly power fractions and burnup increments during the cycle are regarded as the state (i.e., dependent) variables. Zone enrichments are the control (i.e., independent) variables. The model construction procedure is validated and implemented for the initial core of a PWR to determine the optimal BOC kappainfinity profiles for two three-zone scatter loading schemes. The predicted BOC kappainfinity profiles agree with the results of other investigators obtained by different modeling techniques.

  9. Development of a three-dimensional core dynamics analysis program for commercial boiling water reactors

    Microsoft Academic Search

    Yasunori Bessho; Osamu Yokomizo; Yuichiro Yoshimoto; Ryutaro Yamashita; Masumi Ishikawa; Akio Toba

    1997-01-01

    Development and qualification results are described for a three-dimensional, time-domain core dynamics analysis program for commercial boiling water reactors (BWRs). The program allows analysis of the reactor core with a detailed mesh division, which eliminates calculational ambiguity in the nuclear-thermal-hydraulic stability analysis caused by reactor core regional division. During development, emphasis was placed on high calculational speed and large memory

  10. Tendencies of high temperature gas core reactors for NTP and space power plants development

    SciTech Connect

    Glinik, R.A.

    1993-06-01

    This paper examines the development tendencies of the high-temperature gas phase fuel elements (GPFEs) and gas core nuclear reactors (GCRs). Particular attention is given to the developent program for the gaseous fuel elements in an experimental pulse graphite reactor (IGR) with a thermal neutron flux density up to 10 exp 15 t.n./sq cm s. Diagrams of the reactor, the liquid metal fed system, and the combined gas core reactor are presented. 4 refs.

  11. RECENT DEVELOPMENTS OF THE IRIS PROJECT OF INTEREST FOR LATIN AMERICA

    SciTech Connect

    Carelli, M.D.; Petrovic, B.

    2004-10-03

    The IRIS (International Reactor Innovative and Secure) reactor design is being developed by an international consortium of 21 organizations from ten countries, including three members from Brazil and one from Mexico. This reflects the interest that Latin America has for a project which addresses the energy needs of the region. Presented here are some of the most recent developments in the IRIS project. The project's highest priority is the current pre-application licensing with the US NRC, which has required an investigation of the major accident sequences and a preliminary probabilistic risk assessment (PRA). The results of the accident analyses confirmed the outstanding inherent safety of the IRIS configuration and the PRA analyses indicated a core damage frequency due to internal events of the order of 2E-8. This not only highlights the enhanced safety characteristic of IRIS which should enhance its public acceptance, but it has also prompted IRIS to consider the possibility of being licensed without the need for off-site emergency response planning which would have a very positive economic implication. The modular IRIS, with each module rated at {approx} 335 MWe, is of course an ideal size for developing countries as it allows to easily introduce a moderate amount of power on limited electric grids. IRIS can be deployed in single modules in regions only requiring a few hundred MWs or in multiple modules deployed successively at time intervals in large urban areas requiring a larger amount of power increasing with time. IRIS is designed to operate ''hands-off'' as much as possible, with a small crew, having in mind deployment in areas with limited infrastructure. Thus IRIS has a 48-months maintenance interval, long refueling cycles in excess of three years, and is designed to increase as much as possible operational reliability. For example, the project has recently adopted internal control rod drive mechanisms to eliminate vessel head penetrations and the possibility of corrosion cracking as in Davis-Besse and other plants. Latin America, as many other regions on the earth, needs water as much as electricity. IRIS has developed a water desalination co-generation design which can employ a variety of processes as dictated by local and economic conditions. Applications to the arid Brazilian Nord-Este and Mexican Nord-Oeste are being considered.

  12. PRIZMA predictions of in-core detection indications in the VVER-1000 reactor

    NASA Astrophysics Data System (ADS)

    Kandiev, Yadgar Z.; Kashayeva, Elena A.; Malyshin, Gennady N.; Modestov, Dmitry G.; Khatuntsev, Kirill E.

    2014-06-01

    The paper describes calculations which were done by the PRIZMA code(1) to predict indications of in-core rhodium detectors in the VVER-1000 reactor for some core fragments with allowance for fuel and rhodium burnout.

  13. BENCHMARK EVALUATION OF THE START-UP CORE REACTOR PHYSICS MEASUREMENTS OF THE HIGH TEMPERATURE ENGINEERING TEST REACTOR

    SciTech Connect

    John Darrell Bess

    2010-05-01

    The benchmark evaluation of the start-up core reactor physics measurements performed with Japan’s High Temperature Engineering Test Reactor, in support of the Next Generation Nuclear Plant Project and Very High Temperature Reactor Program activities at the Idaho National Laboratory, has been completed. The evaluation was performed using MCNP5 with ENDF/B-VII.0 nuclear data libraries and according to guidelines provided for inclusion in the International Reactor Physics Experiment Evaluation Project Handbook. Results provided include updated evaluation of the initial six critical core configurations (five annular and one fully-loaded). The calculated keff eigenvalues agree within 1s of the benchmark values. Reactor physics measurements that were evaluated include reactivity effects measurements such as excess reactivity during the core loading process and shutdown margins for the fully-loaded core, four isothermal temperature reactivity coefficient measurements for the fully-loaded core, and axial reaction rate measurements in the instrumentation columns of three core configurations. The calculated values agree well with the benchmark experiment measurements. Fully subcritical and warm critical configurations of the fully-loaded core were also assessed. The calculated keff eigenvalues for these two configurations also agree within 1s of the benchmark values. The reactor physics measurement data can be used in the validation and design development of future High Temperature Gas-cooled Reactor systems.

  14. System startup simulation for an in-core thermionic reactor with heat pipe cooling

    NASA Astrophysics Data System (ADS)

    Determan, William R.; Otting, William D.

    1992-01-01

    The heat pipe cooled thermionic (HPTI) reactor relies on in-core sodium heat pipes to provide a redundant means of cooling the 72 thermionic fuel elements (TFEs) which comprise the 40-kWe reactor core assembly. In-core heat pipe cooling was selected for the reactor design due to a requirement for multiple system on-orbit restarts over its lifetime. Powering up the reactor requires the in-core and radiator heat pipes to undergo a thaw cycle with a rapid ascension in power to their operating temperatures. The present study considers how fast the thaw-out and power ascension cycle can be safely accomplished within a reactor core. As part of the study, a transient startup simulator model of the heat pipe cooled reactor system was developed. Results of the startup transient simulation are provided.

  15. IRIS Final Technical Progress Report

    SciTech Connect

    M. D. Carelli

    2003-11-03

    OAK-B135 This NERI project, originally started as the Secure Transportable Autonomous Light Water Reactor (STAR-LW) and currently known as the International Reactor Innovative and Secure (IRIS) project, had the objective of investigating a novel type of water-cooled reactor to satisfy the Generation IV goals: fuel cycle sustainability, enhanced reliability and safety, and improved economics. The research objectives over the three-year (1999-2002) program were as follows: First year: Assess various design alternatives and establish main characteristics of a point design; Second year: Perform feasibility and engineering assessment of the selected design solutions; Third year: Complete reactor design and performance evaluation, including cost assessment These objectives were fully attained and actually they served to launch IRIS as a full fledged project for eventual commercial deployment. The program did not terminate in 2002 at the end of the NERI program, and has just entered in its fifth year. This has been made possible by the IRIS project participants which have grown from the original four member, two-countries team to the current twenty members, nine countries consortium. All the consortium members work under their own funding and it is estimated that the value of their in-kind contributions over the life of the project has been of the order of $30M. Currently, approximately 100 people worldwide are involved in the project. A very important constituency of the IRIS project is the academia: 7 universities from four countries are members of the consortium and five more US universities are associated via parallel NERI programs. To date, 97 students have worked or are working on IRIS; 59 IRIS-related graduate theses have been prepared or are in preparation, and 41 of these students have already graduated with M.S. (33) or Ph.D. (8) degrees. This ''final'' report (final only as far as the NERI program is concerned) summarizes the work performed in the first four years of IRIS, from October 1999 to October 2003. It provides a panoramic of the project status and design effort, with emphasis on the current status, since two previous reports have very extensively documented the work performed, from inception to early 2002.

  16. Plasma core reactor simulations using RF uranium seeded argon discharges

    NASA Technical Reports Server (NTRS)

    Roman, W. C.

    1975-01-01

    An experimental investigation was conducted using the United Technologies Research Center (UTRC) 80 kW and 1.2 MW RF induction heater systems to aid in developing the technology necessary for designing a self-critical fissioning uranium plasma core reactor (PCR). A nonfissioning, steady-state RF-heated argon plasma seeded with pure uranium hexafluoride (UF6) was used. An overall objective was to achieve maximum confinement of uranium vapor within the plasma while simultaneously minimizing the uranium compound wall deposition. Exploratory tests were conducted using the 80 kW RF induction heater with the test chamber at approximately atmospheric pressure and discharge power levels on the order of 10 kW. Four different test chamber flow configurations were tested to permit selection of the configuration offering the best confinement characteristics for subsequent tests at higher pressure and power in the 1.2 MW RF induction heater facility.

  17. Plasma core reactor simulations using RF uranium seeded argon discharges

    NASA Technical Reports Server (NTRS)

    Roman, W. C.

    1976-01-01

    Experimental results are described in which pure uranium hexafluoride was injected into an argon-confined, steady-state, RF-heated plasma to investigate characteristics of plasma core nuclear reactors. The 80 kW (13.56 MHz) and 1.2 MW (5.51 MHz) rf induction heater facilities were used to determine a test chamber flow scheme which offered best uranium confinement with minimum wall coating. The cylindrical fused-silica test chamber walls were 5.7-cm-ID by 10-cm-long. Test conditions included RF powers of 2-85 kW, chamber pressures of 1-12 atm, and uranium hexafluoride mass-flow rates of 0.005-0.13 g/s. Successful techniques were developed for fluid-mechanical confinement of RF-heated plasmas with pure uranium hexafluoride injection.

  18. Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

    DOEpatents

    Peterson, Per F.

    2013-05-14

    A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

  19. Liquid level, void fraction, and superheated steam sensor for nuclear-reactor cores. [PWR; BWR

    DOEpatents

    Tokarz, R.D.

    1981-10-27

    This disclosure relates to an apparatus for monitoring the presence of coolant in liquid or mixed liquid and vapor, and superheated gaseous phases at one or more locations within an operating nuclear reactor core, such as pressurized water reactor or a boiling water reactor.

  20. A NOVEL MICROMEGAS DETECTOR FOR IN-CORE NUCLEAR REACTOR NEUTRON FLUX MEASUREMENTS

    E-print Network

    Paris-Sud XI, Université de

    1 A NOVEL MICROMEGAS DETECTOR FOR IN-CORE NUCLEAR REACTOR NEUTRON FLUX MEASUREMENTS S. ANDRIAMONJE Talence Cedex, France Future fast nuclear reactors designed for energy production and transmutation to neutron detection inside nuclear reactor is given. The advantage of this detector over conventional

  1. Development concept for a small, split-core, heat-pipe-cooled nuclear reactor

    NASA Technical Reports Server (NTRS)

    Lantz, E.; Breitwieser, R.; Niederauer, G. F.

    1974-01-01

    There have been two main deterrents to the development of semiportable nuclear reactors. One is the high development costs; the other is the inability to satisfy with assurance the questions of operational safety. This report shows how a split-core, heat-pipe cooled reactor could conceptually eliminate these deterrents, and examines and summarizes recent work on split-core, heat-pipe reactors. A concept for a small reactor that could be developed at a comparatively low cost is presented. The concept would extend the technology of subcritical radioisotope thermoelectric generators using 238 PuO2 to the evolution of critical space power reactors using 239 PuO2.

  2. Ultrahigh temperature vapor core reactor-MHD system for space nuclear electric power

    NASA Technical Reports Server (NTRS)

    Maya, Isaac; Anghaie, Samim; Diaz, Nils J.; Dugan, Edward T.

    1991-01-01

    The conceptual design of a nuclear space power system based on the ultrahigh temperature vapor core reactor with MHD energy conversion is presented. This UF4 fueled gas core cavity reactor operates at 4000 K maximum core temperature and 40 atm. Materials experiments, conducted with UF4 up to 2200 K, demonstrate acceptable compatibility with tungsten-molybdenum-, and carbon-based materials. The supporting nuclear, heat transfer, fluid flow and MHD analysis, and fissioning plasma physics experiments are also discussed.

  3. American Iris Society

    NSDL National Science Digital Library

    The American Iris Society (AIS) was founded in 1920, "and exists for the sole purpose of promoting the culture and improvement of the Iris." This official AIS website serves as an information resource for iris aficionados and AIS members. The site contains information about AIS awards, membership, upcoming conventions, and the annual Symposium--a "popularity poll of Tall Bearded Iris conducted by the AIS." In addition, the site has sections regarding Iris Registration, Iris Classification, online iris email groups, and related links. Of course the site also contains a small photo gallery featuring beautiful images of award-winning irises, and a brief article on growing and planting irises. The AIS is divided into 24 regions across the U.S. and Canada with local iris organizations in each region. Site visitors will find contact information for numerous AIS regional organizations, and for the AIS region vice presidents.

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

    SciTech Connect

    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)

    2013-07-01

    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)

  5. Evaluation Method for Core Thermohydraulics during Natural Circulation in Fast Reactors

    NASA Astrophysics Data System (ADS)

    Kamide, Hideki; Nagasawa, Kazuyoshi; Kimura, Nobuyuki; Miyakoshi, Hiroyuki

    Decay heat removal using natural circulation is one of significant functions for a reactor. As the decay heat removal system, a direct reactor auxiliary cooling system has been selected in current designs of fast reactors. In this system, cold sodium is provided in an upper plenum of reactor vessel and it covers the reactor core outlet. The cold sodium can penetrate into the gap region between the subassemblies. This gap flow is referred as inter-wrapper flow (IWF). A numerical estimation method for such phenomena was developed, which modeled each subassembly as a rectangular duct with gap region and also the upper plenum. This numerical simulation method was verified by a sodium test and also a water test. We applied this method to the natural circulation in a 600 MWe class fast reactor. The temperature in the core strongly depended on IWF, flow redistribution in the core, and inter-subassembly heat transfer.

  6. Corium Retention for High Power Reactors by An In-Vessel Core Catcher in Combination with External Reactor Vessel Cooling

    SciTech Connect

    Joy L. Rempe; D. L. Knudson; K. G. Condie; K. Y. Suh; F. -B. Cheung; S. -B. Kim

    2004-05-01

    If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel lower head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe Pressurized Water Reactor (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for in-vessel retention (IVR), resulted in the United States Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing Light Water Reactors (LWRs). Accordingly, IVR of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors. However, it is not clear that currently-proposed methods to achieve ERVC will provide sufficient heat removal for higher power reactors. A US–Korean International Nuclear Energy Research Initiative (INERI) project has been initiated in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) will determine if IVR is feasible for reactors up to 1500 MWe. This paper summarizes results from the first year of this 3-year project.

  7. Turbulence coefficients and stability studies for the coaxial flow or dissimiliar fluids. [gaseous core nuclear reactors

    NASA Technical Reports Server (NTRS)

    Weinstein, H.; Lavan, Z.

    1975-01-01

    Analytical investigations of fluid dynamics problems of relevance to the gaseous core nuclear reactor program are presented. The vortex type flow which appears in the nuclear light bulb concept is analyzed along with the fluid flow in the fuel inlet region for the coaxial flow gaseous core nuclear reactor concept. The development of numerical methods for the solution of the Navier-Stokes equations for appropriate geometries is extended to the case of rotating flows and almost completes the gas core program requirements in this area. The investigations demonstrate that the conceptual design of the coaxial flow reactor needs further development.

  8. McCARD for Neutronics Design and Analysis of Research Reactor Cores

    NASA Astrophysics Data System (ADS)

    Shim, Hyung Jin; Park, Ho Jin; Kwon, Soonwoo; Seo, Geon Ho; Hyo Kim, Chang

    2014-06-01

    McCARD is a Monte Carlo (MC) neutron-photon transport simulation code developed exclusively for the neutronics design and analysis of nuclear reactor cores. McCARD is equipped with the hierarchical modeling and scripting functions, the CAD-based geometry processing module, the adjoint-weighted kinetics parameter and source multiplication factor estimation modules as well as the burnup analysis capability for the neutronics design and analysis of both research and power reactor cores. This paper highlights applicability of McCARD for the research reactor core neutronics analysis, as demonstrated for Kyoto University Critical Assembly, HANARO, and YALINA.

  9. FORMOSA-B: A Boiling Water Reactor In-Core Fuel Management Optimization Package II

    Microsoft Academic Search

    Atul A. Karve; Paul J. Turinsky

    2000-01-01

    As part of the continuing development of the boiling water reactor in-core fuel management optimization code FORMOSA-B, the fidelity of the core simulator has been improved and a control rod pattern (CRP) sampling capability has been added. The robustness of the core simulator is first demonstrated by benchmarking against core load-follow depletion predictions of both SIMULATE-3 and MICROBURN-B2 codes. The

  10. Design study for an advanced liquid-metal fast breeder reactor core with a high burnup

    Microsoft Academic Search

    T. Inagaki; H. Kuga; M. Suzuki; T. Yokoyama; M. Yamaoka; K. Kaneto; M. Ohashi; K. Kurihara

    1989-01-01

    Design studies are performed for a commercial liquid-metal fast breeder reactor core that can achieve a burnup of 200 GWd\\/t. A plutonium-type asymmetric parfait core with two different plutonium-enriched zones in the axial direction as well as in the radial direction is studied. This core concept solves core design problems related to high burnup, and it is possible to achieve

  11. Termination of light-water reactor core-melt accidents with a chemical core catcher: the core-melt source reduction system (COMSORS)

    SciTech Connect

    Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W. [Oak Ridge National Lab., TN (United States); Kenton, M.A. [Dames and Moore, Westmont, IL (United States)

    1996-09-01

    The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate light-water reactor core melt accidents and ensure containment integrity. A special dissolution glass is placed under the reactor vessel. If core debris is released onto the glass, the glass melts and the debris dissolves into the molten glass, thus creating a homogeneous molten glass. The molten glass, with dissolved core debris, spreads into a wide pool, distributing the heat for removal by radiation to the reactor cavity above or by transfer to water on top of the molten glass. Expected equilibrium glass temperatures are approximately 600 degrees C. The creation of a low-temperature, homogeneous molten glass with known geometry permits cooling of the glass without threatening containment integrity. This report describes the technology, initial experiments to measure key glass properties, and modeling of COMSORS operations.

  12. Analysis of Sodium Fire in the Containment Building of Prototype Fast Breeder Reactor Under the Scenario of Core Disruptive Accident

    Microsoft Academic Search

    P. M. Rao; N. Kasinathan; S. E. Kannan

    2006-01-01

    The potential for sodium release to reactor containment building from reactor assembly during Core Disruptive Accident (CDA) in Fast Breeder Reactors (FBR) is an important safety issue with reference to the structural integrity of Reactor Containment Building (RCB). For Prototype Fast Breeder Reactor (PFBR), the estimated sodium release under a CDA of 100 MJ energy release is 350 kg. The

  13. Examination of offsite radiological emergency protective measures for nuclear reactor accidents involving core melt

    E-print Network

    Aldrich, David C.

    1979-01-01

    Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted ...

  14. Fluence-limited burnup as a function of fast reactor core parameters

    E-print Network

    Kersting, Alyssa (Alyssa Rae)

    2011-01-01

    The limiting factor in current designs for fast reactors is not only the reactivity, but also the maximum permissible fast-neutron fluence in the cladding, especially for reduced uranium enrichment cores using high-albedo ...

  15. Examination of offsite radiological emergency measures for nuclear reactor accidents involving core melt. [PWR

    Microsoft Academic Search

    D. C. Aldrich; P. E. McGrath; N. C. Rasmussen

    1978-01-01

    Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, ''Melt-through'' and

  16. FREADM1; fast reactor core accident analysis. [GE635; FORTRAN IV

    Microsoft Academic Search

    T. B. Fowler; M. L. Tobias; J. N. Fox; B. E. Lawler; J. U. Koppel; J. R. Triplett; L. L. Lynn; L. A. Waldman; I. Goldberg; P. Greebler; M. D. Kelley; R. A. Davis; C. E. Keck; J. A. Redfield; W. G. Meinhardt

    2008-01-01

    FREADM1 is a fast reactor, multichannel, accident analysis program designed to efficiently simulate a reactor transient from initiation to the point of core disassembly. Models are included for nuclear kinetics (point model), core thermo-hydraulics, voiding, fuel redistribution, failure propagation, programmed reactivity insertion, and the dynamics of primary-system coolant flow. A broad range of assumed accident initiating and propagating activities may

  17. Identification of a nuclear reactor core (VVER) using recurrent neural networks

    Microsoft Academic Search

    Mehrdad Boroushaki; Mohammad B. Ghofrani; Caro Lucas

    2002-01-01

    Recurrent neural networks (RNNs) in identification of complex nonlinear plants like nuclear reactor core, have difficulty in learning long-term dynamics. Therefore, in most papers in this area, the reactor core is used to identify just the short-term dynamics. In this paper we used a multi-NARX (nonlinear autoregressive with exogenous inputs) structure, including neural networks with different time steps and a

  18. A complete fuel development facility utilizing a dual core TRIGA reactor system

    Microsoft Academic Search

    A. Middleton; G. C. Law

    1974-01-01

    A TRIGA Dual Core Reactor System has been chosen by the Romanian Government as the heart of a new fuel development facility which will be operated by the Romanian Institute for Nuclear Technologies. The Facility, which will be operational in 1976, is an integral part of the Romanian National Program for Power Reactor Development, with particular emphasis being placed on

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

    Microsoft Academic Search

    David W. Nigg; Devin A. Steuhm

    2011-01-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of

  20. Performance characteristics of the annular core research reactor fuel motion detection system

    Microsoft Academic Search

    J. G. Kelly; K. T. Stalker

    1983-01-01

    Recent proof tests have shown that the annular core research reactor (ACRR) fuel motion detection system has reached its design goals of providing high temporal and spatial resolution pictures of fuel distributions in the ACRR. The coded aperture imaging system (CAIS) images the fuel by monitoring the fission gamma rays from the fuel that pass through collimators in the reactor

  1. Safety and core design of large liquid-metal cooled fast breeder reactors

    NASA Astrophysics Data System (ADS)

    Qvist, Staffan Alexander

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

  2. Sodium Based Heat Pipe Modules for Space Reactor Concepts: Stainless Steel SAFE100 Core

    Microsoft Academic Search

    James J. Martin; Robert S. Reid

    2004-01-01

    A heat pipe cooled reactor is one of several candidate reactor cores being considered for advanced space power and propulsion systems to support future space exploration applications. Long life heat pipe modules, with designs verified through a combination of theoretical analysis and experimental lifetime evaluations, would be necessary to establish the viability of any of these candidates, including the heat

  3. Investigation of aerosols released at high temperature from nuclear reactor core models

    Microsoft Academic Search

    A Pintér Csordás; L. Matus; A. Czitrovszky; P. Jani; L. Maroti; Z. Hozer; P. Windberg; R. Hummel

    2000-01-01

    Two experiments were performed to simulate severe reactor accident with air ingress into the hot reactor core. The model bundles contained nine PWR type fuel rods. Their cladding was pre-oxidised by argon–oxygen (test 1) and steam (test 2). The released aerosol was measured continuously by laser particle counters. Morphology and elemental composition of the aerosol particles were studied on samples

  4. Irradiation capabilities of LR-0 reactor with VVER-1000 Mock-Up core.

    PubMed

    Koš?ál, Michal; Rypar, Vojt?ch; Svadlenková, Marie; Cvachovec, František; Jánský, Bohumil; Mil?ák, Ján

    2013-12-01

    Even low power reactors, such as zero power reactors, are sufficient for semiconductor radiation hardness effect investigation. This reflects the fact that fluxes necessary for affecting semiconductor electrical resistance are much lower than fluxes necessary to affect material parameters. The paper aims to describe the irradiation possibilities of the LR-0 reactor with a special core arrangement corresponding to VVER-1000 dosimetry Mock-Up. PMID:24050946

  5. FREC-II: An upgrade to SNL's annular core research reactor

    Microsoft Academic Search

    R. A. Rubio; P. J. Cooper; J. F. Schulze; J. W. Bryson; F. M. Morris; F. R. Trowbridge; T. R. Schmidt

    1989-01-01

    The fuel-ringed external cavity, version II (FREC-II), is a recent upgrade to the annular core research reactor (ACRR) at Sandia National Laboratories (SNL). The FREC-II is neutronically coupled to the ACRR, a 2-MW steady-state\\/300-MJ pulse reactor used for a variety of simulation experiments in areas such as reactor safety and weapons effects. The FREC-II was designed to provide a large-volume

  6. Core design investigation for a SUPERSTAR small modular lead-cooled fast reactor demonstrator

    Microsoft Academic Search

    S. Bortot; A. Moisseytsev; J. J. Sienicki; Carlo Artioli

    In this paper a preconceptual neutronics design study for a SUstainable Proliferation-resistance Enhanced Refined Secure Transportable Autonomous Reactor (SUPERSTAR) demonstrator is presented. The main goal of achieving the highest realistic power level limited by natural circulation and transportability, while providing energy security and proliferation resistance thanks to a long core lifetime design has been satisfactorily attained. A preliminary core configuration

  7. Preparations to load, transport, receive, and store the damaged TMI2 (Three Mile Island) reactor core

    Microsoft Academic Search

    H. W. Reno; R. C. Schmitt; G. J. Quinn; A. L. Jr. Ayers; B. J. Jr. Lilburn; D. L. Uhl

    1986-01-01

    The March 1979 incident at the Three Mile Island Nuclear Power Station (TMI) which damaged the core of the Unit 2 reactor resulted in numerous scientific and technical challenges. Some of those challenges involve removing, packaging, and transporting the core debris to the Idaho National Engineering Laboratory (INEL) for storage, examination, and preparation for final disposal. This paper highlights preparations

  8. A study of the structural integrity of the core support structure of a fast breeder reactor

    Microsoft Academic Search

    M. Ueta; M. Ichimiya; H. Hirayama; M. Asano; H. Ikeuchi; K. Sekine; T. Kodama; K. Sato

    1992-01-01

    This paper reports on the core support structure of a fast breeder reactor supports the fuel assemblies, supplies sodium coolant to the fuel assemblies, and maintains the insertability of control rods even during an earthquake. The core support structure is designed as a box fabricated of welded plates, ribs, and cylinders that distribute the load in a diverse manner, in

  9. Thermal Design of an Ultrahigh Temperature Vapor Core Reactor Combined Cycle Nuclear Power Plant

    Microsoft Academic Search

    Samuel E. Bays; Samim Anghaie; Blair Smith; Travis Knight

    2004-01-01

    Current work modeling high temperature compact heat exchangers may demonstrate the design feasibility of a Vapor Core Reactor (VCR) driven combined cycle power plant. For solid nuclear fuel designs, the cycle efficiency is typically limited by a metallurgical temperature limit which is dictated by fuel and structural melting points. In a vapor core, the gas\\/vapor phase nuclear fuel is uniformly

  10. Iridium Interfacial Stack (IRIS)

    NASA Technical Reports Server (NTRS)

    Spry, David James (Inventor)

    2015-01-01

    An iridium interfacial stack ("IrIS") and a method for producing the same are provided. The IrIS may include ordered layers of TaSi.sub.2, platinum, iridium, and platinum, and may be placed on top of a titanium layer and a silicon carbide layer. The IrIS may prevent, reduce, or mitigate against diffusion of elements such as oxygen, platinum, and gold through at least some of its layers.

  11. IRIS Agenda and Literature Searches

    EPA Science Inventory

    IRIS is an EPA database of human health effects that may result from exposure to chemical substances found in the environment. EPA's process for developing IRIS assessments is described in detail on the IRIS Process Web page...

  12. Control rod worth and related nuclear characteristics of an axially heterogeneous liquidmetal fast breeder reactor core

    Microsoft Academic Search

    K. Kawashima; T. Inayaki; K. Inoue; K. Kaneto

    1985-01-01

    An axially heterogeneous core (AHC) concept is applied to a 1000-MW(electric)-class tank-type liquidmetal fast breeder reactor (LMFBR). This AHC is characterized by a disk-shaped internal blanket with a radial thickness adjustment at the core midplane. The nuclear characteristics connected with control rod worth of the AHC are analyzed and compared with those of a homogeneous core (HOC) of the same

  13. Evaluation of LMR (liquid metal reactors) core support concepts under seismic events

    SciTech Connect

    Wu, Ting-shu; Seidensticker, R.W.

    1989-01-01

    In the design of the core support system for liquid metal reactors (LMR) against earthquakes, the major concerns are directed toward the structural integrity as well as the reactivity control. This means that, in addition to the stress levels, maximum displacements and accelerations should also be within their allowable limits. This investigation studies the seismic responses of a large pool-type LMR with different design approaches to support the reactor core. Different core support designs yield different frequency ranges and responses. Responses of these designs to the given floor response spectra are required to satisfy a set of criteria which are common to all designs. 5 refs., 4 figs.

  14. Nuclear reactor with low-level core coolant intake

    DOEpatents

    Challberg, Roy C. (Livermore, CA); Townsend, Harold E. (Campbell, CA)

    1993-01-01

    A natural-circulation boiling-water reactor has skirts extending downward from control rod guide tubes to about 10 centimeters from the reactor vessel bottom. The skirts define annular channels about control rod drive housings that extend through the reactor vessel bottom. Recirculating water is forced in through the low-level entrances to these channels, sweeping bottom water into the channels in the process. The sweeping action prevents cooler water from accumulating at the bottom. This in turn minimizes thermal shock to bottom-dwelling components as would occur when accumulated cool water is swept away and suddenly replaced by warmer water.

  15. Optimization of hydride fueled pressurized water reactor cores

    E-print Network

    Shuffler, Carter Alexander

    2004-01-01

    This thesis contributes to the Hydride Fuels Project, a collaborative effort between UC Berkeley and MIT aimed at investigating the potential benefits of hydride fuel use in light water reactors (LWRs). This pursuit involves ...

  16. Heat exchanger for reactor core and the like

    DOEpatents

    Kaufman, Jay S. (Del Mar, CA); Kissinger, John A. (Del Mar, CA)

    1986-01-01

    A compact bayonet tube type heat exchanger which finds particular application as an auxiliary heat exchanger for transfer of heat from a reactor gas coolant to a secondary fluid medium. The heat exchanger is supported within a vertical cavity in a reactor vessel intersected by a reactor coolant passage at its upper end and having a reactor coolant return duct spaced below the inlet passage. The heat exchanger includes a plurality of relatively short length bayonet type heat exchange tube assemblies adapted to pass a secondary fluid medium therethrough and supported by primary and secondary tube sheets which are releasibly supported in a manner to facilitate removal and inspection of the bayonet tube assemblies from an access area below the heat exchanger. Inner and outer shrouds extend circumferentially of the tube assemblies and cause the reactor coolant to flow downwardly internally of the shrouds over the tube bundle and exit through the lower end of the inner shroud for passage to the return duct in the reactor vessel.

  17. IRIS Process (Pre-2004)

    EPA Science Inventory

    The Integrated Risk Information System (IRIS) is a U.S. Environmental Protection Agency (EPA) database that contains information on human health effects that may result from exposure to chemical substances in the environment. IRIS is maintained by EPA?s Office of Research and Dev...

  18. IRIS Process (2008)

    EPA Science Inventory

    The Integrated Risk Information System (IRIS) is a U.S. Environmental Protection Agency (EPA) database that contains information on human health effects that may result from exposure to chemical substances in the environment. IRIS is maintained by EPA?s Office of Research and De...

  19. IRIS Process (2009 Update)

    EPA Science Inventory

    The Integrated Risk Information System (IRIS) is a U.S. Environmental Protection Agency (EPA) database that contains information on human health effects that may result from exposure to chemical substances in the environment. IRIS is maintained by EPA?s Office of Research and Dev...

  20. Approaches for achieving very high core outlet temperatures in prismatic modular helium reactors

    SciTech Connect

    LaBar, M.; Richards, M.; Shenoy, A. [General Atomics, 3550 General Atomics Court, San Diego, CA 92121-1122 (United States)

    2004-07-01

    High Temperature Gas Reactors (HTGRs) cooled by helium have the capability to develop high core outlet temperatures. The upper temperature bound of HTGRs designed and operated to-date is approximately 950 deg. C. But, the goal for the Next Generation Nuclear Plant (NGNP) is a mixed mean core outlet temperature of 1000 deg. C. The most limiting core design criteria governing core outlet temperature in HTGRs is diffusive fission product release from the fuel. For the reference TRISO coated particle fuel used by HTGRs, the rule of thumb for maintaining fission product releases low enough to meet regulatory limits, is a maximum peak fuel temperature for normal operation of about 1250 deg. C. The Gas-Turbine Modular Helium Reactor (GT-MHR), which employs a prismatic core design, has a design mixed mean core outlet temperature of 850 deg. C. Initial evaluations have been made for a core of the type employed by the GT-MHR to determine the feasibility of increasing the core outlet temperature to 1000 deg. C. The conclusion has been reached that design approaches are available for making the 1000 deg. C temperature potentially achievable, while still meeting peak fuel temperature limits albeit with reduced design margin. These potential design approaches and their effects for achieving a 1000 deg. C core outlet temperature in a prismatic core design are described. (authors)

  1. Internal Control Rod Drive Mechanisms, Design Options for IRIS

    SciTech Connect

    Conway, Lawrence E.; Petrovic, Bojan [Westinghouse Electric Company, Science and Technology Department, 1344 Beulah Rd, Pittsburgh, PA 15235 (United States)

    2004-07-01

    IRIS (International Reactor Innovative and Secure) is a medium-power (335 MWe) PWR with an integral, primary circuit configuration, where all the reactor coolant system components are contained within the reactor vessel. This integral configuration is a key reason for the success of IRIS' 'safety-by-design' approach, whereby accident initiators are eliminated or the accident consequences and/or frequency are reduced. The most obvious example of the IRIS safety by design approach is the elimination of large LOCA's, since the integral reactor coolant system has no large loop piping. Another serious accident scenario that is being addressed in IRIS is the postulated ejection of a reactor control cluster assembly (RCCA). This accident initiator can be eliminated by locating the RCCA drive mechanisms (CRDMs) inside the reactor vessel. This eliminates the mechanical drive rod penetration between the RCCA and the external CRDM, eliminating the potential for differential pressure across the pressure boundary, and thus eliminating 'by design' the possibility for rod ejection accident. Moreover, the elimination of the 'large' drive-rod penetrations and the external CRDM pressure housings decreases the likelihood of boric acid leakage and subsequent corrosion of the reactor pressure boundary (like the Davis-Besse incident). This paper will discuss the IRIS top level design requirements and objectives for internal CRDMs, and provide examples candidate designs and their specific performance characteristics. (authors)

  2. Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor

    SciTech Connect

    B. Boer; A. M. Ougouag

    2010-09-01

    The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge burnup level, while retaining its inherent safety characteristics. Using generic pebble bed reactor cores, this task will perform physics calculations to evaluate the capabilities of the pebble bed reactor to perform utilization and destruction of LWR used-fuel transuranics. The task will use established benchmarked models, and will introduce modeling advancements appropriate to the nature of the fuel considered (high TRU content and high burn-up).

  3. NIOBIUM--PROMISING HIGH-TEMPERATURE REACTOR-CORE MATERIAL

    Microsoft Academic Search

    J. A. DeMastry; R. F. Dickerson

    1960-01-01

    The properties of niobium and niobium-base alloys were studied for high ; temperature reactor use. Results indicated that niobium-based, enriched-uranium ; alloys may be valuable as fuels for use at 1600 to 2000 deg F in a nonoxidizing ; environment. The properties of unalloyed niobium, capture cross sections, high ; temperature strengths, effects of cold work on hardness, electrical resistivity

  4. Analysis of Air-Core Reactors From DC to Very High Frequencies Using PEEC Models

    Microsoft Academic Search

    Mathias Enohnyaket; Jonas Ekman

    2009-01-01

    Faced with the challenges of increasing operational frequencies and switching rates of modern power-electronics devices used in power systems, there is need for high-frequency models (up to a few megahertz) for power components, such as reactors, capacitor banks, and transformers. This paper presents the application of PEEC theory for the creation of high-frequency, electromagnetic (EM) models for air-core reactors. The

  5. A Metropolis algorithm combined with Nelder–Mead Simplex applied to nuclear reactor core design

    Microsoft Academic Search

    Wagner F. Sacco; Hermes Alves Filho; Nélio Henderson; Cassiano R. E. de Oliveira

    2008-01-01

    A hybridization of the recently introduced Particle Collision Algorithm (PCA) and the Nelder–Mead Simplex algorithm is introduced and applied to a core design optimization problem which was previously attacked by other metaheuristics. The optimization problem consists in adjusting several reactor cell parameters, such as dimensions, enrichment and materials, in order to minimize the average peak-factor in a three-enrichment-zone reactor, considering

  6. Potential impact of low-enriched uranium research reactor core conversions on irradiation facilities for BNCT

    SciTech Connect

    Heimberger, L.A.; Aldemir, T.

    1992-01-01

    A critical requirement for boron neutron capture therapy (BNCT) is a strong source of thermal (<1-eV) or epithermal (1-eV to 10-keV) neutrons. The currently available neutron sources with sufficient intensity are research and test reactors. Since the neutrons generated in nuclear reactor cores cover a wide spectrum of energies and the effectiveness of BNCT strongly depends on how well the fast neutron dose to healthy tissue is minimized, the core neutron spectrum is an important consideration in the design of an irradiation facility for BNCT. A large number of research and test reactors in the world use highly enriched uranium (HEU) fuels. In view of the restrictions on the export of HEU from the United States and also the 1986 US Nuclear Regulatory Commission ruling, these reactors are in the process of being converted to use low-enriched uranium (LEU) fuels. The conversion process usually leads to core spectrum hardening. The objective of this paper is to present the results of a three-dimensional Monte Carlo study that investigates the potential impacts of such a core spectrum hardening on an irradiation facility utilizing the reactor as a thermal neutron source.

  7. Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core.

    PubMed

    Lashkari, A; Khalafi, H; Kazeminejad, H

    2013-05-01

    In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change. PMID:24976672

  8. Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

    PubMed Central

    Lashkari, A.; Khalafi, H.; Kazeminejad, H.

    2013-01-01

    In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change. PMID:24976672

  9. Feasibility study on nuclear core design for soluble boron free small modular reactor

    NASA Astrophysics Data System (ADS)

    Rabir, Mohamad Hairie; Hah, Chang Joo; Ju, Cho Sung

    2015-04-01

    A feasibility study on nuclear core design of soluble boron free (SBF) core for small size (150MWth) small modular reactor (SMR) was investigated. The purpose of this study was to design a once through cycle SMR core, where it can be used to supply electricity to a remote isolated area. PWR fuel assembly design with 17×17 arrangement, with 264 fuel rods per assembly was adopted as the basis design. The computer code CASMO-3/MASTER was used for the search of SBF core and fuel assembly analysis for SMR design. A low critical boron concentration (CBC) below 200 ppm core with 4.7 years once through cycle length was achieved using 57 fuel assemblies having 170 cm of active height. Core reactivity controlled using mainly 512 number of 4 wt% and 960 12 wt% Gd rods.

  10. Neutronic analysis of boiling water reactor in-core detector noise

    Microsoft Academic Search

    H. S. Cheng; D. J. Diamond

    1979-01-01

    The response of boiling water reactor in-core detectors undergoing vibration has been calculated. A neutronic model based on calculating the fission activity at a detector position in a planar multibundle environment was employed. The model used eight energy groups and two-dimensional Cartesian geometry in a discrete-ordinates transport approximation. The in-core detector responses due to various detector displacements were calculated as

  11. An on-line fuel management method using in-core detector readings for CANDU reactors

    Microsoft Academic Search

    Chang Joon Jeong

    2002-01-01

    An on-line fuel management method for a CANDU reactor has been developed. In the method, the in-core detector readings are used for channel power generation for refueling channel selection. The in-core detector readings are converted to measured mesh readings, and the Kalman filtering technique is applied to reduce calculation and measurement errors of the mesh readings. Then, the estimated channel

  12. Neutronics analysis of an open-cycle high-impulse gas core reactor concept

    NASA Technical Reports Server (NTRS)

    Whitmarsh, C. L., Jr.

    1972-01-01

    A procedure was developed to calculate the critical fuel mass, including the effects of propellant pressure, for coaxial-flow gas-core reactors operating at 196,600 newtons thrust and 4400 seconds specific impulse. Data were generated for a range of cavity diameter, reflector-moderator thickness, and quantity of structural material. Also presented are such core characteristics as upper limits on cavity pressure, spectral hardening in very-high-temperature hydrogen, and reactivity coefficients.

  13. Feasibility study of boiling water reactor core based on thorium–uranium fuel concept

    Microsoft Academic Search

    Alejandro Núñez-Carrera; Juan Luis François Lacouture; Cecilia Martin del Campo; Gilberto Espinosa-Paredes

    2008-01-01

    The design of a boiling water reactor (BWR) equilibrium core using the thorium–uranium (blanket–seed) concept in the same integrated fuel assembly is presented in this paper. The lattice design uses the thorium conversion capability to 233U in a BWR spectrum. A core design was developed to achieve an equilibrium cycle of one effective full power year in a standard BWR

  14. Development of an advanced core analysis system for boiling water reactor designs

    Microsoft Academic Search

    Hiromi Maruyama; Junichi Koyama; Motoo Aoyama; Kazuya Ishii; Atsushi Zukeran; Takashi Kiguchi; Akira Nishimura

    1997-01-01

    A core analysis system has been developed for the recent advanced designs of boiling water reactors. This system consists of a fuel assembly analysis code VMONT and a three-dimensional core simulator COSNEX. To cope with heterogeneous structures found in the recent high-performance fuel, VMONT employs a Monte Carlo neutron transport calculation method. COSNEX is based on a three-group nodal expansion

  15. Evaluation of the Start-Up Core Physics Tests at Japan's High Temperature Engineering Test Reactor (Annular Core Loadings)

    SciTech Connect

    John D. Bess; Nozomu Fujimoto; James W. Sterbentz; Luka Snoj; Atsushi Zukeran

    2010-03-01

    The High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Agency (JAEA) is a 30 MWth, graphite-moderated, helium-cooled reactor that was constructed with the objectives to establish and upgrade the technological basis for advanced high-temperature gas-cooled reactors (HTGRs) as well as to conduct various irradiation tests for innovative high-temperature research. The core size of the HTTR represents about one-half of that of future HTGRs, and the high excess reactivity of the HTTR, necessary for compensation of temperature, xenon, and burnup effects during power operations, is similar to that of future HTGRs. During the start-up core physics tests of the HTTR, various annular cores were formed to provide experimental data for verification of design codes for future HTGRs. The Japanese government approved construction of the HTTR in the 1989 fiscal year budget; construction began at the Oarai Research and Development Center in March 1991 and was completed May 1996. Fuel loading began July 1, 1998, from the core periphery. The first criticality was attained with an annular core on November 10, 1998 at 14:18, followed by a series of start-up core physics tests until a fully-loaded core was developed on December 16, 1998. Criticality tests were carried out into January 1999. The first full power operation with an average core outlet temperature of 850ºC was completed on December 7, 2001, and operational licensing of the HTTR was approved on March 6, 2002. The HTTR attained high temperature operation at 950 ºC in April 19, 2004. After a series of safety demonstration tests, it will be used as the heat source in a hydrogen production system by 2015. Hot zero-power critical, rise-to-power, irradiation, and safety demonstration testing , have also been performed with the HTTR, representing additional means for computational validation efforts. Power tests were performed in steps from 0 to 30 MW, with various tests performed at each step to confirm core characteristics, thermal-hydraulic properties, and radiation shielding. The high-temperature test operation at 950 ºC represented the fifth and final phase of the rise-to-power tests. The safety tests demonstrated inherent safety features of the HTTR such as slow temperature response during abnormal events due to the large heat capacity of the core and the negative reactivity feedback. The experimental benchmark performed and currently evaluated in this report pertains to the data available for the annular core criticals from the initial six isothermal, annular and fully-loaded, core critical measurements performed at the HTTR. Evaluation of the start-up core physics tests specific to the fully-loaded core is compiled elsewhere (HTTR-GCR-RESR-001).

  16. Nodal Diffusion Burnable Poison Treatment for Prismatic Reactor Cores

    SciTech Connect

    A. M. Ougouag; R. M. Ferrer

    2010-10-01

    The prismatic block version of the High Temperature Reactor (HTR) considered as a candidate Very High Temperature Reactor (VHTR)design may use burnable poison pins in locations at some corners of the fuel blocks (i.e., assembly equivalent structures). The presence of any highly absorbing materials, such as these burnable poisons, within fuel blocks for hexagonal geometry, graphite-moderated High Temperature Reactors (HTRs) causes a local inter-block flux depression that most nodal diffusion-based method have failed to properly model or otherwise represent. The location of these burnable poisons near vertices results in an asymmetry in the morphology of the assemblies (or blocks). Hence the resulting inadequacy of traditional homogenization methods, as these “spread” the actually local effect of the burnable poisons throughout the assembly. Furthermore, the actual effect of the burnable poison is primarily local with influence in its immediate vicinity, which happens to include a small region within the same assembly as well as similar regions in the adjacent assemblies. Traditional homogenization methods miss this artifact entirely. This paper presents a novel method for treating the local effect of the burnable poison explicitly in the context of a modern nodal method.

  17. A Burst Mode, Ultrahigh Temperature UF4 Vapor Core Reactor Rankine Cycle Space Power System Concept

    NASA Technical Reports Server (NTRS)

    Dugan, E. T.; Kahook, S. D.; Diaz, N. J.

    1996-01-01

    Static and dynamic neutronic analyses have been performed on an innovative burst mode (100's of MW output for a few thousand seconds) Ulvahigh Temperature Vapor Core Reactor (UTVR) space nuclear power system. The NVTR employs multiple, neutronically-coupled fissioning cores and operates on a direct, closed Rankine cycle using a disk Magnetohydrodynamic (MHD) generater for energy conversion. The UTVR includes two types of fissioning core regions: (1) the central Ultrahigh Temperature Vapor Core (UTVC) which contains a vapor mixture of highly enriched UF4 fuel and a metal fluoride working fluid and (2) the UF4 boiler column cores located in the BeO moderator/reflector region. The gaseous nature of the fuel the fact that the fuel is circulating, the multiple coupled fissioning cores, and the use of a two phase fissioning fuel lead to unique static and dynamic neutronic characteristics. Static neutronic analysis was conducted using two-dimensional S sub n, transport theory calculations and three-dimensional Monte Carlo transport theory calculations. Circulating-fuel, coupled-core point reactor kinetics equations were used for analyzing the dynamic behavior of the UTVR. In addition to including reactivity feedback phenomena associated with the individual fissioning cores, the effects of core-to-core neutronic and mass flow coupling between the UTVC and the surrounding boiler cores were also included in the dynamic model The dynamic analysis of the UTVR reveals the existence of some very effectlve inherent reactivity feedback effects that are capable of quickly stabilizing this system, within a few seconds, even when large positive reactivity insertions are imposed. If the UTVC vapor fuel density feedback is suppressed, the UTVR is still inherently stable because of the boiler core liquid-fuel volume feedback; in contrast, suppression of the vapor fuel density feedback in 'conventional" gas core cavity reactors causes them to become inherently unstable. Due to the strength of the negative reactivity feedback in the UTVR, it is found that external reactivity insertions alone are inadequate for bringing about significant power level changes during normal reactor operations. Additional methods of reactivity control such as variations in the gaseous fuel mass flow rate, are needed to achieve the desired power level oontrol.

  18. PREDICTION OF MASS FLOW RATE AND PRESSURE DROP IN THE COOLANT CHANNEL OF THE TRIGA 2000 REACTOR CORE

    Microsoft Academic Search

    Efrizon Umar

    2001-01-01

    PREDICTION OF MASS FLOW RATE AND PRESSURE DROP IN THE COOLANT CHANNEL OF THE TRIGA 2000 REACTOR CORE. Prediction of the mass flow rate and pressure drop in the coolant channel of the TRIGA 2000 reactor core have been carried out and the value of mass flow rate and pressure drop for the maximum and average powered fuel rod have

  19. Cosmic Ray Radiography of the Damaged Cores of the Fukushima Reactors

    E-print Network

    Konstantin Borozdin; Steven Greene; Zarija Luki?; Edward Cas Milner; Haruo Miyadera; Christopher Morris; John Perry

    2012-09-13

    The passage of muons through matter is dominated by the Coulomb interaction with electrons and nuclei. The interaction with the electrons leads to continuous energy loss and stopping of the muons. The interaction with nuclei leads to angle diffusion. Two muon imaging methods that use flux attenuation and multiple Coulomb scattering of cosmic-ray muons are being studied as tools for diagnosing the damaged cores of the Fukushima reactors. Here we compare these two methods. We conclude that the scattering method can provide detailed information about the core. Attenuation has low contrast and little sensitivity to the core.

  20. Implications for accident management of adding water to a degrading reactor core

    SciTech Connect

    Kuan, P.; Hanson, D.J.; Pafford, D.J.; Quick, K.S.; Witt, R.J. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1994-02-01

    This report evaluates both the positive and negative consequences of adding water to a degraded reactor core during a severe accident. The evaluation discusses the earliest possible stage at which an accident can be terminated and how plant personnel can best respond to undesired results. Specifically discussed are (a) the potential for plant personnel to add water for a range of severe accidents, (b) the time available for plant personnel to act, (c) possible plant responses to water added during the various stages of core degradation, (d) plant instrumentation available to understand the core condition and (e) the expected response of the instrumentation during the various stages of severe accidents.

  1. Cosmic ray radiography of the damaged cores of the Fukushima reactors.

    PubMed

    Borozdin, Konstantin; Greene, Steven; Luki?, Zarija; Milner, Edward; Miyadera, Haruo; Morris, Christopher; Perry, John

    2012-10-12

    The passage of muons through matter is dominated by the Coulomb interaction with electrons and nuclei. The interaction with the electrons leads to continuous energy loss and stopping of the muons. The interaction with nuclei leads to angle "diffusion." Two muon-imaging methods that use flux attenuation and multiple Coulomb scattering of cosmic-ray muons are being studied as tools for diagnosing the damaged cores of the Fukushima reactors. Here, we compare these two methods. We conclude that the scattering method can provide detailed information about the core. Attenuation has low contrast and little sensitivity to the core. PMID:23102302

  2. Reducing numerical costs for core wide nuclear reactor CFD simulations by the Coarse-Grid-CFD

    NASA Astrophysics Data System (ADS)

    Viellieber, Mathias; Class, Andreas G.

    2013-11-01

    Traditionally complete nuclear reactor core simulations are performed with subchannel analysis codes, that rely on experimental and empirical input. The Coarse-Grid-CFD (CGCFD) intends to replace the experimental or empirical input with CFD data. The reactor core consists of repetitive flow patterns, allowing the general approach of creating a parametrized model for one segment and composing many of those to obtain the entire reactor simulation. The method is based on a detailed and well-resolved CFD simulation of one representative segment. From this simulation we extract so-called parametrized volumetric forces which close, an otherwise strongly under resolved, coarsely-meshed model of a complete reactor setup. While the formulation so far accounts for forces created internally in the fluid others e.g. obstruction and flow deviation through spacers and wire wraps, still need to be accounted for if the geometric details are not represented in the coarse mesh. These are modelled with an Anisotropic Porosity Formulation (APF). This work focuses on the application of the CGCFD to a complete reactor core setup and the accomplishment of the parametrization of the volumetric forces.

  3. Experimental Breeder Reactor II (EBR-II): Instrumentation for core surveillance

    SciTech Connect

    Christensen, L.J.

    1989-01-01

    EBR-II has operated for 25 years in support of several major programs. During this time period, several of the original, non-replaceable, flow sensors, RDT sensors and thermocouples have failed in the primary system. This has led to the development of new sensors and the use of calculated values using computer models of the plant. It is important for the next generation of LMR reactors to minimize or eliminate the use of non-replaceable sensors. EBR-II is perhaps the best modeled reactor in the world, thanks to a dedicated T-H analysis program. The success of this program relied on excellent measurements of temperature and flow in subassemblies in the core. The instrumented subassemblies of the XX series provided that measurement capability. From this test series, EBR-II calculations showed that the core could withstand a loss-of-flow without scram accident and a loss-of-heat sink without scram accident from full reactor power without core damage. From this, reactor designers can now design with confidence, inherently safe reactors. 11 refs., 8 figs.

  4. Evaluation of core damage sequences initiated by loss of reactor coolant pump seal cooling

    Microsoft Academic Search

    S. Mitra; R. Baradaran; R. Youngblood

    1986-01-01

    This report is concerned with core damage accident sequences initiated by loss of component cooling water, leading to loss of reactor coolant pump seal cooling, subsequent primary coolant leakage, and failure to make up the coolant loss. Three plants are considered: Indian Point Unit 3, Midland Unit 2, and Calvert Cliffs Unit 1. It is shown that design differences in

  5. Preliminary fracture analysis of the core pressure boundary tube for the Advanced Neutron Source Research Reactor

    Microsoft Academic Search

    K. C. Schulz; G. T. Yahr

    1995-01-01

    The outer core pressure boundary tube (CPBT) of the Advanced neutron Source (ANS) reactor being designed at Oak Ridge National Laboratory is currently specified as being composed of 6061-T6 aluminum. ASME Boiler and Pressure Vessel Code fracture analysis rules for nuclear components are based on the use of ferritic steels; the expressions, tables, charts and equations were all developed from

  6. A fuel freezing model for liquid-metal fast breeder reactor hypothetical core disruptive accidents

    Microsoft Academic Search

    F. R. Best; C. Erdman; D. Wayne

    1985-01-01

    A proposed fuel freezing mechanism for molten UO2 fuel penetrating a steel channel was investigated in the course of liquid-metal-cooled fast breeder reactor hypothetical core disruptiv accident safety studies. The fuel crust deposited on an underlying melting steel wall was analyzed as being subjected to two stresses one due to the pressure difference between the flowing fuel and the stagnant

  7. Spring design for use in the core of a nuclear reactor

    DOEpatents

    Willard, Jr., H. James (Bethel Park, PA)

    1993-01-01

    A spring design particularly suitable for use in the core of a nuclear reactor includes one surface having a first material oriented in a longitudinal direction, and another surface having a second material oriented in a transverse direction. The respective surfaces exhibit different amounts of irraditation induced strain.

  8. Shear Strengths of Copper\\/Insulation Interfaces After In-Core Reactor Irradiations

    Microsoft Academic Search

    S. D. Grandlienard; M. W. Hooker; P. E. Fabian; M. E. Sawan; T. H. Newton; A. R. Grein

    2007-01-01

    The adhesive shear strengths of copper\\/insulation interfaces were evaluated using a novel specimen design in which thin copper foils were embedded in laminate structures. In each instance, the copper surface was either chemically or physically treated prior to laminate fabrication. Once produced, the specimens were subjected to in-core reactor irradiations providing total radiation doses of 10, 22, and 100 MGy

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

    NASA Technical Reports Server (NTRS)

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

    2009-01-01

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

  10. Effect of debris bed pressure, particle size, and distribution on degraded nuclear reactor core coolability

    Microsoft Academic Search

    D. Squarer; A. T. Pieczynski; L. E. Hochreiter

    1982-01-01

    In the worst hypothetical accident of a light water reactor (LWR), when all protection systems fail, the core could melt and be converted to a deep particulate bed as a result of molten-fuel-coolant interaction. The containment of such an accident depends on the coolability of the heat generating particulate bed. This paper summarizes published theoretical analyses that may predict bed

  11. Temperature monitoring of nuclear reactor cores with multiplexed fiber Bragg grating sensors

    Microsoft Academic Search

    Alberto F. Fernandez; Andrei I. Gusarov; Benoit Brichard; S. Bodart; K. Lammens; Francis Berghmans; Marc C. Decreton; Patrice Megret; Michel Blondel; Alain Delchambre

    2002-01-01

    In-core temperature measurement is a critical issue for the safe operation of nuclear reactors. Classical thermocouples require shielded connections and are known to drift under high neutron fluence. As an alternative, we propose to take advantage of the multiplexing capabilities of fiber Bragg grating (FBG) temperature sensors. Our experiments show that sensitivity to radiation depends on both the radiation field

  12. FEASIBILITY STUDIES--NONDESTRUCTIVE TESTING OF THE ENRICO FERMI REACTOR CORE B FUEL ELEMENT

    Microsoft Academic Search

    McClung

    1962-01-01

    A series of feasibility studies which werc conducted to determine the ; capabilities and limitatnnons of several nondestructive testing methods as ; applnned to the Core B fuel element of the Enrico Fermi Fast Breeder Reactor are ; discussed. An eddy-current technique is demonstrated to be capable of measuring ; the fuel plate-clad thickness wtth an accuracy of plus or

  13. Energetics of core disruptive accident for different fuels for a medium sized fast reactor

    Microsoft Academic Search

    Om Pal Singh; R Harish

    2002-01-01

    A comparative study has been made on the mechanical energy released in a core disruptive accident resulting from an unprotected loss of flow accident (LOFA) in a medium sized liquid metal fast breeder reactor with oxide, carbide and metal fuels. The study is conducted by ignoring the passive safety features incorporated in the design so that the accident scenario culminates

  14. Man--machine communication system for boiling water reactor core management planning

    Microsoft Academic Search

    O. Yokomizo; H. Motoda; T. Kiguchi; R. Takeda

    1976-01-01

    A man-machine communication system has been developed for boiling water reactor (BWR) core management planning to provide a very flexible tool, which is complementary to automated optimization programs that maximize or minimize one particular performance index under certain constraints. A three-dimensional BWR simulator, which can cover a wide range of BWR operating conditions, has been developed and coupled with a

  15. Application of gaseous core reactors for transmutation of nuclear waste

    NASA Technical Reports Server (NTRS)

    Schnitzler, B. G.; Paternoster, R. R.; Schneider, R. T.

    1976-01-01

    An acceptable management scheme for high-level radioactive waste is vital to the nuclear industry. The hazard potential of the trans-uranic actinides and of key fission products is high due to their nuclear activity and/or chemical toxicity. Of particular concern are the very long-lived nuclides whose hazard potential remains high for hundreds of thousands of years. Neutron induced transmutation offers a promising technique for the treatment of problem wastes. Transmutation is unique as a waste management scheme in that it offers the potential for "destruction" of the hazardous nuclides by conversion to non-hazardous or more manageable nuclides. The transmutation potential of a thermal spectrum uranium hexafluoride fueled cavity reactor was examined. Initial studies focused on a heavy water moderated cavity reactor fueled with 5% enriched U-235-F6 and operating with an average thermal flux of 6 times 10 to the 14th power neutrons/sq cm-sec. The isotopes considered for transmutation were I-129, Am-241, Am-242m, Am-243, Cm-243, Cm-244, Cm-245, and Cm-246.

  16. A feasibility study of ferro-boron as in-core shield material in fast breeder reactors

    Microsoft Academic Search

    D. Sunil Kumar; R. S. Keshavamurthy; P. Mohanakrishnan; S. C. Chetal

    2010-01-01

    Shields around core and blankets form major part of reactor assembly in fast reactors as the incident neutron spectrum is hard with negligible thermal component and has anisotropic angular distribution. In this paper, a study is presented on the use of ferro-boron as neutron shield material in pool type fast reactors. The reference case chosen is the Prototype Fast Breeder

  17. Evolution of the core physics concept for the Canadian supercritical water reactor

    SciTech Connect

    Pencer, J.; Colton, A.; Wang, X.; Gaudet, M.; Hamilton, H.; Yetisir, M. [Atomic Energy of Canada, Ltd., Chalk River Laboratories, Chalk River, ON (Canada)

    2013-07-01

    The supercritical water cooled reactor (SCWR) is one of the advanced reactor concepts chosen by the GEN-IV International Forum (GIF) for research and development efforts. Canada's contribution is the Canadian SCWR, a heavy water moderated, pressure tube supercritical light water cooled reactor. Recent developments in the SCWR lattice and core concepts, primarily the introduction of a large central flow tube filled with coolant combined with a two-ring fuel assembly, have enabled significant improvements compared to earlier concepts. These improvements include a reduction in coolant void reactivity (CVR) by more than 10 mk, and an almost 40% increase in fuel exit burnup, which is achieved via balanced power distribution between the fuel pins in the fuel assembly. In this paper the evolution of the physics concept is reviewed, and the present lattice and core physics concepts are presented.

  18. New Concept of a Small Passive-Safety Reactor with UO{sub 2}-Graphite-Water Core

    SciTech Connect

    Tetsuo Matsumura; Takanori Kameyama; Yasushi Nauchi; Izumi Kinoshita [Central Research Institute of Electric Power Industry (CRIEPI), 2-11-l. lwado Kita, Komae-shi, Tokyo 201-8.511 (Japan)

    2002-07-01

    New concept of a passive-safety reactor with I/O:-graphite-water core is proposed, which has negligible possibility of core melting and, flexibility of total reactor power. Present concept has simple plant system design without a reactor pressure vessel, ECCS, recirculation systems (of BWR) and others. Therefore construction cost per electric power generation is expected to be slightly low comparing with conventional large scale WRs. (authors)

  19. Partial Safety Analysis for a Reduced Uranium Enrichment Core for the High Flux Isotope Reactor

    SciTech Connect

    Primm, Trent [ORNL; Gehin, Jess C [ORNL

    2009-04-01

    A computational model of the reactor core of the High Flux Isotope Rector (HFIR) was developed in order to analyze non-destructive accidents caused by transients during reactor operation. The reactor model was built for the latest version of the nuclear analysis software package called Program for the Analysis of Reactor Transients (PARET). Analyses performed with the model constructed were compared with previous data obtained with other tools in order to benchmark the code. Finally, the model was used to analyze the behavior of the reactor under transients using a different nuclear fuel with lower enrichment of uranium (LEU) than the fuel currently used, which has a high enrichment of uranium (HEU). The study shows that the presence of fertile isotopes in LEU fuel, which increases the neutron resonance absorption, reduces the impact of transients on the fuel and enhances the negative reactivity feedback, thus, within the limitations of this study, making LEU fuel appear to be a safe alternative fuel for the reactor core.

  20. Full core reactor analysis: Running Denovo on Jaguar

    SciTech Connect

    Jarrell, J. J.; Godfrey, A. T.; Evans, T. M.; Davidson, G. G. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States)

    2012-07-01

    Fully-consistent, full-core, 3D, deterministic neutron transport simulations using the orthogonal mesh code Denovo were run on the massively parallel computing architecture Jaguar XT5. Using energy and spatial parallelization schemes, Denovo was able to efficiently scale to more than 160 k processors. Cell-homogenized cross sections were used with step-characteristics, linear-discontinuous finite element, and trilinear-discontinuous finite element spatial methods. It was determined that using the finite element methods gave considerably more accurate eigenvalue solutions for large-aspect ratio meshes than using step-characteristics. (authors)

  1. Full Core Reactor Analysis: Running Denovo on Jaguar

    SciTech Connect

    Jarrell, Joshua J [ORNL; Godfrey, Andrew T [ORNL; Evans, Thomas M [ORNL; Davidson, Gregory G [ORNL

    2012-01-01

    Fully-consistent, full-core, 3D, deterministic simulations using the orthogonal mesh code Denovo were run on the massively parallel computing architecture Jaguar XT5. Using energy and spatial parallelization schemes, Denovo was able to efficiently scale to over 160k processors. Cell-homogenized cross-sections were used with Step-Characteristics, Linear-Discontinuous Finite Element, and Tri-Linear-Discontinuous Finite Element spatial methods. It was determined that using the finite element methods gave considerably more accurate eigenvalue solutions for large aspect ratios meshes than those using Step-Characteristics.

  2. Full Core Reactor Analysis: Running Denovo on Jaguar

    SciTech Connect

    Jarrell, Joshua J [ORNL] [ORNL; Godfrey, Andrew T [ORNL] [ORNL; Evans, Thomas M [ORNL] [ORNL; Davidson, Gregory G [ORNL] [ORNL

    2013-01-01

    Fully-consistent, full-core, 3D, deterministic neutron transport simulations using the orthogonal mesh code Denovo were run on the massively parallel computing architecture Jaguar XT5. Using energy and spatial parallelization schemes, Denovo was able to efficiently scale to more than 160k processors. Cell-homogenized cross sections were used with step-characteristics, linear-discontinuous finite element, and trilinear-discontinuous finite element spatial methods. It was determined that using the finite element methods gave considerably more accurate eigenvalue solutions for large-aspect ratio meshes than using step-characteristics.

  3. Measurements of the HEU and LEU in-core spectra at the Ford Nuclear Reactor

    SciTech Connect

    Wehe, D.K.; King, J.S.; Lee, J.C.; Martin, W.R.

    1984-01-01

    The Ford Nuclear Reactor (FNR) at the University of Michigan has been serving as the test site for a low-enriched uranium (LEU) fuel whole-core demonstration. As part of the experimental program, the differential neutron spectrum has been measured in a high-enriched uranium (HEU) core and an LEU core. The HEU and LEU spectra were determined by unfolding the measured activities of foils that were irradiated in the reactor. When the HEU and LEU spectra are compared from 1 MeV to 10 MeV, significant differences between the two spectra are apparent below 10 eV. These are probably caused by the additional /sup 238/U resonance absorption in the LEU fuel. No measurable difference occurs in the shape of the spectra above 1 MeV. 7 refs., 6 figs., 2 tabs.

  4. The Detection of Reactor Antineutrinos for Reactor Core Monitoring: an Overview

    NASA Astrophysics Data System (ADS)

    Fallot, M.

    2014-06-01

    There have been new developments in the field of applied neutrino physics during the last decade. The International Atomic Energy Agency (IAEA) has expressed interest in the potentialities of antineutrino detection as a new tool for reactor monitoring and has created an ad hoc Working Group in late 2010 to follow the associated research and development. Several research projects are ongoing around the world to build antineutrino detectors dedicated to reactor monitoring, to search for and develop innovative detection techniques, or to simulate and study the characteristics of the antineutrino emission of actual and innovative nuclear reactor designs. We give, in these proceedings, an overview of the relevant properties of antineutrinos, the possibilities of and limitations on their detection, and the status of the development of a variety of compact antineutrino detectors for reactor monitoring.

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

    SciTech Connect

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

    2013-07-01

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

  6. Split-core heat-pipe reactors for out-of-pile thermionic power systems.

    NASA Technical Reports Server (NTRS)

    Niederauer, G.; Lantz, E.; Breitweiser, R.

    1971-01-01

    Description of the concept of splitting a heat-pipe reactor for out-of-core thermionics into two identical halves and using the resulting center gap for reactivity control. Short Li-W reactor heat pipes penetrate the axial reflectors and form a heat exchanger with long heat pipes which wind through the shield to the thermionic diodes. With one reactor half anchored to the shield, the other is attached to a long arm with a pivot behind the shield and swings through a small arc for reactivity control. A safety shim prevents large reactivity inputs, and a fueled control arm drive shaft acts as a power stabilizer. Reactors fueled with U-235C and with U-233C have been studied.-

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

    SciTech Connect

    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

    2010-09-01

    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.

  8. Validation of the neutron and gamma fields in the JSI TRIGA reactor using in-core fission and ionization chambers.

    PubMed

    Žerovnik, Gašper; Kaiba, Tanja; Radulovi?, Vladimir; Jazbec, Anže; Rupnik, Sebastjan; Barbot, Loïc; Fourmentel, Damien; Snoj, Luka

    2015-02-01

    CEA developed fission chambers and ionization chambers were utilized at the JSI TRIGA reactor to measure neutron and gamma fields. The measured axial fission rate distributions in the reactor core are generally in good agreement with the calculated values using the Monte Carlo model of the reactor thus verifying both the computational model and the fission chambers. In future, multiple absolutely calibrated fission chambers could be used for more accurate online reactor thermal power monitoring. PMID:25479432

  9. Core design of long life-cycle fast reactors operating without reactivity margin

    SciTech Connect

    Aristova, E. N.; Baydin, D. F.; Gol'din, V. Y.; Pestryakova, G. A.; Stoynov, M. I. [Keldysh Inst. of Applied Mathematics RAS, Miusskaya sq., bld.4, 125047, Moscow (Russian Federation)

    2012-07-01

    In this paper we consider a possibility of designing a fast reactor core that operates without reactivity margin for a long time. This study is based on the physical principle of fast reactor operating in a self-adjustable neutron-nuclear regime (SANNR-1) introduced by L.P. Feoktistov (1988-1993) and improved by V. Ya. Gol'din SANNR-2 (1995). The mathematical modeling of active zones of fast reactors in SANNR modes is held by authors since 1992. The numerical simulation is based on solving the neutron transport equation coupled with quasi-diffusion equations. The calculations have been performed using standard 26 energy groups. We use a hierarchy of spatial models of 1D, 1.5D, 2D, and 3D geometries. The spatial models of higher dimensionality are used for verification of results. The calculations showed that operation of the reactor in this mode increases its efficiency, safety and simplifies management. It is possible to achieve continuous work of the reactor in SANNR-2 during 7-10 years without fuel overloads by means of further optimization of the mode. Small reactivity margin is used only for the reactor start up. After first 10-15 days the reactor in SANNR-2 operates without reactivity margin. (authors)

  10. Nuclear design of a vapor core reactor for space nuclear propulsion

    SciTech Connect

    Dugan, E.T.; Watanabe, Y.; Kuras, S.A.; Maya, I.; Diaz, N.J. (Innovative Nuclear Space Power and Propulsion Institute, University of Florida, Gainesville, Florida 32611 (United States))

    1993-01-15

    Neutronic analysis methodology and results are presented for the nuclear design of a vapor core reactor for space nuclear propulsion. The Nuclear Vapor Thermal Reactor (NVTR) Rocket Engine uses modified NERVA geometry and systems which the solid fuel replaced by uranium tetrafluoride vapor. The NVTR is an intermediate term gas core thermal rocket engine with specific impulse in the range of 1000--1200 seconds; a thrust of 75,000 lbs for a hydrogen flow rate of 30 kg/s; average core exit temperatures of 3100 K to 3400 K; and reactor thermal powers of 1400 to 1800 MW. Initial calculations were performed on epithermal NVTRs using ZrC fuel elements. Studies are now directed at thermal NVTRs that use fuel elements made of C-C composite. The large ZrC-moderated reactors resulted in thrust-to-weight ratios of only 1 to 2; the compact C-C composite systems yield thrust-to-weight ratios of 3 to 5.

  11. Nuclear design of a vapor core reactor for space nuclear propulsion

    NASA Astrophysics Data System (ADS)

    Dugan, Edward T.; Watanabe, Yoichi; Kuras, Stephen A.; Maya, Isaac; Diaz, Nils J.

    1993-01-01

    Neutronic analysis methodology and results are presented for the nuclear design of a vapor core reactor for space nuclear propulsion. The Nuclear Vapor Thermal Reactor (NVTR) Rocket Engine uses modified NERVA geometry and systems which the solid fuel replaced by uranium tetrafluoride vapor. The NVTR is an intermediate term gas core thermal rocket engine with specific impulse in the range of 1000-1200 seconds; a thrust of 75,000 lbs for a hydrogen flow rate of 30 kg/s; average core exit temperatures of 3100 K to 3400 K; and reactor thermal powers of 1400 to 1800 MW. Initial calculations were performed on epithermal NVTRs using ZrC fuel elements. Studies are now directed at thermal NVTRs that use fuel elements made of C-C composite. The large ZrC-moderated reactors resulted in thrust-to-weight ratios of only 1 to 2; the compact C-C composite systems yield thrust-to-weight ratios of 3 to 5.

  12. Determination of the kinetic parameters of the CALIBAN metallic core reactor from stochastic neutron measurements

    SciTech Connect

    Casoli, P.; Authier, N.; Chapelle, A. [Commissariat a l'Energie Atomique et Aux Energies Alternatives, CEA, DAM, F-21120 Is sur Tille (France)

    2012-07-01

    Several experimental devices are operated by the Criticality and Neutron Science Research Dept. of the CEA Valduc Laboratory. One of these is the Caliban metallic core reactor. The purpose of this study is to develop and perform experiments allowing to determinate some of fundamental kinetic parameters of the reactor. The prompt neutron decay constant and particularly its value at criticality can be measured with reactor noise techniques such as Rossi-{alpha} and Feynman variance-to-mean methods. Subcritical, critical, and even supercritical experiments were performed. Fission chambers detectors were put nearby the core and measurements were analyzed with the Rossi-{alpha} technique. A new value of the prompt neutron decay constant at criticality was determined, which allows, using the Nelson number method, new evaluations of the effective delayed neutron fraction and the in core neutron lifetime. As an introduction of this paper, some motivations of this work are given in part 1. In part 2, principles of the noise measurements experiments performed at the CEA Valduc Laboratory are reminded. The Caliban reactor is described in part 3. Stochastic neutron measurements analysis techniques used in this study are then presented in part 4. Results of fission chamber experiments are summarized in part 5. Part 6 is devoted to the current work, improvement of the experimental device using He 3 neutron detectors and first results obtained with it. Finally, conclusions and perspectives are given in part 7. (authors)

  13. Advanced Core Design And Fuel Management For Pebble-Bed Reactors

    SciTech Connect

    Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

    2004-10-01

    A method for designing and optimizing recirculating pebble-bed reactor cores is presented. At the heart of the method is a new reactor physics computer code, PEBBED, which accurately and efficiently computes the neutronic and material properties of the asymptotic (equilibrium) fuel cycle. This core state is shown to be unique for a given core geometry, power level, discharge burnup, and fuel circulation policy. Fuel circulation in the pebble-bed can be described in terms of a few well?defined parameters and expressed as a recirculation matrix. The implementation of a few heat?transfer relations suitable for high-temperature gas-cooled reactors allows for the rapid estimation of thermal properties critical for safe operation. Thus, modeling and design optimization of a given pebble-bed core can be performed quickly and efficiently via the manipulation of a limited number key parameters. Automation of the optimization process is achieved by manipulation of these parameters using a genetic algorithm. The end result is an economical, passively safe, proliferation-resistant nuclear power plant.

  14. Core design and safety studies for a small modular fast reactor

    SciTech Connect

    Yang, W. S.; Cahalan, J. E.; Dunn, F. E. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2006-07-01

    The paper describes the core design and performance characteristics and the safety analysis results for a 50 MWe small modular fast reactor design that was developed jointly by ANL, CEA, and JNC as an international collaborative effort. The main goal in the core design was to achieve a 30-year lifetime with no refueling. In order to minimize the burnup reactivity swing, metal fuel with a high heavy metal volume fraction was selected. To enhance the proliferation resistance and actinide transmutation, all the transuranic (TRU) elements recovered from light water reactor spent fuel were used in a ternary alloy form of U-TRU-10Zr. A 125 MWt core design was developed, for which the burnup reactivity swing was only 1.6$ over the 30-year core lifetime. The average discharge burnup was 87 MWd/kg, and the maximum sodium void worth was 4.65$. The evaluated reactivity coefficients provided sufficient negative feedbacks. Shutdown margins of control systems were confirmed. Steady-state thermal-hydraulic analysis results showed that peak 2{sigma} cladding inner-wall and fuel centerline temperatures were less than design limits with sufficient margins. Detailed transient analyses for the total loss of power to reactor and intermediate coolant pumps showed that no fuel damage or cladding failure would occur, even when multiple safety systems were assumed to malfunction. (authors)

  15. Passive Safety Small Reactor for Distributed Energy Supply: Heavy Water Mixing Core

    SciTech Connect

    Ken-ichi Sawada; Naoteru Odano [National Maritime Research Institute, 6-38-1, Shinkawa, Mitaka-shi, Tokyo 181-0004 (Japan); Toshihisa Ishida [Kobe University, Kobe 657-8501 (Japan)

    2006-07-01

    The purpose of this paper is to study the core performance of passive safety small reactor for distributed energy supply by changing the heavy water (D{sub 2}O) concentration in the mixed coolant together with the fuel pitch. The long core life with conditions of the excessive reactivity of 2 %{delta}k/k, the reactivity shutdown margin of 1 %{delta}k/k and the negative coolant temperature reactivity coefficient is attained for the case of D{sub 2}O concentration of 60 % with 10 % enrichment gadolinia (Gd{sub 2}O{sub 3}) doped fuel rods. This D{sub 2}O core has a shorter core life 4.14 years than the original light water (H{sub 2}O) core 4.76 years, while it needs a larger core size. However, changing the D{sub 2}O concentration on the way during the burn-up shows a possibility of extending more the core life than that of the original H{sub 2}O core. (authors)

  16. Results in the application of pattern recognition methods to nuclear reactor core component surveillance

    SciTech Connect

    Gonzalez, R.C.; Fry, D.N.; Kryter, R.C.

    1973-01-01

    From nuclear science symposium; San Francisco, California, USA (14 Nov 1973). Pattern recognition methods were applied to analyze and interpret neutron noise data from the High Flux Isotope Reactor (HFIR) at ORNL. The results show that it is feasible to detect some core component failures by means of machine- discernible differences in the time-dependent noise power spectra. These neutron spectra (signatures) were analyzed by using a clusterseeking algorithm to derive a set of templates for automatic computer evaluation of the reactor's mechanical integrity and soundness. (auth)

  17. The effects of aging on Boiling Water Reactor core isolation cooling system

    SciTech Connect

    Lee, Bom Soon

    1994-06-01

    A study was performed to assess the effects of aging on the Reactor Core Isolation Cooling system in commercial Boiling Water Reactors. This study is part of the Nuclear Plant Aging Research program sponsored by the US Nuclear Regulatory Commission. The failure data, from national databases, as well as plant specific data were reviewed and analyzed to understand the effects of aging on the RCIC system. This analysis identified important components that should receive the highest priority in terms of aging management. The aging characterization provided information on the effects of aging on component failure frequency, failure modes, and failure causes.

  18. Analyzing the thermionic reactor critical experiments. [thermal spectrum of uranium 235 core

    NASA Technical Reports Server (NTRS)

    Niederauer, G. F.

    1973-01-01

    The Thermionic Reactor Critical Experiments (TRCE) consisted of fast spectrum highly enriched U-235 cores reflected by different thicknesses of beryllium or beryllium oxide with a transition zone of stainless steel between the core and reflector. The mixed fast-thermal spectrum at the core reflector interface region poses a difficult neutron transport calculation. Calculations of TRCE using ENDF/B fast spectrum data and GATHER library thermal spectrum data agreed within about 1 percent for the multiplication factor and within 6 to 8 percent for the power peaks. Use of GAM library fast spectrum data yielded larger deviations. The results were obtained from DOT R Theta calculations with leakage cross sections, by region and by group, extracted from DOT RZ calculations. Delineation of the power peaks required extraordinarily fine mesh size at the core reflector interface.

  19. Core design and reactor physics of a breed and burn gas-cooled fast reactor

    E-print Network

    Yarsky, Peter

    2005-01-01

    In order to fulfill the goals set forth by the Generation IV International Forum, the current NERI funded research has focused on the design of a Gas-cooled Fast Reactor (GFR) operating in a Breed and Burnm (B&B) fuel cycle ...

  20. DCT-Based Iris Recognition

    Microsoft Academic Search

    Donald M. Monro; Soumyadip Rakshit; Dexin Zhang

    2007-01-01

    This paper presents a novel iris coding method based on differences of discrete cosine transform (DCT) coefficients of overlapped angular patches from normalized iris images. The feature extraction capabilities of the DCT are optimized on the two largest publicly available iris image data sets, 2,156 images of 308 eyes from the CASIA database and 2,955 images of 150 eyes from

  1. Trends in Iris Recognition Algorithms

    Microsoft Academic Search

    Atul Bansal; Ravinder Agarwal; R. K. Sharma

    2010-01-01

    Authentication of persons using machine has always been a very attractive problem. Biometric systems for authentication based on human characteristics such as face, finger, voice and iris is becoming the prominent research area. Iris recognition among these is considered the most accurate and reliable biometric identification system. Iris recognition system finds application in the various security systems including at airports,

  2. Iris mattress suture: a technique for sectoral iris defect repair.

    PubMed

    Tsao, Sean W; Holz, Huck A

    2015-03-01

    Achieving a cosmetic and functional outcome from iris defect repair is a surgical challenge. We describe an adaptation of techniques to address a case of 2.5 clock hours of sectoral iris tissue defect. Our method combines Siepser's modified closed-chamber sliding knot technique with the placement of a double-armed iris mattress suture to approximate iris tissue to the scleral wall and thereby create a pseudo-iris root. This technique reduces glare and achieves a cosmetic outcome for the patient. PMID:24879808

  3. A new advanced fixed in-core instrumentation for a PWR reactor

    NASA Astrophysics Data System (ADS)

    Barbet, M.; Guillery, M.

    Gamma thermometer studies have been done at E.D.F. for four years. These studies started in France with a feasibility study in 1975. E.D.F.'s scope was to develop a new fixed "in-core" instrumentation for PWR based on the gamma heat measurements. The advanced gamma thermometer design has been done in such a way to be able to manufacture strings of 6 to 9 detectors each. The results of gamma thermometer make up in 1976 were encouraging and E.D.F. went on to develop a gamma thermometer assembly for a reactor application. Before being mounted on the reactor vessel, the gamma thermometer strings are calibrated in a loop test by means of an electrical current giving the ?T versus the specific power ( W/ g). The loop test simulates the thermohydraulic conditions in the reactor tube guide. Two gamma thermometer strings have been installed in the BUGEY 5 reactor since June 1979. Four gamma thermometer strings are provided for insertion in the TRICASTIN 2 reactor and four more gamma thermometer strings are manufactured to be ready for the start up of the TRICASTIN 3 reactor in 1980.

  4. Toward accurate and fast iris segmentation for iris biometrics.

    PubMed

    He, Zhaofeng; Tan, Tieniu; Sun, Zhenan; Qiu, Xianchao

    2009-09-01

    Iris segmentation is an essential module in iris recognition because it defines the effective image region used for subsequent processing such as feature extraction. Traditional iris segmentation methods often involve an exhaustive search of a large parameter space, which is time consuming and sensitive to noise. To address these problems, this paper presents a novel algorithm for accurate and fast iris segmentation. After efficient reflection removal, an Adaboost-cascade iris detector is first built to extract a rough position of the iris center. Edge points of iris boundaries are then detected, and an elastic model named pulling and pushing is established. Under this model, the center and radius of the circular iris boundaries are iteratively refined in a way driven by the restoring forces of Hooke's law. Furthermore, a smoothing spline-based edge fitting scheme is presented to deal with noncircular iris boundaries. After that, eyelids are localized via edge detection followed by curve fitting. The novelty here is the adoption of a rank filter for noise elimination and a histogram filter for tackling the shape irregularity of eyelids. Finally, eyelashes and shadows are detected via a learned prediction model. This model provides an adaptive threshold for eyelash and shadow detection by analyzing the intensity distributions of different iris regions. Experimental results on three challenging iris image databases demonstrate that the proposed algorithm outperforms state-of-the-art methods in both accuracy and speed. PMID:19574626

  5. Experimental Plans for Subsystems of a Shock Wave Driven Gas Core Reactor

    NASA Technical Reports Server (NTRS)

    Kazeminezhad, F.; Anghai, S.

    2008-01-01

    This Contractor Report proposes a number of plans for experiments on subsystems of a shock wave driven pulsed magnetic induction gas core reactor (PMI-GCR, or PMD-GCR pulsed magnet driven gas core reactor). Computer models of shock generation and collision in a large-scale PMI-GCR shock tube have been performed. Based upon the simulation results a number of issues arose that can only be addressed adequately by capturing experimental data on high pressure (approx.1 atmosphere or greater) partial plasma shock wave effects in large bore shock tubes ( 10 cm radius). There are three main subsystems that are of immediate interest (for appraisal of the concept viability). These are (1) the shock generation in a high pressure gas using either a plasma thruster or pulsed high magnetic field, (2) collision of MHD or gas dynamic shocks, their interaction time, and collision pile-up region thickness, and (3) magnetic flux compression power generation (not included here).

  6. A Metal Fuel Core Concept for 1000 MWt Advanced Burner Reactor

    SciTech Connect

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

    2007-07-01

    This paper describes the core design and performance characteristics of a metal fuel core concept for a 1000 MWt Advanced Burner Reactor. A ternary metal fuel form of U-TRU-Zr was assumed with weapons grade plutonium feed for the startup core and TRU recovered from LWR spent fuel for the recycled equilibrium core. A compact burner core was developed by trade-off between the burnup reactivity loss and TRU conversion ratio, with a fixed cycle length of one-year. In the startup core, the average TRU enrichment is 15.5%, the TRU conversion ratio is 0.81, and the burnup reactivity loss over a cycle is 3.6% {delta}k. The heavy metal and TRU inventories are 13.1 and 2.0 metric tons, respectively. The average discharge burnup is 93 MWd/kg, and the TRU consumption rate is 55.5 kg/year. For the recycled equilibrium core, the average TRU enrichment is 22.1 %, the TRU conversion ratio is 0.73, and the burnup reactivity loss is 2.2% {delta}k. The TRU inventory and consumption rate are 2.9 metric tons and 81.6 kg/year, respectively. The evaluated reactivity coefficients provide sufficient negative feedbacks. The control systems provide shutdown margins that are more than adequate. The integral reactivity parameters for quasi-static reactivity balance analysis indicate favorable passive safety features, although detailed safety analyses are required to verify passive safety behavior. (authors)

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

    Microsoft Academic Search

    E. A. Hoffman; W. S. Yang; R. N. Hill

    2008-01-01

    A consistent set of designs for 1000 MWt commercial-scale sodium-cooled Advance Burner Reactors (ABR) have been developed for both metal and oxide-fueled cores with conversion ratios from breakeven (CR=1.0) to fertile-free (CR=0.0). These designs are expected to satisfy thermal and irradiation damage limits based on the currently available data. The very low conversion ratio designs require fuel that is beyond

  8. Optimal control of a coupled-core nuclear reactor by a singular perturbation method

    Microsoft Academic Search

    PARVATHAREDDY B. REDDY; PEDDAPULLAIAH SANNUTI

    1975-01-01

    Optimal control of a two-core coupled nuclear reactor system is considered. The mathematical description of this system leads to an eighth-order nonlinear time delay model. This model is written in such a way that when a scalar parameter is perturbed, it reduces to a second-order model without time delays. Using the recently developed singular perturbation theory, an approximate solution valid

  9. System for monitoring of energy release in the core of a boiling-water reactor

    Microsoft Academic Search

    Yu. I. Leshchenko; V. P. Sadulin; I. I. Semidotskii

    1988-01-01

    Results are discussed from an investigation into a system for the physical monitoring of energy release in the core of the VK-50 boiling water reactor. Movable self-powered detectors are used in this system as energy-release neutron detectors. Rhodium serves as the emitter in these detectors. A number of parameters for these detectors were experimentally measured; they include the ratio of

  10. FORMOSA-B: A Boiling Water Reactor In-Core Fuel Management Optimization Package III

    Microsoft Academic Search

    Atul A. Karve; Paul J. Turinsky

    2001-01-01

    As part of the continuing development of the boiling water reactor in-core fuel management optimization code FORMOSA-B, the cold shutdown margin (SDM) constraint evaluator has been improved. The SDM evaluator in FORMOSA-B had been a first-order accurate Rayleigh quotient variational technique. It was deemed unreliable for difficult perturbed loading patterns (LPs) and thus was replaced by a high-fidelity, robust, computationally

  11. The Annular Core Research Reactor (ACRR) postulated limiting event initial and building source terms

    SciTech Connect

    Restrepo, L F

    1992-08-01

    As part of the update of the Safety analysis Report (SAR) for the Annular Core Research Reactor (ACRR), operational limiting events under the category of inadvertent withdrawal of an experiment while at power or during a power pulse were determined to be the most limiting event(s) for this reactor. This report provides a summary of the assumptions, modeling, and results in evaluation of: Reactivity and thermal hydraulics analysis to determine the amount of fuel melt or fuel damage ratios; The reactor inventories following the limiting event; A literature review of post NUREG-0772 release fraction experiment results on severe fuel damages; Decontamination factors due to in-pool transport; and In-building transport modeling and building source term analysis.

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

    SciTech Connect

    Jeannot, J. P.; Rodriguez, G.; Jammes, C.; Bernardin, B.; Portier, J. L.; Jadot, F. [Commissariat a l'Energie Atomique, Saint-Paul-lez-Durance, 13108 (France); Maire, S.; Verrier, D. [Advanced Projects and Decommissioning Div. Plant Sector AREVA NP - NEPL-FT, Lyon, 69000 (France); Loisy, F. [EDF - EDF R and D STEP Dept., 6 Quai Watier, Chatou, 78401 (France); Prele, G. [EDF, Generation/Nuclear Engineering, Basic Design Dept., Villeurbanne, 69628 (France)

    2011-07-01

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

  13. Optimization of a heterogeneous fast breeder reactor core with improved behavior during unprotected transients

    SciTech Connect

    Poumerouly, S.; Schmitt, D.; Massara, S.; Maliverney, B. [EDF R and D, 1 avenue du general de Gaulle, 92140 Clamart (France)

    2012-07-01

    Innovative Sodium-cooled Fast Reactors (SFRs) are currently being investigated by CEA, AREVA and EDF in the framework of a joint French collaboration, and the construction of a GEN IV prototype, ASTRID (Advanced Sodium Technical Reactor for Industrial Demonstration), is scheduled in the years 2020. Significant improvements are expected so as to improve the reactor safety: the goal is to achieve a robust safety demonstration of the mastering of the consequences of a Core Disruptive Accident (CDA), whether by means of prevention or mitigation features. In this framework, an innovative design was proposed by CEA in 2010. It aims at strongly reducing the sodium void effect, thereby improving the core behavior during unprotected loss of coolant transients. This design is strongly heterogeneous and includes, amongst others, a fertile plate, a sodium plenum associated with a B{sub 4}C upper blanket and a stepwise modulation of the fissile height of the core (onwards referred to as the 'diabolo shape'). In this paper, studies which were entirely carried out at EDF are presented: the full potential of this heterogeneous concept is thoroughly investigated using the SDDS methodology. (authors)

  14. Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

    DOE PAGESBeta

    None

    2014-11-04

    Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental keff come from uncertainties in the manganese content and impurities in the stainless steel fuel cladding asmore »well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9?) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2? uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

  15. Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

    SciTech Connect

    None

    2014-11-04

    Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental keff come from uncertainties in the manganese content and impurities in the stainless steel fuel cladding as well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9?) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2? uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

  16. Benchmark Evaluation of the Neutron Radiography (NRAD) Reactor Upgraded LEU-Fueled Core

    SciTech Connect

    John D. Bess

    2001-09-01

    Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. The final upgraded core configuration with 64 fuel elements has been completed. Evaluated benchmark measurement data include criticality, control-rod worth measurements, shutdown margin, and excess reactivity. Dominant uncertainties in keff include the manganese content and impurities contained within the stainless steel cladding of the fuel and the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 nuclear data are approximately 1.4% greater than the benchmark model eigenvalue, supporting contemporary research regarding errors in the cross section data necessary to simulate TRIGA-type reactors. Uncertainties in reactivity effects measurements are estimated to be ~10% with calculations in agreement with benchmark experiment values within 2s. The completed benchmark evaluation de-tails are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Experiments (IRPhEP Handbook). Evaluation of the NRAD LEU cores containing 56, 60, and 62 fuel elements have also been completed, including analysis of their respective reactivity effects measurements; they are also available in the IRPhEP Handbook but will not be included in this summary paper.

  17. Low-Enriched Fuel Design Concept for the Prismatic Very High Temperature Reactor Core

    SciTech Connect

    Sterbentz, James W

    2007-05-01

    A new non-TRISO fuel and clad design concept is proposed for the prismatic, heliumcooled Very High Temperature Reactor core. The new concept could substantially reduce the current 10-20 wt% TRISO uranium enrichments down to 4-6 wt% for both initial and reload cores. The proposed fuel form would be a high-temperature, high-density uranium ceramic, for example UO2, configured into very small diameter cylindrical rods. The small diameter fuel rods significantly increase core reactivity through improved neutron moderation and fuel lumping. Although a high-temperature clad system for the concept remains to be developed, recent success in tube fabrication and preliminary irradiation testing of silicon carbide (SiC) cladding for light water reactor applications offers good potential for this application, and for future development of other carbide clad designs. A high-temperature ceramic fuel, together with a high-temperature clad material, could also lead to higher thermal safety margins during both normal and transient reactor conditions relative to TRISO fuel. The calculated neutronic results show that the lowenrichment, small diameter fuel rods and low thermal neutron absorbing clad retain the strong negative Doppler fuel temperature coefficient of reactivity that ensures inherent safe operation of the VHTR, and depletion studies demonstrate that an 18-month power cycle can be achieved with the lower enrichment fuel.

  18. Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

    DOE PAGESBeta

    None

    2014-11-04

    Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental keff come from uncertainties in the manganese content and impurities in the stainless steel fuel cladding as well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9?) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2? uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

  19. Checkerboard seed-blanket thorium fuel core concepts for heavy water moderated reactors

    SciTech Connect

    Bromley, B.P.; Hyland, B. [Atomic Energy of Canada Limited, Chalk River Laboratories, 1 Plant Road, Chalk River, Ontario, K0J 1J0 (Canada)

    2013-07-01

    New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen was a 35-element bundle made with a homogeneous mixture of reactor grade Pu (about 67 wt% fissile) and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several checkerboard heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that various checkerboard core concepts can achieve a fissile utilization that is up to 26% higher than that achieved in a PT-HWR using more conventional natural uranium fuel bundles. Up to 60% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 303 kg/year of Pa-233/U-233/U-235 are produced. Checkerboard cores with about 50% of low-power blanket bundles may require power de-rating (65% to 74%) to avoid exceeding maximum limits for channel and bundle powers and linear element ratings. (authors)

  20. Annular seed-blanket thorium fuel core concepts for heavy water moderated reactors

    SciTech Connect

    Bromley, B.P.; Hyland, B. [Atomic Energy of Canada Limited, Chalk River Laboratories, 1 Plant Road, Chalk River, Ontario, K0J 1J0 (Canada)

    2013-07-01

    New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen is a 35-element bundle made with a homogeneous mixture of reactor grade Pu and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several annular heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that the various core concepts can achieve a fissile utilization that is up to 30% higher than is currently achieved in a PT-HWR using conventional natural uranium fuel bundles. Up to 67% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 363 kg/year of U-233 is produced. Seed-blanket cores with ?50% content of low-power blanket bundles may require power de-rating (?58% to 65%) to avoid exceeding maximum limits for peak channel power, bundle power and linear element ratings. (authors)

  1. Lunar Nuclear Power Plant With Solid Core Reactor, Heatpipes and Thermoelectric Conversion

    SciTech Connect

    Sayre, Edwin D. [Engineering Consultant, 218 Brooke Acres Drive, Los Gatos, CA 95032 (United States); Ring, Peter J. [Advanced Methods and Materials, 1190 Mountain View-Alviso Rd. Suite P, Sunnyvale, CA 94089 (United States); Brown, Neil [Engineering Consultant, 5134 Cordoy Lane, San Jose, CA 95124 (United States); Elsner, Norbert B.; Bass, John C. [Hi-Z Technology, Inc., 7606 Miramar Rd. Suite 7400, San Diego, CA 92126 (United States)

    2008-01-21

    This is a lunar nuclear power plant with the advantages of minimum mass, with no moving parts, no pumped liquid coolant, a solid metal rugged core, with no single point of failure. The electrical output is 100 kilowatts with a 500 kilowatt thermal reactor. The thermoelectric converters surround the potassium heatpipes from the core and water heatpipes surround the converter and connect to the radiator. The solid core reactor is made from HT9 alloy. The fuel is uranium oxide with 90% enrichment. The thermoelectric converter is bonded to the outside of the 1.10 inch ID heat pipe and is 30 inches long. The thermoelectric couple is Si/SiGe-Si/SiC Quantum Well with over 20% efficiency with an 890 K hot side and a 490 K cold side and produces 625 Watts. 176 converters produce 110 kWe. With less than 10% loss in controls this yields 100 kWe for use. The cylindrical thermoelectric converter is designed and fabricated by HIPing to keep brittle materials in compression and to ensure conductivity. The solid core is fabricated by machining the heatpipe tubes with 6 grooves that are diffusion bonded together by HIPing to form the fuel tubes. The maximum temperature of the heat pipes is 940 K and the return flow temperature is 890 K. The reactor core is hexagonal shaped, 61 cm. wide and 76.2 cm high with 12 rotating control drums surrounding it. There is shielding to protect components and human habitation. The radiator is daisy shaped at 45 degrees with each petal 5.5 meters long. The design life is ten years.

  2. Iris recognition technology

    Microsoft Academic Search

    G. O. Williams

    1996-01-01

    IriScan Inc. has for the past two years, been developing an identification\\/verification system capable of positively identifying and verifying the identity of individuals without physical contact or a person in the loop. Personal identification has historically been based on what a person possesses (a card); knows (a Personal Identification Number); or is (an inherent physiological or behavioral characteristic). Facial features

  3. Final Report, NERI Project: ''An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model''

    SciTech Connect

    Dmitriy Y. Anistratov; Marvin L. Adams; Todd S. Palmer; Kord S. Smith; Kevin Clarno; Hikaru Hiruta; Razvan Nes

    2003-08-04

    OAK (B204) Final Report, NERI Project: ''An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model'' The present generation of reactor analysis methods uses few-group nodal diffusion approximations to calculate full-core eigenvalues and power distributions. The cross sections, diffusion coefficients, and discontinuity factors (collectively called ''group constants'') in the nodal diffusion equations are parameterized as functions of many variables, ranging from the obvious (temperature, boron concentration, etc.) to the more obscure (spectral index, moderator temperature history, etc.). These group constants, and their variations as functions of the many variables, are calculated by assembly-level transport codes. The current methodology has two main weaknesses that this project addressed. The first weakness is the diffusion approximation in the full-core calculation; this can be significantly inaccurate at interfaces between different assemblies. This project used the nodal diffusion framework to implement nodal quasidiffusion equations, which can capture transport effects to an arbitrary degree of accuracy. The second weakness is in the parameterization of the group constants; current models do not always perform well, especially at interfaces between unlike assemblies. The project developed a theoretical foundation for parameterization and homogenization models and used that theory to devise improved models. The new models were extended to tabulate information that the nodal quasidiffusion equations can use to capture transport effects in full-core calculations.

  4. Investigation on the Core Bypass Flow in a Very High Temperature Reactor

    SciTech Connect

    Hassan, Yassin

    2013-10-22

    Uncertainties associated with the core bypass flow are some of the key issues that directly influence the coolant mass flow distribution and magnitude, and thus the operational core temperature profiles, in the very high-temperature reactor (VHTR). Designers will attempt to configure the core geometry so the core cooling flow rate magnitude and distribution conform to the design values. The objective of this project is to study the bypass flow both experimentally and computationally. Researchers will develop experimental data using state-of-the-art particle image velocimetry in a small test facility. The team will attempt to obtain full field temperature distribution using racks of thermocouples. The experimental data are intended to benchmark computational fluid dynamics (CFD) codes by providing detailed information. These experimental data are urgently needed for validation of the CFD codes. The following are the project tasks: • Construct a small-scale bench-top experiment to resemble the bypass flow between the graphite blocks, varying parameters to address their impact on bypass flow. Wall roughness of the graphite block walls, spacing between the blocks, and temperature of the blocks are some of the parameters to be tested. • Perform CFD to evaluate pre- and post-test calculations and turbulence models, including sensitivity studies to achieve high accuracy. • Develop the state-of-the art large eddy simulation (LES) using appropriate subgrid modeling. • Develop models to be used in systems thermal hydraulics codes to account and estimate the bypass flows. These computer programs include, among others, RELAP3D, MELCOR, GAMMA, and GAS-NET. Actual core bypass flow rate may vary considerably from the design value. Although the uncertainty of the bypass flow rate is not known, some sources have stated that the bypass flow rates in the Fort St. Vrain reactor were between 8 and 25 percent of the total reactor mass flow rate. If bypass flow rates are on the high side, the quantity of cooling flow through the core may be considerably less than the nominal design value, causing some regions of the core to operate at temperatures in excess of the design values. These effects are postulated to lead to localized hot regions in the core that must be considered when evaluating the VHTR operational and accident scenarios.

  5. Preparations to load, transport, receive, and store the damaged TMI-2 (Three Mile Island) reactor core

    SciTech Connect

    Reno, H.W.; Schmitt, R.C.; Quinn, G.J.; Ayers, A.L. Jr.; Lilburn, B.J. Jr.; Uhl, D.L.

    1986-03-01

    The March 1979 incident at the Three Mile Island Nuclear Power Station (TMI) which damaged the core of the Unit 2 reactor resulted in numerous scientific and technical challenges. Some of those challenges involve removing, packaging, and transporting the core debris to the Idaho National Engineering Laboratory (INEL) for storage, examination, and preparation for final disposal. This paper highlights preparations for transporting the core debris from TMI to INEL and receiving and storing that material at INEL. Issues discussed include interfacing of equipment and facilities at TMI, loading operations, transportation activities using a newly designed cask, receiving and storing operations at INEL, and criticality control during storage. Key to the transportation effort was designing, testing, fabricating, and licensing two rail casks which individually provide double containment of the damaged fuel. 27 figs.

  6. Simulated Verification of Fuel Element Inventory in a Small Reactor Core Using the Nuclear Materials Identification System (NMIS)

    SciTech Connect

    Grogan, Brandon R [ORNL; Mihalczo, John T [ORNL

    2009-01-01

    The International Panel on Climate Change projects that by 2050 the world energy demand may double. Although the primary focus for new nuclear power plants in industrialized nations is on large plants in the 1000-1600 MWe range, there is an increasing demand for small and medium reactors (SMRs). About half of the innovative SMR concepts are for small (<300 MWe) reactors with a 5-30 year life without on-site refueling. This type of reactor is also known as a battery-type reactor. These reactors are particularly attractive to countries with small power grids and for non-electrical purposes such as heating, hydrogen production, and seawater desalination. Traditionally, this type of reactor has been used in a nautical propulsion role. This type of reactor is designed as a permanently sealed unit to prevent the diversion of the uranium in the core by the user. However, after initial fabrication it will be necessary to verify that the newly fabricated reactor core contains the quantity of uranium that initially entered the fuel fabrication plant. In most instances, traditional inspection techniques can be used to perform this verification, but in certain situations the core design will be considered sensitive. Non-intrusive verification techniques must be utilized in these situations. The Nuclear Materials Identification System (NMIS) with imaging uses active interrogation and a fast time correlation processor to characterize fissile material. The MCNP-PoliMi computer code was used to simulate NMIS measurements of a small, sealed reactor core. Because most battery-type reactor designs are still in the early design phase, a more traditional design based on a Russian icebreaker core was used in the simulations. These simulations show how the radiography capabilities of the NMIS could be used to detect the diversion of fissile material by detecting void areas in the assembled core where fuel elements have been removed.

  7. Enhancement of the inherent self-protection of the fast sodium reactor cores with oxide fuel

    SciTech Connect

    Eliseev, V.A.; Malisheva, I.V.; Matveev, V.I.; Egorov, A.V.; Maslov, P.A. [SSC RF - IPPE, Obninsk, Kaluga region (Russian Federation)

    2013-07-01

    With the development and research into the generation IV fast sodium reactors, great attention is paid to the enhancement of the core inherent self-protection characteristics. One of the problems dealt here is connected with the reduction of the reactivity margin so that the control rods running should not result in the core overheating and melting. In this paper we consider the possibilities of improving the core of BN-1200 with oxide fuel by a known method of introducing an axial fertile layer into the core. But unlike earlier studies this paper looks at the possibility of using such a layer not only for improving breeding, but also for reducing sodium void reactivity effect (SVRE). This proposed improvement of the BN-1200 core does not solve the problem of strong interference in control and protection system (CPS) rods of BN-1200, but they reduce significantly the reactivity margin for burn-up compensation. This helps compensate all the reactivity balances in the improved core configurations without violating constraints on SVRE value.

  8. THERMIT: a computer program for three-dimensional thermal-hydraulic analysis of light-water-reactor cores. Final report

    Microsoft Academic Search

    J. Loomis; W. H. Reed; A. Schor; H. B. Stewart; L. Wolf

    1981-01-01

    THERMIT is a computer code that solves the time dependent two-fluid thermal-hydraulic equations in three space dimensions along with the fuel pin and clad temperatures. Its purpose is to predict the thermal-hydraulic response of a modeling of a reactor core to situations which do not lead to reconfiguration of the core geometry. Both boiling and pressurized water reactors may be

  9. A point kernel model for the energy deposited on samples from gamma radiation in a research reactor core

    Microsoft Academic Search

    M. Varvayanni; N. Catsaros; M. Antonopoulos-Domis

    2008-01-01

    A basic safety requirement for a research reactor is the reliable estimation of the gamma heating of samples irradiated in the reactor core. A three-dimensional numerical code of gamma heating using a point kernel parameterization is developed. The heating due to ?-rays, produced from U235 fission and from (n,?) reactions with the core materials is considered. The dose build-up due

  10. Design of an overmoderated fuel and a full MOX core for plutonium consumption in boiling water reactors

    Microsoft Academic Search

    Juan Luis François; Cecilia Mart??n del Campo; Joel Hernández

    2002-01-01

    The use of uranium–plutonium mixed oxide fuel (MOX) in light water reactors (LWR) is nowadays a current practice in several countries. Generally 1\\/3 of the reactor core is loaded with MOX fuel assemblies and the other 2\\/3 with uranium assemblies. Nevertheless the plutonium utilization could be more effective if the full core could be loaded with MOX fuel. In this

  11. Safety apparatus for nuclear reactor to prevent structural damage from overheating by core debris

    DOEpatents

    Gabor, John D. (Western Springs, IL); Cassulo, John C. (Stickney, IL); Pedersen, Dean R. (Naperville, IL); Baker, Jr., Louis (Downers Grove, IL)

    1986-01-01

    The invention teaches safety apparatus that can be included in a nuclear reactor, either when newly fabricated or as a retrofit add-on, that will minimize proliferation of structural damage to the reactor in the event the reactor is experiencing an overheating malfunction whereby radioactive nuclear debris might break away from and be discharged from the reactor core. The invention provides a porous bed or sublayer on the lower surface of the reactor containment vessel so that the debris falls on and piles up on the bed. Vapor release elements upstand from the bed in some laterally spaced array. Thus should the high heat flux of the debris interior vaporize the coolant at that location, the vaporized coolant can be vented downwardly to and laterally through the bed to the vapor release elements and in turn via the release elements upwardly through the debris. This minimizes the pressure buildup in the debris and allows for continuing infiltration of the liquid coolant into the debris interior.

  12. Radiation transport in a liquid metal fast breeder rector during a loss of sodium coolant reactor core disassembly

    Microsoft Academic Search

    Rzepecki

    1986-01-01

    The time-dependent, neutron radiation transport equation for a liquid metal fast breeder reactor undergoing a loss of sodium coolant reactor core disassembly accident is analyzed and the models being used evaluated. Time-dependent neutron transport is calculated using both a discrete ordinates and a diffusion theory solution for the real flux shape. It was found that diffusion theory would underpredict reactivity

  13. CORE ANALYSIS, DESIGN AND OPTIMIZATION OF A DEEP-BURN PEBBLE BED REACTOR

    SciTech Connect

    B. Boer; A. M. Ougouag

    2010-05-01

    Achieving a high burnup in the Deep-Burn pebble bed reactor design, while remaining within the limits for fuel temperature, power peaking and temperature reactivity feedback, is challenging. The high content of Pu and Minor Actinides in the Deep-Burn fuel significantly impacts the thermal neutron energy spectrum. This can result in power and temperature peaking in the pebble bed core in locally thermalized regions near the graphite reflectors. Furthermore, the interplay of the Pu resonances of the neutron absorption cross sections at low-lying energies can lead to a positive temperature reactivity coefficient for the graphite moderator at certain operating conditions. To investigate the aforementioned effects a code system using existing codes has been developed for neutronic, thermal-hydraulic and fuel depletion analysis of Deep-Burn pebble bed reactors. A core analysis of a Deep-Burn Pebble Bed Modular Reactor (400 MWth) design has been performed for two Deep-Burn fuel types and possible improvements of the design with regard to power peaking and temperature reactivity feedback are identified.

  14. Operating experience with the multienrichment initial core of the boiling water reactor Kashiwazaki-Kariwa Unit 5

    Microsoft Academic Search

    Takaaki Mochida; Mitsunari Nakamura; Junichi Yamashita; Hiromi Maruyama; Sakae Muto; Shigeru Kasai

    1996-01-01

    The multienrichment boiling water reactor (BWR) initial core design was first applied to the Kashiwazaki-Kariwa Nuclear Power Station Unit 5 [1100-MW (electric) BWR] in Japan. This core is designed to improve fuel discharge exposure, capacity factors, and operability. The design study shows that three types of fuel bundles with different enrichments are suitable to achieve the design targets. Three bundle

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

    SciTech Connect

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

    2010-09-01

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

  16. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    SciTech Connect

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

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

    SciTech Connect

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

    2012-07-01

    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)

  18. Conceptual studies for pressurised water reactor cores employing plutonium erbium zirconium oxide inert matrix fuel assemblies

    NASA Astrophysics Data System (ADS)

    Stanculescu, A.; Kasemeyer, U.; Paratte, J.-M.; Chawla, R.

    1999-08-01

    The most efficient way to enhance plutonium consumption in light water reactors is to eliminate the production of plutonium all together. This requirement leads to fuel concepts in which the uranium is replaced by an inert matrix. At PSI, studies have focused on employing ZrO 2 as inert matrix. Adding a burnable poison to such a fuel proves to be necessary. As a result of scoping studies, Er 2O 3 was identified as the most suitable burnable poison material. The results of whole-core three-dimensional neutronics analyses indicated, for a present-day 1000 MW e pressurised water reactor (PWR), the feasibility of an asymptotic equilibrium four-batch cycle fuelled solely with the proposed PuO 2-Er 2O 3-ZrO 2 inert matrix fuel (IMF). The present paper presents the results of more recent investigations related to `real-life' situations, which call for transition configurations in which mixed IMF and UO 2 assembly loadings must be considered. To determine the influence of the introduction of IMF assemblies on the characteristics of a UO 2-fuelled core, three-dimensional full-core calculations have been performed for a present-day 1000 MW e PWR containing up to 12 optimised IMF assemblies.

  19. Integrated risk information system (IRIS)

    SciTech Connect

    Tuxen, L. [Environmental Protection Agency, Washington, DC (United States)

    1990-12-31

    The Integrated Risk Information System (IRIS) is an electronic information system developed by the US Environmental Protection Agency (EPA) containing information related to health risk assessment. IRIS is the Agency`s primary vehicle for communication of chronic health hazard information that represents Agency consensus following comprehensive review by intra-Agency work groups. The original purpose for developing IRIS was to provide guidance to EPA personnel in making risk management decisions. This original purpose for developing IRIS was to guidance to EPA personnel in making risk management decisions. This role has expanded and evolved with wider access and use of the system. IRIS contains chemical-specific information in summary format for approximately 500 chemicals. IRIS is available to the general public on the National Library of Medicine`s Toxicology Data Network (TOXNET) and on diskettes through the National Technical Information Service (NTIS).

  20. Physics-based multiscale coupling for full core nuclear reactor simulation

    DOE PAGESBeta

    Gaston, Derek R.; Permann, Cody J.; Peterson, John W.; Slaughter, Andrew E.; Andrš, David; Wang, Yaqi; Short, Michael P.; Perez, Danielle M.; Tonks, Michael R.; Ortensi, Javier; et al

    2015-10-01

    Numerical simulation of nuclear reactors is a key technology in the quest for improvements in efficiency, safety, and reliability of both existing and future reactor designs. Historically, simulation of an entire reactor was accomplished by linking together multiple existing codes that each simulated a subset of the relevant multiphysics phenomena. Recent advances in the MOOSE (Multiphysics Object Oriented Simulation Environment) framework have enabled a new approach: multiple domain-specific applications, all built on the same software framework, are efficiently linked to create a cohesive application. This is accomplished with a flexible coupling capability that allows for a variety of different datamore »exchanges to occur simultaneously on high performance parallel computational hardware. Examples based on the KAIST-3A benchmark core, as well as a simplified Westinghouse AP-1000 configuration, demonstrate the power of this new framework for tackling—in a coupled, multiscale manner—crucial reactor phenomena such as CRUD-induced power shift and fuel shuffle. 2014 The Authors. Published by Elsevier Ltd. This is an open access article under the CC BY-NC-SA license« less

  1. Physics-based multiscale coupling for full core nuclear reactor simulation

    SciTech Connect

    Derek R. Gaston; Cody J. Permann; John W. Peterson; Slaughter; David Andrs; Yaqui Wang; Michael P. Short; Danielle M. Perez; Michael R. Tonks; Javier Ortensi; Ling Zou; Richard C. Martineau

    2014-11-01

    Numerical simulation of nuclear reactors is a key technology in the quest for improvements in efficiency, safety, and reliability of both existing and future reactor designs. Historically, simulation of an entire reactor was accomplished by linking together multiple existing codes that each simulated a subset of the relevant multiphysics phenomena. Recent advances in the MOOSE (Multiphysics Object Oriented Simulation Environment) framework have enabled a new approach: multiple domain-specific applications, all built on the same software framework, are efficiently linked to create a cohesive application. This is accomplished with a flexible coupling capability that allows for a variety of different data exchanges to occur simultaneously on high performance parallel computational hardware. Examples based on the KAIST-3A benchmark core, as well as a simplified Westinghouse AP-1000 configuration, demonstrate the power of this new framework for tackling—in a coupled, multiscale manner—crucial reactor phenomena such as CRUD-induced power shift and fuel shuffle. 2014 The Authors. Published by Elsevier Ltd. This is an open access article under the CC BY-NC-SA license

  2. A survey of alternative once-through fast reactor core designs

    SciTech Connect

    Fei, T.; Richard, J. G.; Kersting, A. R.; Don, S. M.; Oi, C.; Driscoll, M. J.; Shwageraus, E. [Nuclear Science and Engineering Dept., Massachusetts Inst. of Technology, 77 Massachusetts Ave., Cambridge, MA 02139 (United States)

    2012-07-01

    Reprocessing of Light Water Reactor (LWR) spent fuel to recover plutonium or transuranics for use in Sodium cooled Fast Reactors (SFRs) is a distant prospect in the U.S.A. This has motivated our evaluation of potentially cost-effective operation of uranium startup fast reactors (USFRs) in a once-through mode. This review goes beyond findings reported earlier based on a UC fueled MgO reflected SFR to describe a broader parametric study of options. Cores were evaluated for a variety of fuel/coolant/reflector combinations: UC/UZr/UO{sub 2}/UN;Na/Pb; MgO/SS/Zr. The challenge is achieving high burnup while minimizing enrichment and respecting both cladding fluence/dpa and reactivity lifetime limits. These parametric studies show that while UC fuel is still the leading contender, UO{sub 2} fuel and ZrH 1.7 moderated metallic fuel are also attractive if UC proves to be otherwise inadequate. Overall, these findings support the conclusion that a competitive fuel cycle cost and uranium utilization compared to LWRs is possible for SFRs operated on a once-through uranium fueled fuel cycle. In addition, eventual transition to TRU recycle mode is studied, as is a small test reactor to demonstrate key features. (authors)

  3. An improved resonance self-shielding method for heterogeneous fast reactor assembly and core calculations

    SciTech Connect

    Lee, C. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439-4842 (United States); Yang, W. S. [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47907-2017 (United States)

    2013-07-01

    An improved resonance self-shielding method has been developed to accurately estimate the effective multigroup cross sections for heterogeneous fast reactor assembly and core calculations. In the method, the heterogeneity effect is considered by the use of isotopic escape cross sections while the resonance interference effect is accounted for through the narrow resonance approximation or slowing-down calculations for specific compositions. The isotopic escape cross sections are calculated by solving fixed-source transport equations with the method of characteristics for the whole problem domain. This method requires no pre-calculated resonance integral tables or parameters that are typically necessary in the subgroup method. Preliminary results for multi pin-cell fast reactor problems show that the escape cross sections estimated from the explicit-geometry fixed source calculations produce more accurate eigenvalue and self-shielded effective cross sections than those from conventional one-dimensional geometry models. (authors)

  4. Performance of metal and oxide fuel cores during accidents in large liquid-metal-cooled reactors

    SciTech Connect

    Royl, P.H.; Kussmaul, G. (Kernforschungszentrum Karlsruhe GmbH (Germany)); Cahalan, J.E.; Wigeland, R.A. (Argonne National Lab., IL (United States)); Friedel, G. (Interatom, W-5060 Bergisch-Gladbach 1 (DE)); Moreau, J. (Centre d'Etude Nucleaires de Cadarache, 13108 Saint-Paul-lez-Durance (FR)); Perks, M. (AEA-Technology, Risley, Warrington, Cheshire WA3 6AT (GB))

    1992-02-01

    This paper reports on a cooperative effort among European and U.S. analysts, which is an assessment of the comparative safety performance of metal and oxide fuels during accidents in a 3500-MW (thermal), pool-type, liquid-metal-cooled reactor (LMR) is performed. The study focuses on three accident initiators with failure to scram: the unprotected loss-of-flow (ULOF), the unprotected transient overpower, and the unprotected loss-of-heat-sink (ULOHS). Core designs with a similar power output that have been previously analyzed in Europe under ULOF accident conditions are also included in this comparison. Emphasis is placed on identification of design features that provide passive, self-limiting responses to postulated accident conditions and quantification of relative safety margins. The analyses show that in ULOF and ULOHS sequences, metal-fueled LMRs with pool-type primary systems provide larger temperature margins to coolant boiling than do oxide-fueled reactors of the same design.

  5. Monte Carlo Neutronics and Thermal Hydraulics Analysis of Reactor Cores with Multilevel Grids

    NASA Astrophysics Data System (ADS)

    Bernnat, W.; Mattes, M.; Guilliard, N.; Lapins, J.; Zwermann, W.; Pasichnyk, I.; Velkov, K.

    2014-06-01

    Power reactors are composed of assemblies with fuel pin lattices or other repeated structures with several grid levels, which can be modeled in detail by Monte Carlo neutronics codes such as MCNP6 using corresponding lattice options, even for large cores. Except for fresh cores at beginning of life, there is a varying material distribution due to burnup in the different fuel pins. Additionally, for power states the fuel and moderator temperatures and moderator densities vary according to the power distribution and cooling conditions. Therefore, a coupling of the neutronics code with a thermal hydraulics code is necessary. Depending on the level of detail of the analysis, a very large number of cells with different materials and temperatures must be regarded. The assignment of different material properties to all elements of a multilevel grid is very elaborate and may exceed program limits if the standard input procedure is used. Therefore, an internal assignment is used which overrides uniform input parameters. The temperature dependency of continuous energy cross sections, probability tables for the unresolved resonance region and thermal neutron scattering laws is taken into account by interpolation, requiring only a limited number of data sets generated for different temperatures. The method is applied with MCNP6 and proven for several full core reactor models. For the coupling of MCNP6 with thermal hydraulics appropriate interfaces were developed for the GRS system code ATHLET for liquid coolant and the IKE thermal hydraulics code ATTICA-3D for gaseous coolant. Examples will be shown for different applications for PWRs with square and hexagonal lattices, fast reactors (SFR) with hexagonal lattices and HTRs with pebble bed and prismatic lattices.

  6. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    SciTech Connect

    Armstrong, J.; Hamilton, H.; Hyland, B. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)

    2013-07-01

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  7. RELAP5 model for advanced neutron source reactor thermal-hydraulic transients, three-element-core design

    SciTech Connect

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L.

    1996-02-01

    In order to utilize reduced enrichment fuel, the three-element-core design has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. However, the total flow rate through the core is greater and the pressure drop across the core is less so that the primary coolant pumps and heat exchangers are operating at a different point in their performance curves. This report describes the new RELAP5 input for the core components.

  8. Advanced neutron source reactor conceptual safety analysis report, three-element-core design: Chapter 15, accident analysis

    SciTech Connect

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L.; Harrington, R.M.

    1996-02-01

    In order to utilize reduced enrichment fuel, the three-element-core design for the Advanced Neutron Source has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. To assess the impact of changes in the core region configuration and the thermal-hydraulic steady-state conditions, the safety analysis has been updated. This report gives the safety margins for the loss-of-off-site power and pressure-boundary fault accidents based on the RELAP5 results. AU margins are greater for the three-element-core simulations than those calculated for the two-element core.

  9. Iris mammillations: significance and associations.

    PubMed

    Ragge, N K; Acheson, J; Murphree, A L

    1996-01-01

    Iris mammillations are rarely described, distinctive villiform protuberances that can cover the iris. In the majority of reported cases they are unilateral and sporadic, and are seen in association with oculodermal melanosis. In past literature and current clinical practice they are frequently confused with the iris nodules seen in neurofibromatosis type 1. Their clinical significance is not established, although it has been suggested that iris mammillations may be an external sign of ocular hypertension or intraocular malignancy. We report a series of 9 patients between the ages of 3 and 28 years with iris mammillations. The mammillations appear as regularly spaced, deep brown, smooth, conical elevations on the iris, of uniform height or increasing in height as the pupil margin is approached. They often overlie a naevus or an exceptionally deeply pigmented iris, such as that seen in melanosis oculi. One case had an associated ciliary body mass. They tend to occur in more highly pigmented ethnic groups and can be dominantly inherited. Iris mammillations may occur in association with systemic conditions including phakomatosis pigmentovascularis type IIb and neurofibromatosis type 1 when they may even coexist with iris hamartomas. PMID:8763309

  10. BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III

    SciTech Connect

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.

    1981-06-01

    This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given.

  11. Wireless, in-vessel neutron monitor for initial core-loading of advanced breeder reactors

    NASA Technical Reports Server (NTRS)

    Delorenzo, J. T.; Kennedy, E. J.; Blalock, T. V.; Rochelle, J. M.; Chiles, M. M.; Valentine, K. H.

    1981-01-01

    An experimental wireless, in-vessel neutron monitor was developed to measure the reactivity of an advanced breeder reactor as the core is loaded for the first time to preclude an accidental critically incident. The environment is liquid sodium at a temperature of approx. 220 C, with negligible gamma or neutron radiation. With ultrasonic transmission of neutron data, no fundamental limitation was observed after tests at 230 C for 2000 h. The neutron sensitivity was approx. 1 count/s-nv, and the potential data transmission rate was approx. 10,000 counts/s.

  12. COREMAP: Graphical user interface for displaying reactor core data in an interactive hexagon map

    SciTech Connect

    Muscat, F.L.; Derstine, K.L.

    1995-06-01

    COREMAP is a Graphical User Interface (GUI) designed to assist users read and check reactor core data from multidimensional neutronic simulation models in color and/or as text in an interactive 2D planar grid of hexagonal subassemblies. COREMAP is a complete GEODST/RUNDESC viewing tool which enables the user to access multi data set files (e.g. planes, moments, energy groups ,... ) and display up to two data sets simultaneously, one as color and the other as text. The user (1) controls color scale characteristics such as type (linear or logarithmic) and range limits, (2) controls the text display based upon conditional statements on data spelling, and value. (3) chooses zoom features such as core map size, number of rings and surrounding subassemblies, and (4) specifies the data selection for supplied popup subwindows which display a selection of data currently off-screen for a selected cell, as a list of data and/or as a graph. COREMAP includes a RUNDESC file editing tool which creates ``proposed`` Run-description files by point and click revisions to subassembly assignments in an existing EBRII Run-description file. COREMAP includes a fully automated printing option which creates high quality PostScript color or greyscale images of the core map independent of the monitor used, e.g. color prints can be generated with a session from a color or monochrome monitor. The automated PostScript output is an alternative to the xgrabsc based printing option. COREMAP includes a plotting option which creates graphs related to a selected cell. The user specifies the X and Y coordinates types (planes, moment, group, flux ,... ) and a parameter, P, when displaying several curves for the specified (X, Y) pair COREMAP supports hexagonal geometry reactor core configurations specified by: the GEODST file and binary Standard Interface Files and the RUNDESC ordering.

  13. Core and System Design of Reduced-Moderation Water Reactor with Passive Safety Features

    SciTech Connect

    Iwamura, Takamichi; Okubo, Tsutomu; Yonomoto, Taisuke [Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Ibaraki-ken, 319-1195 (Japan); Takeda, Renzo; Moriya, Kumiaki [Hitachi, Ltd. (Japan); Kanno, Minoru [The Japan Atomic Power Company (Japan)

    2002-07-01

    In order to ensure the sustainable energy supply in Japan, research and developments of reduced-moderation water reactor (RMWR) have been performed. The RMWR can attain the favorable characteristics such as high burn-up, long operation cycle, multiple recycling of plutonium and effective utilization of uranium resources, based on the matured LWR technologies. MOX fuel assemblies in the tight-lattice fuel rod arrangement are used to reduce the moderation of neutron, and hence, to increase the conversion ratio. The conceptual design has been accomplished for the small 330 MWe RMWR core with the discharge burn-up of 60 GWd/t and the operation cycle of 24 months, under the natural circulation cooling of the core. A breeding ratio of 1.01 and the negative void reactivity coefficient are simultaneously realized in the design. In the plant system design, the passive safety features are intended to be utilized mainly to improve the economy. At present, a hybrid one under the combination of the passive and the active components, and a fully passive one are proposed. The former has been evaluated to reduce the cost for the reactor components. (authors)

  14. Hybrid parallel code acceleration methods in full-core reactor physics calculations

    SciTech Connect

    Courau, T.; Plagne, L.; Ponicot, A. [EDF R and D, 1, Avenue du General de Gaulle, 92141 Clamart Cedex (France); Sjoden, G. [Nuclear and Radiological Engineering, Georgia Inst. of Technology, Atlanta, GA 30332 (United States)

    2012-07-01

    When dealing with nuclear reactor calculation schemes, the need for three dimensional (3D) transport-based reference solutions is essential for both validation and optimization purposes. Considering a benchmark problem, this work investigates the potential of discrete ordinates (Sn) transport methods applied to 3D pressurized water reactor (PWR) full-core calculations. First, the benchmark problem is described. It involves a pin-by-pin description of a 3D PWR first core, and uses a 8-group cross-section library prepared with the DRAGON cell code. Then, a convergence analysis is performed using the PENTRAN parallel Sn Cartesian code. It discusses the spatial refinement and the associated angular quadrature required to properly describe the problem physics. It also shows that initializing the Sn solution with the EDF SPN solver COCAGNE reduces the number of iterations required to converge by nearly a factor of 6. Using a best estimate model, PENTRAN results are then compared to multigroup Monte Carlo results obtained with the MCNP5 code. Good consistency is observed between the two methods (Sn and Monte Carlo), with discrepancies that are less than 25 pcm for the k{sub eff}, and less than 2.1% and 1.6% for the flux at the pin-cell level and for the pin-power distribution, respectively. (authors)

  15. Design of a Resistively Heated Thermal Hydraulic Simulator for Nuclear Rocket Reactor Cores

    NASA Technical Reports Server (NTRS)

    Litchford, Ron J.; Foote, John P.; Ramachandran, Narayanan; Wang, Ten-See; Anghaie, Samim

    2007-01-01

    A preliminary design study is presented for a non-nuclear test facility which uses ohmic heating to replicate the thermal hydraulic characteristics of solid core nuclear reactor fuel element passages. The basis for this testing capability is a recently commissioned nuclear thermal rocket environments simulator, which uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce high-temperature pressurized hydrogen flows representative of reactor core environments, excepting radiation effects. Initially, the baseline test fixture for this non-nuclear environments simulator was configured for long duration hot hydrogen exposure of small cylindrical material specimens as a low cost means of evaluating material compatibility. It became evident, however, that additional functionality enhancements were needed to permit a critical examination of thermal hydraulic effects in fuel element passages. Thus, a design configuration was conceived whereby a short tubular material specimen, representing a fuel element passage segment, is surrounded by a backside resistive tungsten heater element and mounted within a self-contained module that inserts directly into the baseline test fixture assembly. With this configuration, it becomes possible to create an inward directed radial thermal gradient within the tubular material specimen such that the wall-to-gas heat flux characteristics of a typical fuel element passage are effectively simulated. The results of a preliminary engineering study for this innovative concept are fully summarized, including high-fidelity multi-physics thermal hydraulic simulations and detailed design features.

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

    SciTech Connect

    David W. Nigg

    2013-09-01

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

  17. High temperature UF6 RF plasma experiments applicable to uranium plasma core reactors

    NASA Technical Reports Server (NTRS)

    Roman, W. C.

    1979-01-01

    An investigation was conducted using a 1.2 MW RF induction heater facility to aid in developing the technology necessary for designing a self critical fissioning uranium plasma core reactor. Pure, high temperature uranium hexafluoride (UF6) was injected into an argon fluid mechanically confined, steady state, RF heated plasma while employing different exhaust systems and diagnostic techniques to simulate and investigate some potential characteristics of uranium plasma core nuclear reactors. The development of techniques and equipment for fluid mechanical confinement of RF heated uranium plasmas with a high density of uranium vapor within the plasma, while simultaneously minimizing deposition of uranium and uranium compounds on the test chamber peripheral wall, endwall surfaces, and primary exhaust ducts, is discussed. The material tests and handling techniques suitable for use with high temperature, high pressure, gaseous UF6 are described and the development of complementary diagnostic instrumentation and measurement techniques to characterize the uranium plasma, effluent exhaust gases, and residue deposited on the test chamber and exhaust system components is reported.

  18. MORE ON IRIS YELLOW SPOT

    Technology Transfer Automated Retrieval System (TEKTRAN)

    Iris yellow spot, caused by Iris yellow spot tospovirus, is an emerging disease of onion in the U.S. and world. Yield losses vary, but may range from undetectable to nearly 100% in onion seed crops. This article presents recent advances in understanding the etiology, epidemiology, and management o...

  19. IRIS Process (2004-2008)

    EPA Science Inventory

    The Integrated Risk Information System (IRIS) is a U.S. Environmental Protection Agency (EPA) database that contains information on human health effects that may result from exposure to chemical substances in the environment. IRIS is maintained by EPA?s Office of Research and Dev...

  20. Georgia Institute of Technology research on the Gas Core Actinide Transmutation Reactor (GCATR)

    NASA Technical Reports Server (NTRS)

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

    1976-01-01

    The program reviewed is a study of the feasibility, design, and optimization of the GCATR. The program is designed to take advantage of initial results and to continue work carried out on the Gas Core Breeder Reactor. The program complements NASA's program of developing UF6 fueled cavity reactors for power, nuclear pumped lasers, and other advanced technology applications. The program comprises: (1) General Studies--Parametric survey calculations performed to examine the effects of reactor spectrum and flux level on the actinide transmutation for GCATR conditions. The sensitivity of the results to neutron cross sections are to be assessed. Specifically, the parametric calculations of the actinide transmutation are to include the mass, isotope composition, fission and capture rates, reactivity effects, and neutron activity of recycled actinides. (2) GCATR Design Studies--This task is a major thrust of the proposed research program. Several subtasks are considered: optimization criteria studies of the blanket and fuel reprocessing, the actinide insertion and recirculation system, and the system integration. A brief review of the background of the GCATR and ongoing research is presented.

  1. [Complications after cosmetic iris implantation].

    PubMed

    Jonsson, N J; Sahlmüller, M C; Ruokonen, P C; Torun, N; Rieck, P

    2011-05-01

    We report the case of a 37-year-old patient with ocular complications associated with the implantation of cosmetic iris implants. Implantation of silicone iris implants for the purpose of changing iris colour has been performed since 2004. Diaphragms are implanted in the anterior chamber. Up to now only little information exists about side effects of this method. In the literature severe ocular complications shortly after cosmetic iris implantation are reported in single cases. In our case 5 months after surgery optic nerve damage caused by elevated intraocular pressure (IOP) was diagnosed. Nuclear opacity of both lenses and a decreased number of corneal endothelial cells were observed at the first visit. Because of recurrent IOP elevation despite maximum antiglaucoma therapy, explantation of the iris implants was required. Damage to the trabecular meshwork, opacity of the lenses as well as the reduced number of endothelial cells are permanent and will probably lead to further complications like corneal decompensation and progressing glaucoma. PMID:21344246

  2. Detailed analyses of key phenomena in core disruptive accidents of sodium-cooled fast reactors by the COMPASS code

    Microsoft Academic Search

    Koji Morita; Shuai Zhang; Seiichi Koshizuka; Yoshiharu Tobita; Hidemasa Yamano; Noriyuki Shirakawa; Fusao Inoue; Hiroaki Yugo; Masanori Naitoh; Hidetoshi Okada; Yuichi Yamamoto; Masashi Himi; Etsujo Hirano; Sensuke Shimizu; Masaya Oue

    A five-year research project has been initiated in 2005 to develop a code based on the MPS (Moving Particle Semi-implicit) method for detailed analysis of key phenomena in core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). The code is named COMPASS (Computer Code with Moving Particle Semi-implicit for Reactor Safety Analysis). The key phenomena include (1) fuel pin failure

  3. Noise Reduction of a Three-Phase Reactor by Optimization of Gaps Between Cores Considering Electromagnetism and Magnetostriction

    Microsoft Academic Search

    Yanhui Gao; Masahide Nagata; Kazuhiro Muramatsu; Koji Fujiwara; Yoshiyuki Ishihara; Shigemasa Fukuchi

    2011-01-01

    To reduce the noise of a reactor connected to an inverter power supply, we have already proposed an improved one-phase reactor, in which the hard materials are chosen as the insulators inserted in the gaps between the cores, and gaps are added in the yokes, using a three-dimensional (3-D) coupled magnetic field and vibration analyses taking account of both the

  4. VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 3. Programmer's manual. Final report. [PWR; BWR

    SciTech Connect

    Stewart, C.W.; Koontz, A.S.; Cuta, J.M.; Montgomery, S.D.

    1983-05-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear-reactor-core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This is Volume 3, the Programmer's Manual. It explains the codes' structures and the computer interfaces.

  5. VIPRE-01. a thermal-hydraulic analysis code for reactor cores. Volume 1. Mathematical modeling. [PWR; BWR

    SciTech Connect

    Stewart, C.W.; Cuta, J.M.; Koontz, A.S.; Kelly, J.M.; Basehore, K.L.; George, T.L.; Rowe, D.S.

    1983-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 1: Mathematical Modeling) explains the major thermal hydraulic models and supporting correlations in detail.

  6. Reactor moderator, pressure vessel, and heat rejection system of an open-cycle gas core nuclear rocket concept

    NASA Technical Reports Server (NTRS)

    Taylor, M. F.; Whitmarsh, C. L., Jr.; Sirocky, P. J., Jr.; Iwanczyke, L. C.

    1973-01-01

    A preliminary design study of a conceptual 6000-megawatt open-cycle gas-core nuclear rocket engine system was made. The engine has a thrust of 196,600 newtons (44,200 lb) and a specific impulse of 4400 seconds. The nuclear fuel is uranium-235 and the propellant is hydrogen. Critical fuel mass was calculated for several reactor configurations. Major components of the reactor (reflector, pressure vessel, and waste heat rejection system) were considered conceptually and were sized.

  7. Mass flow rate sensitivity and uncertainty analysis in natural circulation boiling water reactor core from Monte Carlo simulations

    Microsoft Academic Search

    Gilberto Espinosa-Paredes; Surendra P. Verma; Alejandro Vázquez-Rodríguez; Alejandro Nuñez-Carrera

    2010-01-01

    Our aim was to evaluate the sensitivity and uncertainty of mass flow rate in the core on the performance of natural circulation boiling water reactor (NCBWR). This analysis was carried out through Monte Carlo simulations of sizes up to 40,000, and the size, i.e., repetition of 25,000 was considered as valid for routine applications. A simplified boiling water reactor (SBWR)

  8. Simultaneous measurement of neutron and gamma-ray radiation levels from a TRIGA reactor core using silicon carbide semiconductor detectors

    Microsoft Academic Search

    A. R. Dulloo; F. H. Ruddy; J. G. Seidel; C. Davison; T. Flinchbaugh; T. Daubenspeck

    1998-01-01

    The ability of a SiC detector to measure neutron and gamma radiation levels in a TRIGA reactor's mixed neutron\\/gamma field was demonstrated. Linear responses to an epicadmium neutron fluence rate (up to 3×107 cm-2 s-1) and to a gamma dose rate (0.6-234 krad-Si h-1) were obtained with the detector. Axial profiles of the reactor core's neutron and gamma-ray radiation levels

  9. Hurricane Iris Hits Belize

    NASA Technical Reports Server (NTRS)

    2002-01-01

    Hurricane Iris hit the small Central American country of Belize around midnight on October 8, 2001. At the time, Iris was the strongest Atlantic hurricane of the season, with sustained winds up to 225 kilometers per hour (140 mph). The hurricane caused severe damage-destroying homes, flooding streets, and leveling trees-in coastal towns south of Belize City. In addition, a boat of American recreational scuba divers docked along the coast was capsized by the storm, leaving 20 of the 28 passengers missing. Within hours the winds had subsided to only 56 kph (35 mph), a modest tropical depression, but Mexico, Guatemala, El Salvador, and Honduras were still expecting heavy rains. The above image is a combination of visible and thermal infrared data (for clouds) acquired by a NOAA Geostationary Operational Environmental Satellite (GOES-8) on October 8, 2001, at 2:45 p.m., and the Moderate-resolution Imaging Spectroradiometer (MODIS) (for the color of the ground). The three-dimensional view is from the south-southeast (north is towards the upper left). Belize is off the image to the left. Image courtesy Marit Jentoft-Nilsen, NASA GSFC Visualization Analysis Lab

  10. Natural Nuclear Reactor Oklo and Variation of Fundamental Constants Part 1: Computation of Neutronics of Fresh Core

    E-print Network

    Yu. V. Petrov; A. I. Nazarov; M. S. Onegin; V. Yu. Petrov; E. G. Sakhnovsky

    2005-09-15

    Using modern methods of reactor physics we have performed full-scale calculations of the natural reactor Oklo. For reliability we have used recent version of two Monte Carlo codes: Russian code MCU REA and world wide known code MCNP (USA). Both codes produce similar results. We have constructed a computer model of the reactor Oklo zone RZ2 which takes into account all details of design and composition. The calculations were performed for three fresh cores with different uranium contents. Multiplication factors, reactivities and neutron fluxes were calculated. We have estimated also the temperature and void effects for the fresh core. As would be expected, we have found for the fresh core a significant difference between reactor and Maxwell spectra, which was used before for averaging cross sections in the Oklo reactor. The averaged cross section of Sm-149 and its dependence on the shift of resonance position (due to variation of fundamental constants) are significantly different from previous results. Contrary to results of some previous papers we find no evidence for the change of the fine structure constant in the past and obtain new, most accurate limits on its variation with time: -4 10^{-17}year^{-1} < d alpha/dt/alpha < 3 10^{-17} year^{-1} A further improvement in the accuracy of the limits can be achieved by taking account of the core burnup. These calculations are in progress.

  11. SAS4A: A computer model for the analysis of hypothetical core disruptive accidents in liquid metal reactors

    SciTech Connect

    Tentner, A.M.; Birgersson, G.; Cahalan, J.E.; Dunn, F.E.; Kalimullah; Miles, K.J.

    1987-01-01

    To ensure that the public health and safety are protected under any accident conditions in a Liquid Metal Fast Breeder Reactor (LMFBR), many accidents are analyzed for their potential consequences. The SAS4A code system, described in this paper, provides such an analysis capability, including the ability to analyze low probability events such as the Hypothetical Core Disruptive Accidents (HCDAs). The SAS4A code system has been designed to simulate all the events that occur in a LMFBR core during the initiating phase of a Hypothetical Core Disruptive Accident. During such postulated accident scenarios as the Loss-of-Flow and Transient Overpower events, a large number of interrelated physical phenomena occur during a relatively short time. These phenomena include transient heat transfer and hydrodynamic events, coolant boiling and fuel and cladding melting and relocation. During to the strong neutronic feedback present in a nuclear reactor, these events can significantly influence the reactor power. The SAS4A code system is used in the safety analysis of nuclear reactors, in order to estimate the energetic potential of very low probability accidents. The results of SAS4A simulations are also used by reactor designers in order to build safer reactors and eliminate the possibility of any accident which could endanger the public safety.

  12. Personal Identification Based on Iris Texture Analysis

    Microsoft Academic Search

    Li Ma; Tieniu Tan; Yunhong Wang; Dexin Zhang

    2003-01-01

    Abstract - With an increasing emphasis on security, automated personal identification based on biometrics has been receiving extensive attention over the past decade Iris recognition, as an emerging biometric recognition approach, is becoming a very active topic in both research and practical applications In general, a typical iris recognition system includes iris imaging, iris liveness detection, and recognition This paper

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

    SciTech Connect

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

    2012-09-01

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

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

    SciTech Connect

    David W. Nigg; Devin A. Steuhm

    2011-09-01

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

  15. Flowing gas, non-nuclear experiments on the gas core reactor

    NASA Technical Reports Server (NTRS)

    Kunze, J. F.; Cooper, C. G.; Macbeth, P. J.

    1973-01-01

    Variations in cavity wall and injection configurations of the gas core reactor were aimed at establishing flow patterns that give a maximum of the nuclear criticality eigenvalue. Correlation with the nuclear effect was made using multigroup diffusion theory normalized by previous benchmark critical experiments. Air was used to simulate the hydrogen propellant in the flow tests, and smoked air, argon, or Freon to simulate the central nuclear fuel gas. Tests were run both in the down-firing and upfiring directions. Results showed that acceptable flow patterns with volume fraction for the simulated nuclear fuel gas and high flow rate ratios of propellant to fuel can be obtained. Using a point injector for the fuel, good flow patterns are obtained by directing the outer gas at high velocity long the cavity wall, using louvered injection schemes. Recirculation patterns were needed to stabilize the heavy central gas when different gases are used.

  16. Core design study of a supercritical light water reactor with double row fuel rods

    SciTech Connect

    Zhao, C.; Wu, H.; Cao, L.; Zheng, Y. [School of Nuclear Science and Technology, Xi'an Jiaotong Univ., No. 28, Xianning West Road, Xi'an, ShannXi, 710049 (China); Yang, J.; Zhang, Y. [China Nuclear Power Technology Research Inst., Yitian Road, ShenZhen, GuangDong, 518026 (China)

    2012-07-01

    An equilibrium core for supercritical light water reactor has been designed. A novel type of fuel assembly with dual rows of fuel rods between water rods is chosen and optimized to get more uniform assembly power distributions. Stainless steel is used for fuel rod cladding and structural material. Honeycomb structure filled with thermal isolation is introduced to reduce the usage of stainless steel and to keep moderator temperature below the pseudo critical temperature. Water flow scheme with ascending coolant flow in inner regions is carried out to achieve high outlet temperature. In order to enhance coolant outlet temperature, the radial power distributions needs to be as flat as possible through operation cycle. Fuel loading pattern and control rod pattern are optimized to flatten power distribution at inner regions. Axial fuel enrichment is divided into three parts to control axial power peak, which affects maximum cladding surface temperature. (authors)

  17. The Getty Iris

    NSDL National Science Digital Library

    Launched in 2010, the Iris is the Getty's online magazine, written by staff, volunteers, scholars, interns and other specialists at the Getty's Los Angeles campuses. The goal of this publication is "to offer news, stories, and discoveries about art, conservation, research, and philanthropy and to provide an entertaining and substantive behind-the-scenes look at the inner workings of the Getty." On the homepage, visitors can look through nine different sections, including Behind the Scenes, Art, Conservation, Publications, Research, and Voices. This last area features first-person perspectives from members of the Getty community. These posts include observations on conservation science, graffiti art, and pointed pieces like "Does Text Still Matter?" Lastly, the Philanthropy area features thoughts on the Getty's work through strategic arts education initiatives.

  18. Bartus Iris biometrics

    SciTech Connect

    Johnston, R.; Grace, W.

    1996-07-01

    This is the final report of a one-year, Laboratory-Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). We won a 1994 R&D 100 Award for inventing the Bartas Iris Verification System. The system has been delivered to a sponsor and is no longer available to us. This technology can verify the identity of a person for purposes of access control, national security, law enforcement, forensics, counter-terrorism, and medical, financial, or scholastic records. The technique is non-invasive, psychologically acceptable, works in real-time, and obtains more biometric data than any other biometric except DNA analysis. This project sought to develop a new, second-generation prototype instrument.

  19. Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 39, No. 11, p. 11691181 (November 2002) Conceptual Design of a Modular Island Core Fast Breeder Reactor "RAPID-M"

    E-print Network

    Laughlin, Robert B.

    Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 39, No. 11, p. 1169­1181 (November 2002) Conceptual Design of a Modular Island Core Fast Breeder Reactor "RAPID-M" Mitsuru KAMBE Central Research Institute and accepted September 10, 2002) A metal fueled modular island core sodium cooled fast breeder reactor concept

  20. The design and installation of a core discharge monitor for CANDU-type reactors

    SciTech Connect

    Halbig, J.K. (Los Alamos National Lab., NM (USA)); Monticone, A.C.; Ksiezak, L. (International Atomic Energy Agency, Vienna (Austria)); Smiltnieks, V. (International Atomic Energy Agency, Toronto, ON (Canada). Regional Office)

    1990-01-01

    A new type of surveillance systems that monitors neutron and gamma radiation in a reactor containment is being installed at the Ontario Hydro Darlington Nuclear Generating Station A, Unit 2. Unlike video or film surveillance that monitors mechanical motion, this system measures fuel-specific radiation emanating from irradiated fuel as it is pushed from the core of CANDU-type reactors. Proof-of-principle measurements have been carried out at Bruce Nuclear Generating Station A, Unit 3. The system uses ({gamma},n) threshold detectors and ionization detectors. A microprocessor-based electronics package, GRAND-II (Gamma Ray and Neutron Detector electronics package), provides detector bias, preamplifier power, and signal processing. Firmware in the GRAND-2 controls the surveillance activities, including data acquisition and a level of detector authentication, and it handles authenticated communication with a central data logging computer. Data from the GRAND-II are transferred to an MS-DOS-compatible computer and stored. These data are collected and reviewed for fuel-specific radiation signatures from the primary detector and proper ratios of signals from secondary detectors. 5 figs.

  1. Preliminary fracture analysis of the core pressure boundary tube for the Advanced Neutron Source Research Reactor

    SciTech Connect

    Schulz, K.C. [Univ. of Turabo, Gurabo, Puerto (Puerto Rico). College of Engineering; Yahr, G.T. [Oak Ridge National Lab., TN (United States)

    1995-08-01

    The outer core pressure boundary tube (CPBT) of the Advanced neutron Source (ANS) reactor being designed at Oak Ridge National Laboratory is currently specified as being composed of 6061-T6 aluminum. ASME Boiler and Pressure Vessel Code fracture analysis rules for nuclear components are based on the use of ferritic steels; the expressions, tables, charts and equations were all developed from tests and analyses conducted for ferritic steels. Because of the nature of the Code, design with thin aluminum requires analytical approaches that do not directly follow the Code. The intent of this report is to present a methodology comparable to the ASME Code for ensuring the prevention of nonductile fracture of the CPBT in the ANS reactor. 6061-T6 aluminum is known to be a relatively brittle material; the linear elastic fracture mechanics (LEFM) approach is utilized to determine allowable flaw sizes for the CPBT. A J-analysis following the procedure developed by the Electric Power Research Institute was conducted as a check; the results matched those for the LEFM analysis for the cases analyzed. Since 6061-T6 is known to embrittle when irradiated, the reduction in K{sub Q} due to irradiation is considered in the analysis. In anticipation of probable requirements regarding maximum allowable flaw size, a survey of nondestructive inspection capabilities is also presented. A discussion of probabilistic fracture mechanics approaches, principally Monte Carlo techniques, is included in this report as an introduction to what quantifying the probability of nonductile failure of the CPBT may entail.

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

    SciTech Connect

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

    2006-07-31

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

  3. Shape Adaptive, Robust Iris Feature Extraction from Noisy Iris Images

    PubMed Central

    Ghodrati, Hamed; Dehghani, Mohammad Javad; Danyali, Habibolah

    2013-01-01

    In the current iris recognition systems, noise removing step is only used to detect noisy parts of the iris region and features extracted from there will be excluded in matching step. Whereas depending on the filter structure used in feature extraction, the noisy parts may influence relevant features. To the best of our knowledge, the effect of noise factors on feature extraction has not been considered in the previous works. This paper investigates the effect of shape adaptive wavelet transform and shape adaptive Gabor-wavelet for feature extraction on the iris recognition performance. In addition, an effective noise-removing approach is proposed in this paper. The contribution is to detect eyelashes and reflections by calculating appropriate thresholds by a procedure called statistical decision making. The eyelids are segmented by parabolic Hough transform in normalized iris image to decrease computational burden through omitting rotation term. The iris is localized by an accurate and fast algorithm based on coarse-to-fine strategy. The principle of mask code generation is to assign the noisy bits in an iris code in order to exclude them in matching step is presented in details. An experimental result shows that by using the shape adaptive Gabor-wavelet technique there is an improvement on the accuracy of recognition rate. PMID:24696801

  4. GPU Based General-Purpose Parallel computing to Solve Nuclear Reactor In-Core fuel Management Design and Operation Problem

    SciTech Connect

    Prayudhatama, D.; Waris, A.; Kurniasih, N.; Kurniadi, R. [Bosscha Laboratory, Department of Physics, Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia)

    2010-06-22

    In-core fuel management study is a crucial activity in nuclear power plant design and operation. Its common problem is to find an optimum arrangement of fuel assemblies inside the reactor core. Main objective for this activity is to reduce the cost of generating electricity, which can be done by altering several physical properties of the nuclear reactor without violating any of the constraints imposed by operational and safety considerations. This research try to address the problem of nuclear fuel arrangement problem, which is, leads to the multi-objective optimization problem. However, the calculation of the reactor core physical properties itself is a heavy computation, which became obstacle in solving the optimization problem by using genetic algorithm optimization.This research tends to address that problem by using the emerging General Purpose Computation on Graphics Processing Units (GPGPU) techniques implemented by C language for CUDA (Compute Unified Device Architecture) parallel programming. By using this parallel programming technique, we develop parallelized nuclear reactor fitness calculation, which is involving numerical finite difference computation. This paper describes current prototype of the parallel algorithm code we have developed on CUDA, that performs one hundreds finite difference calculation for nuclear reactor fitness evaluation in parallel by using GPU G9 Hardware Series developed by NVIDIA.

  5. Analysis of in-core experiment activities for the MIT Research Reactor using the ORIGEN computer code

    E-print Network

    Helvenston, Edward M. (Edward March)

    2006-01-01

    The objective of this study is to devise a method for utilizing the ORIGEN-S computer code to calculate the activation products generated in in-core experimental assemblies at the MIT Research Reactor (MITR-II). ORIGEN-S ...

  6. Inrush behaviour of a plunger core reactor with parallel winding paths obtained from field-circuit coupled finite element analyses

    Microsoft Academic Search

    Erich Schmidt; Peter Hamberger

    2008-01-01

    The inrush operational behaviour of a plunger core reactor with four parallel connected winding paths is calculated from voltage driven finite element analyses. The 3D finite element model utilizes a direct circuit coupling to take into account the the parallel windings paths and the applied voltage of the mains. Consequently, the distribution of the coil currents within the parallel connected

  7. Parameters of a Plunger Core Reactor with Parallel Winding Paths Obtained from Voltage Driven Finite Element Analyses

    Microsoft Academic Search

    Erich Schmidt; Peter Hamberger

    2006-01-01

    The steady-state operational parameters of a plunger core reactor with four parallel connected winding paths are calculated from voltage driven finite element analyses. Thereby, the 3D finite element model utilizes a direct circuit coupling to take into account for the applied sinusoidal voltage and the parallel windings paths. Consequently, the distribution of the coil currents is obtained with the full

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

    SciTech Connect

    Varivtsev, A. V., E-mail: vav3@niiar.ru; Zhemkov, I. Yu. [JSC “SSC RIAR,” Dimitrovgrad-10 (Russian Federation)

    2014-12-15

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

  9. The fuzzy clearing approach for a niching genetic algorithm applied to a nuclear reactor core design optimization problem

    Microsoft Academic Search

    Wagner F. Sacco; Marcelo D. Machado; Cláudio M. N. A. Pereira; Roberto Schirru

    2004-01-01

    This article extends previous efforts on genetic algorithms (GAs) applied to a core design optimization problem. We introduce the application of a new Niching Genetic Algorithm (NGA) to this problem and compare its performance to these previous works. The optimization problem consists in adjusting several reactor cell parameters, such as dimensions, enrichment and materials, in order to minimize the average

  10. Design of a boiling water reactor core based on an integrated blanket–seed thorium–uranium concept

    Microsoft Academic Search

    Alejandro Núñez-Carrera; Juan Luis François; Cecilia Martín-del-Campo; Gilberto Espinosa-Paredes

    2005-01-01

    This paper is concerned with the design of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material in an integrated blanket–seed (BS) assembly. The integrated BS concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned out in a once-through cycle. The idea behind the

  11. Predicted impact of core flow rate on the corrosion mitigation effectiveness of hydrogen water chemistry for Kuosheng boiling water reactor

    Microsoft Academic Search

    Mei-Ya Wang; Tsung-Kuang Yeh; Charles F. Chu; Ching Chang

    2009-01-01

    It is currently a common practice that a boiling water reactor (BWR) adopts hydrogen water chemistry (HWC) for mitigating corrosion in structural components in its primary coolant circuit. When the core flow rate (CFR) in a BWR is changed, the coolant residence time in the primary coolant circuit would be different. The concentrations of major redox species (i.e. hydrogen, oxygen,

  12. Load follow simulation of three-dimensional boiling water reactor core by PACS32 parallel microprocessor system

    Microsoft Academic Search

    T. Hoshino; T. Shirakawa

    1982-01-01

    The three-dimensional boiling water reactor (BWR) core following the daily load was simulated by the use of the processor array for continuum simulation (PACS-32), a newly developed parallel microprocessor system. The PACS system consists of 32 processing units (PUs) (microprocessors) and has a multiinstruction, multidata type architecture, being optimum to the numerical simulation of the partial differential equations. The BWR

  13. Kinematic dynamo action in a network of screw motions; application to the core of a fast breeder reactor

    Microsoft Academic Search

    F. Plunian; P. Marty; A. Alemany

    1999-01-01

    Most of the studies concerning the dynamo effect are motivated by astrophysical and geophysical applications. The dynamo effect is also the subject of some experimental studies in fast breeder reactors (FBR) for they contain liquid sodium in motion with magnetic Reynolds numbers larger than unity. In this paper, we are concerned with the flow of sodium inside the core of

  14. Development of Inspection Modality for Shell Weld of Core Support Structure of a Fast Breeder Reactor Using Civa

    Microsoft Academic Search

    Chelamchala Babu Rao; Govind Kumar Sharma; Tammana Jayakumar; Philippe Benoist; Baldev Raj

    2010-01-01

    The core support structure is welded with a 40 mm thick base plate of the main vessel, of 500 MWe Prototype Fast Breeder Reactor, Kalpakkam, India, along the circumference. This `shell weld' situated at a distance of about 435 mm away from the weld overlay, is in-accessible to contact mode ultrasonic testing during in-service inspection. An unconventional ultrasonic methodology was

  15. Bi-content Gadolinia as Burnable Absorber in PWR to Improve the Reactor Core Behaviour

    SciTech Connect

    Zheng, S. [AREVA, AREVA NP Fuel Sector, 10, Rue Juliette Recamier 69456 Lyon cedex (France)

    2007-07-01

    The gadolinia product is one of the standard burnable absorbers used in the PWR long and low leakage fuel cycle in order to control the radial power distribution and to hold down the initial core reactivity. This product presents a large number of advantages such as the high efficiency with only a small number of gadolinia-bearing rods, the easy adjustment between the number and the content of the gadolinia-bearing rods according to the cycle length need and the initial reactivity hold-down, no increasing of boron concentration versus cycle depletion, no additional increasing of internal pressure in poisoned rods, very low additional manufacture cost. On the other hand, some unfavourable phenomena are also observed during the utilization of the gadolinia: amplification of the asymmetrical power distribution and more negative axial offset. Based on the correlation between the gadolinia burnout and its content, the use of gadolinia bi-content will improve the parameters indicated here above. The gadolinia bi-content have been used in BWR for more than 20 years. In this paper, the comparison of the main reactor core physical parameters in PWR, calculated with the AREVA NP standard neutronic code package SCIENCE, is made by using the mono- and bi-content of the gadolinia products in the same fuel assembly. The results show that the asymmetrical axial and azimuthal power distribution can be improved in the case of the bi-content gadolinia product. (authors)

  16. A Computational Fluid Dynamic and Heat Transfer Model for Gaseous Core and Gas Cooled Space Power and Propulsion Reactors

    NASA Technical Reports Server (NTRS)

    Anghaie, S.; Chen, G.

    1996-01-01

    A computational model based on the axisymmetric, thin-layer Navier-Stokes equations is developed to predict the convective, radiation and conductive heat transfer in high temperature space nuclear reactors. An implicit-explicit, finite volume, MacCormack method in conjunction with the Gauss-Seidel line iteration procedure is utilized to solve the thermal and fluid governing equations. Simulation of coolant and propellant flows in these reactors involves the subsonic and supersonic flows of hydrogen, helium and uranium tetrafluoride under variable boundary conditions. An enthalpy-rebalancing scheme is developed and implemented to enhance and accelerate the rate of convergence when a wall heat flux boundary condition is used. The model also incorporated the Baldwin and Lomax two-layer algebraic turbulence scheme for the calculation of the turbulent kinetic energy and eddy diffusivity of energy. The Rosseland diffusion approximation is used to simulate the radiative energy transfer in the optically thick environment of gas core reactors. The computational model is benchmarked with experimental data on flow separation angle and drag force acting on a suspended sphere in a cylindrical tube. The heat transfer is validated by comparing the computed results with the standard heat transfer correlations predictions. The model is used to simulate flow and heat transfer under a variety of design conditions. The effect of internal heat generation on the heat transfer in the gas core reactors is examined for a variety of power densities, 100 W/cc, 500 W/cc and 1000 W/cc. The maximum temperature, corresponding with the heat generation rates, are 2150 K, 2750 K and 3550 K, respectively. This analysis shows that the maximum temperature is strongly dependent on the value of heat generation rate. It also indicates that a heat generation rate higher than 1000 W/cc is necessary to maintain the gas temperature at about 3500 K, which is typical design temperature required to achieve high efficiency in the gas core reactors. The model is also used to predict the convective and radiation heat fluxes for the gas core reactors. The maximum value of heat flux occurs at the exit of the reactor core. Radiation heat flux increases with higher wall temperature. This behavior is due to the fact that the radiative heat flux is strongly dependent on wall temperature. This study also found that at temperature close to 3500 K the radiative heat flux is comparable with the convective heat flux in a uranium fluoride failed gas core reactor.

  17. INTEGRATED RISK INFORMATION SYSTEM (IRIS)

    EPA Science Inventory

    The Integrated Risk Information System (IRIS), prepared and maintained by the U.S. Environmental Protection Agency (U.S. EPA), is an electronic data base containing information on human health effects that may result from exposure to vario...

  18. Nonlinear analysis of nuclear coupled density wave instability in time domain for a boiling water reactor core undergoing core-wide and regional modes of oscillations

    Microsoft Academic Search

    Goutam Dutta; Jagdeep B. Doshi

    2009-01-01

    The objective of the paper is to develop a nuclear coupled thermal-hydraulic model in order to simulate core-wide (in-phase) and regional (out-of-phase) stability analysis in time domain within the limitation of desktop research facility for a boiling water reactor subjected to operational transients. The integrated numerical tool, which is a combination of thermal-hydraulic, neutronic and fuel heat conduction models, is

  19. Reliability comparison of computer based core temperature monitoring system with two and three thermocouples per sub-assembly for Fast Breeder Reactors

    Microsoft Academic Search

    R. Dheenadhayalan; M. Sakthivel; A. J. Arul; K. Madhusoodanan; P. Mohanakrishnan

    2010-01-01

    Prototype Fast Breeder Reactor (PFBR) is a mixed oxide fuelled, sodium cooled, 500 MWe, pool type fast breeder reactor under construction at Kalpakkam, India. The reactor core consists of fuel pins assembled in a number of hexagonal shaped, vertically stacked SubAssemblies (SA). Sodium flows from the bottom of the SAs, takes heat from the fission reaction, comes out through the

  20. Monte Carlo Calculation of Core Reactivity and Fluxes for the Development of the BNCT Neutron Source at the Kyiv Research Reactor

    Microsoft Academic Search

    Olena Gritzay; Oleksandr Kalchenko; Nataliya Klimova; Volodymyr Razbudey; Andriy Sanzhur; Stephen Binney

    2005-01-01

    The presented results show our consecutive steps in developing a neutron source with parameters required by Boron Neutron Capture Therapy (BNCT) at the Kyiv Research Reactor (KRR). The main goal of this work was to analyze the influence of installation of different types of uranium converters close to the reactor core on neutron beam characteristics and on level of reactor

  1. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    SciTech Connect

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2014-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  2. Computational investigations of the dynamical characteristics of the core of a periodic pulse reactor in a system with cascade neutron multiplication

    Microsoft Academic Search

    A. V. Gulevich; P. P. D’yachenko; O. F. Kukharchuk; Yu. I. Likhachev; D. V. Razumovskii; D. A. Rogov; E. N. Kravchenko; O. G. Fokina

    2004-01-01

    The core dynamics of a fast reactor in a cascade reactor system operating in a periodic-pulse regime are examined. A model of a BN-600 fuel element is used as a computational model. Computational studies of the neutron kinetics processes in a fast rector-subcritical assembly system and the thermal dynamics of a fuel element in the core of a periodic-pulse reactor

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

    SciTech Connect

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

    2008-05-05

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

  4. Analysis of core damage frequency due to external events at the DOE (Department of Energy) N-Reactor

    SciTech Connect

    Lambright, J.A.; Bohn, M.P.; Daniel, S.L. (Sandia National Labs., Albuquerque, NM (USA)); Baxter, J.T. (Westinghouse Hanford Co., Richland, WA (USA)); Johnson, J.J.; Ravindra, M.K.; Hashimoto, P.O.; Mraz, M.J.; Tong, W.H.; Conoscente, J.P. (EQE, Inc., San Francisco, CA (USA)); Brosseau, D.A. (ERCE, Inc., Albuquerque, NM (USA))

    1990-11-01

    A complete external events probabilistic risk assessment has been performed for the N-Reactor power plant, making full use of all insights gained during the past ten years' developments in risk assessment methodologies. A detailed screening analysis was performed which showed that all external events had negligible contribution to core damage frequency except fires, seismic events, and external flooding. A limited scope analysis of the external flooding risk indicated that it is not a major risk contributor. Detailed analyses of the fire and seismic risks resulted in total (mean) core damage frequencies of 1.96E-5 and 4.60E-05 per reactor year, respectively. Detailed uncertainty analyses were performed for both fire and seismic risks. These results show that the core damage frequency profile for these events is comparable to that found for existing commercial power plants if proposed fixes are completed as part of the restart program. 108 refs., 85 figs., 80 tabs.

  5. VIPRE (Versatile Internals and Component Program for Reactors; EPRI)-01: A thermal-hydraulic code for reactor cores: Volume 4, Applications: Final report

    SciTech Connect

    Cuta, J.M.; Stewart, C.W.; Koontz, A.S.; Montgomery, S.D.

    1987-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 4: Applications) contains extensive comparisons of VIPRE calculations to experimental data. There are also sensitivity studies and evaluations of code numerical and computational performance. In addition, calculations performed by member utilities using VIPRE for comparisons with transient CHF data, and FSAR plant analyses are presented. Comparisons are also presented of plant thermal-hydraulic calculations with VIPRE and other COBRA codes. These calculations demonstrate the suitability of VIPRE for PWR core thermal-hydraulic analysis.

  6. Mini-cavity plasma core reactors for dual-mode space nuclear power/propulsion systems. M.S. Thesis

    NASA Technical Reports Server (NTRS)

    Chow, S.

    1976-01-01

    A mini-cavity plasma core reactor is investigated for potential use in a dual-mode space power and propulsion system. In the propulsive mode, hydrogen propellant is injected radially inward through the reactor solid regions and into the cavity. The propellant is heated by both solid driver fuel elements surrounding the cavity and uranium plasma before it is exhausted out the nozzle. The propellant only removes a fraction of the driver power, the remainder is transferred by a coolant fluid to a power conversion system, which incorporates a radiator for heat rejection. Neutronic feasibility of dual mode operation and smaller reactor sizes than those previously investigated are shown to be possible. A heat transfer analysis of one such reactor shows that the dual-mode concept is applicable when power generation mode thermal power levels are within the same order of magnitude as direct thrust mode thermal power levels.

  7. Predictions of Displacement Damage and Count Rate for SiC Detectors in IRIS

    Microsoft Academic Search

    B. Khorsandi; T. E. Blue; D. Miller; J. Kulisek; B. Lohan

    2006-01-01

    Silicon carbide (SiC) semiconductor diode detectors may be useful as neutron power monitors in the International Reactor Innovative and Secure (IRIS) nuclear reactor, due to their very high band-gap (which allows high temperature operation and also mitigates many of the effects of radiation damage) and their small volume (which allows high fluence rate operation and detector redundancy). Pulse mode operation

  8. Design of a boiling water reactor equilibrium core using thorium-uranium fuel

    SciTech Connect

    Francois, J-L.; Nunez-Carrera, A.; Espinosa-Paredes, G.; Martin-del-Campo, C.

    2004-10-06

    In this paper the design of a Boiling Water Reactor (BWR) equilibrium core using thorium is presented; a heterogeneous blanket-seed core arrangement concept was adopted. The design was developed in three steps: in the first step two different assemblies were designed based on the integrated blanket-seed concept, they are the blanket-dummy assembly and the blanket-seed assembly. The integrated blanketseed concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned-out in a once-through cycle. In the second step, a core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average 235U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the assembly. In the third step an in-house code was developed to evaluate the thorium equilibrium core under transient conditions. A stability analysis was also performed. Regarding the stability analysis, five operational states were analyzed; four of them define the traditional instability region corner of the power-flow map and the fifth one is the operational state for the full power condition. The frequency and the boiling length were calculated for each operational state. The frequency of the analyzed operational states was similar to that reported for BWRs; these are close to the unstable region that occurs due to the density wave oscillation phenomena in some nuclear power plants. Four transient analyses were also performed: manual SCRAM, recirculation pumps trip, main steam isolation valves closure and loss of feed water. The results of these transients are similar to those obtained with the traditional UO2 nuclear fuel.

  9. Comparison of oxide- and metal-core behavior during CRBRP (Clinch River Breeder Reactor Plant) station blackout

    SciTech Connect

    Polkinghorne, S T; Atkinson, S A

    1986-01-01

    A resurrected concept that could significantly improve the inherently safe response of Liquid-Metal cooled Reactors (LMRs) during severe undercooling transients is the use of metallic fuel. Analytical studies have been reported on for the transient behavior of metal-fuel cores in innovative, inherently safe LMR designs. This paper reports on an analysis done, instead, for the Clinch River Breeder Reactor Plant (CRBRP) design with the only innovative change being the incorporation of a metal-fuel core. The SSC-L code was used to simulate a protected station blackout accident in the CRBRP with a 943 MWt Integral Fast Reactor (IFR) metal-fuel core. The results, compared with those for the oxide-fueled CRBRP, show that the margin to boiling is greater for the IFR core. However, the cooldown transient is more severe due to the faster thermal response time of metallic fuel. Some additional calculations to assess possible LMR design improvements (reduced primary system pressure losses, extended flow coastdown) are also discussed. 8 refs., 13 figs., 2 tabs.

  10. VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 2. User's manual. [PWR; BWR

    SciTech Connect

    Cuta, J.M.; Koontz, A.S.; Stewart, C.W.; Montgomery, S.D.

    1983-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear energy reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 2: User's Manual) describes the input requirements of VIPRE and its auxiliary programs, SPECSET, ASP and DECCON, and lists the input instructions for each code.

  11. Detecting a Contacts Lens During Iris Recognition, page 1 Detecting a Contact Lens During Iris Recognition

    E-print Network

    Barrett, William A.

    Detecting a Contacts Lens During Iris Recognition, page 1 Detecting a Contact Lens During Iris to a high security area. #12;Detecting a Contacts Lens During Iris Recognition, page 2 False recognition image capture. · Patterned, or cosmetic, contact lenses will interfere with the iris pattern analysis

  12. IRIS TOXICOLOGICAL REVIEW OF ACROLEIN (2003 Final)

    EPA Science Inventory

    EPA is announcing the release of the final report, Toxicological Review of Acrolein: in support of the Integrated Risk Information System (IRIS) . The updated Summary for Acrolein and accompanying Quickview have also been added to the IRIS Database....

  13. IRIS Toxicological Review of Hexachloroethane (2011 Final)

    EPA Science Inventory

    EPA is announcing the release of the final report Toxicological Review of Hexachloroethane: in support of the Integrated Risk Information System (IRIS). The updated Summary for Acrylamide and accompanying Quickview have also been added to the IRIS database. ...

  14. IRIS TOXICOLOGICAL REVIEW OF PHOSGENE (2006 Final)

    EPA Science Inventory

    EPA is announcing the release of the final report, Toxicological Review of Phosgene: in support of the Integrated Risk Information System (IRIS) . The updated Summary for Phosgene and accompanying Quickview have also been added to the IRIS Database. ...

  15. Image Quality Assessment for Iris Biometric

    Microsoft Academic Search

    N. D. Kalka; V. Dorairaj; Y. N. Shah; N. A. Schmid; B. Cukic

    Iris recognition, the ability to recognize and distinguish individuals by their iris pattern, is the most reliable biometric in terms of recognition and identification performance. However, performance of these systems is affected by poor quality imaging. In this work, we extend previous research efforts on iris quality assessment by analyzing the effect of seven quality factors: defocus blur, motion blur,

  16. SECURE IRIS VERIFICATION Shenglin Yang1

    E-print Network

    Verbauwhede, Ingrid

    . Index Terms-- iris, personal verification, data security 1. INTRODUCTION Biometric verification provides the biometrics with modern cryptographic techniques [3]. 2. RELATED WORK Beginning from 1987, automatic irisSECURE IRIS VERIFICATION Shenglin Yang1 and Ingrid Verbauwhede1,2 1 Department of Electrical

  17. TE10 resonant iris with angular alignment

    E-print Network

    Bornemann, Jens

    TE10 resonant iris with angular alignment TE101 mode cavities TM110 mode cavities TE01 (TE10) resonant iris 1a 1b Fig. 1: Filter configurations utilizing cavity and iris resonances. Resonant irises and resonant irises. Two different configurations, which allow precise control of the direct couplings between

  18. Non-linear classification for iris patterns

    Microsoft Academic Search

    S V Sheela; P A Vijaya

    2011-01-01

    Biometric authentication based on iris patterns is used for personal identification. Important attributes to identity applications include accuracy, speed and template size. Iris patterns are segmented by considering the maximum area of the connected components in the binary images. The iris region is decomposed into subregions. Hu moments are applied to the minimum variance subregions (MVS). The summation of the

  19. A novel approach for iris recognition

    Microsoft Academic Search

    U. M. Chaskar; M. S. Sutaone

    2010-01-01

    In comparison with other biometric traits, iris recognition systems have many advantages. Iris is a protected internal organ whose random texture is stable throughout life, it can serve as a kind of living password that one need not remember but one always carries along. Since the degree of freedom of iris textures is extremely high, the probability of finding two

  20. A Potential IRI Based Phishing Strategy

    Microsoft Academic Search

    Anthony Y. Fu; Xiaotie Deng; Wenyin Liu

    2005-01-01

    We anticipate a potential phishing strategy by obfuscation of Web links using Internationalized Resource Identifier (IRI). In the IRI scheme, the glyphs of many characters look very similar while their Unicodes are different. Hence, certain different IRIs may show high similarity. The potential phishing attacks based on this strategy are very likely to happen in the near future with the

  1. Fission Product Impact Reduction via Protracted In-core Retention in Very High Temperature Reactor (VHTR) Transmutation Scenarios 

    E-print Network

    Alajo, Ayodeji Babatunde

    2011-08-08

    Studies of Texas A&M University in partial fulfillment of the requirements for the degree of DOCTOR OF PHILOSOPHY May 2010 Major Subject: Nuclear Engineering FISSION PRODUCT IMPACT REDUCTION VIA PROTRACTED IN...-CORE RETENTION IN VERY HIGH TEMPERATURE REACTOR (VHTR) TRANSMUTATION SCENARIOS A Dissertation by AYODEJI BABATUNDE ALAJO Submitted to the Office of Graduate Studies of Texas A&M University in partial fulfillment of the requirements...

  2. First 3-D calculation of core disruptive accident in a large-scale sodium-cooled fast reactor

    Microsoft Academic Search

    Hidemasa Yamano; Yoshiharu Tobita; Satoshi Fujita; Werner Maschek

    2009-01-01

    The SIMMER-IV computer code is a three-dimensional fluid-dynamics code coupled with a fuel-pin model and a space- and energy-dependent neutron transport kinetics model. The present study has attempted the first application of SIMMER-IV to a core disruptive accident in a large-scale sodium-cooled fast reactor. A principal point of this study was to investigate reactivity effects with fuel relocation under three-dimensional

  3. Theoretical and experimental analysis of the bistable convection loop for liquid-metal-cooled reactor emergency core cooling

    Microsoft Academic Search

    G. Anand; R. N. Christensen

    1992-01-01

    In this paper, an emergency core cooling system incorporating a bistable convection loop (BCL) for current passive liquid-metal-cooled reactors is proposed. The system has two stable operating modes. During the off mode, the system is in a pure conduction mode and transfers very little heat. In the on mode, the system switches to the low-resistance configuration of a closed natural

  4. Pressurized water reactor in-core nuclear fuel management by tabu search

    E-print Network

    Hill, Natasha J.; Parks, Geoffrey T.

    2014-08-24

    Optimization of the arrangement of fuel assemblies and burnable poisons when reloading pressurized water reactors has, in the past, been performed with many di erent algorithms in an attempt to make reactors more economic and fuel effi cient...

  5. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual report for FY 2009

    SciTech Connect

    Chandler, David [ORNL; Freels, James D [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Primm, Trent [ORNL; Sease, John D [ORNL; Guida, Tracey [University of Pittsburgh; Jolly, Brian C [ORNL

    2010-02-01

    This report documents progress made during FY 2009 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Studies are reported of the application of a silicon coating to surrogates for spheres of uranium-molybdenum alloy. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. A description of the progress in developing a finite element thermal hydraulics model of the LEU core is provided.

  6. Transient Temperature and Pressure in the Reactor Room During a Core Meltdown Accident

    SciTech Connect

    Shadday, M.A.

    2001-07-17

    The purpose of this numerical model is to determine the optimum ventilation exhaust flow rate for the reactor room. The influence of steam produced in the reactor vessel, on the reactor room pressures, is included in the model. A parametric study of the affect of various steam mass flow rates is included in this document. The affect of steam on the conditions in the reactor room is significant at modest flow rates.

  7. Iris: The VAO SED Application

    NASA Astrophysics Data System (ADS)

    Doe, S.; Bonaventura, N.; Busko, I.; D'Abrusco, R.; Cresitello-Dittmar, M.; Ebert, R.; Evans, J.; Laurino, O.; McDowell, J.; Pevunova, O.; Refsdal, B.

    2012-09-01

    We present Iris, the VAO (Virtual Astronomical Observatory) application for analyzing SEDs (spectral energy distributions). Iris is the result of one of the major science initiatives of the VAO, and the first version was released in September 2011. Iris seamlessly combines key features of several existing astronomical software applications to streamline and enhance the SED analysis process. With Iris, users may read in and display SEDs, select data ranges for analysis, fit models to SEDs, and calculate confidence limits on best-fit parameters. SED data may be uploaded into the application from IVOA-compliant VOTable and FITS format files, or retrieved directly from NED (the NASA/IPAC Extragalactic Database). Data written in unsupported formats may be converted for upload using SedImporter, a new application provided with the package. The components of Iris have been contributed by members of the VAO. Specview, contributed by STScI (the Space Telescope Science Institute), provides a GUI for reading, editing, and displaying SEDs, as well as defining model expressions and setting initial model parameter values. Sherpa, contributed by the Chandra project at SAO (the Smithsonian Astrophysical Observatory), provides a library of models, fit statistics, and optimization methods for analyzing SEDs; the underlying I/O library, SEDLib, is a VAO product written by SAO to current IVOA (International Virtual Observatory Alliance) data model standards. NED is a service provided by IPAC (the Infrared Processing and Analysis Center) at Caltech for easy location of data for a given extragalactic astronomical source, including SEDs. SedImporter is a new tool for converting non-standard SED data files into a format supported by Iris. We demonstrate the use of SedImporter to retrieve SEDs from a variety of sources-from the NED SED service, from the user's own data, and from other VO applications using SAMP (Simple Application Messaging Protocol). We also demonstrate the use of Iris to read, display, select ranges from, and fit models to SEDs. Finally, we discuss the architecture of Iris, and the use of IVOA standards so that Specview, Sherpa, SEDLib and SedImporter work together seamlessly.

  8. Physics design of experimental metal fuelled fast reactor cores for full scale demonstration

    Microsoft Academic Search

    K. Devan; Abhitab Bachchan; A. Riyas; T. Sathiyasheela; P. Mohanakrishnan; S. C. Chetal

    Fast breeder reactors based on metal fuel are planned to be in operation for the year beyond 2025 to meet the growing energy demand in India. A road map is laid towards the development of technologies required for launching 1000MWe commercial metal breeder reactors with closed fuel cycle. Construction of a test reactor with metallic fuel is also envisaged to

  9. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY-2011 Activities

    SciTech Connect

    Michael A. Pope

    2011-10-01

    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.

  10. Evolution of fast reactor core spectra in changing a heavy liquid metal coolant by molten PB-208

    SciTech Connect

    Blokhin, D. A.; Mitenkova, E. F. [Nuclear Safety Inst., Russian Academy of Sciences, B. Tulskaya 52, Moscow, 115119 (Russian Federation); Khorasanov, G. L.; Zemskov, E. A.; Blokhin, A. I. [State Scientific Center, Russian Federation, Inst. of Physics and Power Engineering, Bondarenko Square 1, Obninsk, 249033 (Russian Federation)

    2012-07-01

    In the paper neutron spectra of fast reactor cooled with lead-bismuth or lead-208 are given. It is shown that in changing the coolant from lead-bismuth to lead-208 the core neutron spectra of the fast reactor FR RBEC-M are hardening in whole by several percents when a little share of low energy neutrons (5 eV - 50 keV) is slightly increasing. The shift of spectra to higher energies permits to enhance the fuel fission while the increased share of low energy neutrons provides more effective conversion of uranium-238 into plutonium due to peculiarity of {sup 238}U neutron capture cross section. Good neutron and physical features of molten {sup 208}Pb permit to assume it as perspective coolant for fast reactors and accelerator driven systems. The one-group cross sections of neutron radiation capture, {sigma}(n,g), by {sup 208}Pb, {sup 238}U, {sup 99}Tc, mix of lead and bismuth, {sup nat}Pb-Bi, averaged over neutron spectra of the fast reactor RBEC-M are given. It is shown that one-group cross sections of neutron capture by material of the liquid metal coolant consisted from lead enriched with the stable lead isotope, {sup 208}Pb, are by 4-7 times smaller {sigma}(n,g) for the coolant {sup nat}Pb-Bi. The economy of neutrons in the core cooled with {sup 208}Pb can be used for reducing reactor's initial fuel load, increasing fuel breeding and transmutation of long lived fission products, for example {sup 99}Tc. Good neutron and physical features of lead enriched with {sup 208}Pb permit to consider it as a perspective low neutron absorbing coolant for fast reactors and accelerator driven systems. (authors)

  11. Mass estimates of very small reactor cores fueled by Uranium-235, U-233 and Cm-245

    NASA Astrophysics Data System (ADS)

    Wright, Steven A.; Lipinski, Ronald J.

    2001-02-01

    This paper explores the possibility of manufacturing very small reactors from U-235, U-233 and Cm-245. Pin type reactor systems fueled with uranium or curium metal zirconium hydride (UZrH or CmZrH) are compared with similar designs using U-235. Criticality measurements of homogeneous water uranium systems, suggest that reactor subsystem masses have a broad minimum for hydrogen-to-uranium atom ratios that vary from 25-250. This paper compares the masses of metal-hydride fueled reactor systems that use U-235, U-233, and Cm-245 fuel with hydrogen-to-metal atom ratios from 20-300 when cooled by gas (HeXe), liquid metal (Na), and water. The results indicate that water cooled reactors in general have the smallest reactor subsystem mass. For gas and liquid-metal cooled reactors U-233 subsystems have total masses that are about 1/2 those of similarly designed U-235 fuel reactors. Reactor subsystems consisting of 11.2% enriched Cm-245 (balance Cm-244) that can be obtained from fuel reprocessing have system masses comparable to that of U-233. The smallest reactor subsystem masses were on the order of 60-80 kg for U-233 fueled water cooled reactors. .

  12. Validation of finite difference core diffusion calculation methods with FEM and NEM for VVER-1000 MWe reactor

    SciTech Connect

    Jagannathan, V. [Light Water Reactor Physics Section, Reactor Physics Design Div., Bhabha Atomic Research Centre, Mumbai - 400 085 (India); RPDD, Central Complex, BARC, Mumbai - 400085 (India); Singh, T. [Reactor Physics and Nuclear Engineering Section, Reactor Group, BARC, Mumbai (India); Pal, U.; Karthikeyan, R. [Light Water Reactor Physics Section, Reactor Physics Design Div., Bhabha Atomic Research Centre, Mumbai - 400 085 (India); Sundaram, G. [Nuclear Safety Group, KK-NPC, Mumbai (India)

    2006-07-01

    India is developing several in-house fuel management codes for the design evaluation of WER-1000 M We reactors, being built at Kudankulam, Tamil Nadu in collaboration with Russian Federation. A lattice burnup code EXCEL provides the few group lattice parameters of various fuel assembly types constituting the core. The core diffusion analyses have been performed by two methods. In the first method the entire fuel assembly is treated as a single homogenized cell. Each fuel assembly cell is divided into 6n{sup 2} triangles, where 'n' is the number of uniform divisions on a side of the hexagon. Regular triangular meshes are used in the active core as well as in surrounding reflector regions. This method is incorporated in the code TRIHEXFA. In the second method a pin by pin description of the core is accomplished by considering the few group lattice parameters generated by EXCEL code for various fuel and non-fuel cells in each fuel assembly. Regular hexagonal cells of one pin pitch are considered in the core and reflector regions. This method is incorporated in HEXPIN code. Both these codes use centre mesh finite difference method (FDM) for regular triangular or hexagonal meshes. It is well known that the large size of the WER fuel assembly, the zigzag structure of the core-baffle zone, the distribution of water tubes of different diameter in this baffle zone and the surrounding steel and water layers of different thickness, all lead to a very complex description of the core-reflector interface. We are analyzing the WER core in fresh state by two other approaches to obtain independent benchmark reference solutions. They are finite element method (FEM) and nodal expansion method (NEM). The few group cross sections of EXCEL are used in the FEM and NEM analyses. The paper would present the comparison of the results of core followup simulations of FD codes with those of FEM and NEM analyses. (authors)

  13. Demonstration of long-term optical fiber thermometry in the in-core region of a nuclear reactor

    NASA Astrophysics Data System (ADS)

    Jensen, Fredrik B. H.; Nakazawa, Masaharu; Kakuta, Tsunemi; Shikama, Tatsuo; Narui, Minoru; Sagawa, Tsutomu

    1997-11-01

    An experimental demonstration of fiber optic temperature sensing in the in-core region of Japan Materials Testing Reactor from 250 to 750 degrees C is described. Temperature data could be obtained for two full-power weeks with neutron fluxes of approximately 1014 n/cm2/s and gamma dose rates of approximately 5 X 103 Gy/s. The measurements were based on thermally generated IR light within the optical fiber itself. The fiber thus served as both signal generator and signal transmitter to the out-of-core region. The fibers utilized in the experiments where of high OH pure-silica-core type and showed good radiation resistance. In the IR region the transmission of the fibers was only weakly affected by the incident radiation. Radiation induced luminescence and Cerenkov radiation in the optical fibers were found to have small influence on the signal in the IR window. The high OH content of the fibers used in the present experiment precluded the use of the spectral regions at 945, 1245, and 1390 nm, due to the high intrinsic and radiation induced absorption at these wavelengths. The use of silica fibers limited the maximum temperature to < 1000 degrees C. The present experiments show that optical sensors based on IR emission can be used to monitor temperature in the in-core region of nuclear reactors for extended periods of time.

  14. MORECA: A computer code for simulating modular high-temperature gas-cooled reactor core heatup accidents

    SciTech Connect

    Ball, S.J. (Oak Ridge National Lab., TN (United States))

    1991-10-01

    The design features of the modular high-temperature gas-cooled reactor (MHTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. This report describes the ORNL MORECA code, which was developed for analyzing postulated long-term core heatup scenarios for which active cooling systems used to remove afterheat following the accidents can be assumed to the unavailable. Simulations of long-term loss-of-forced-convection accidents, both with and without depressurization of the primary coolant, have shown that maximum core temperatures stay below the point at which any significant fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. MORECA models the US Department of Energy reference design of a standard MHTGR.

  15. Analysis of Sodium Fire in the Containment Building of Prototype Fast Breeder Reactor Under the Scenario of Core Disruptive Accident

    SciTech Connect

    Rao, P.M.; Kasinathan, N. [Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Kannan, S.E. [Atomic Energy Regulatory Board, Niyamak Bhavan, Anushaktinagar, Mumbai 400 094 (India)

    2006-07-01

    The potential for sodium release to reactor containment building from reactor assembly during Core Disruptive Accident (CDA) in Fast Breeder Reactors (FBR) is an important safety issue with reference to the structural integrity of Reactor Containment Building (RCB). For Prototype Fast Breeder Reactor (PFBR), the estimated sodium release under a CDA of 100 MJ energy release is 350 kg. The ejected sodium reacts easily with air in RCB and causes temperature and pressure rise in the RCB. For estimating the severe thermal consequences in RCB, different modes of sodium fires like pool and spray fires were analyzed by using SOFIRE -- II and NACOM sodium fire computer codes. Effects of important parameters like amount of sodium, area of pool, containment air volume and oxygen concentration have been investigated. A peak pressure rise of 7.32 kPa is predicted by SOFIRE II code for 350 kg sodium pool fire in 86,000 m{sup 3} RCB volume. Under sodium release as spray followed by unburnt sodium as pool fire mode analysis, the estimated pressure rise is 5.85 kPa in the RCB. In the mode of instantaneous combustion of sodium, the estimated peak pressure rise is 13 kPa. (authors)

  16. EXPOSURE SUMMARIES FOR IRIS CHEMICALS.

    EPA Science Inventory

    The Integrated Risk Information System (IRIS), prepared and maintained by the National Center for Environmental Assessment (NCEA) of the U.S. Environmental Protection Agency (U.S. EPA), is an electronic database containing information on human health effects that may result from ...

  17. IrisBased Biometric Cryptosystems

    E-print Network

    Uhl, Andreas

    cryptographic key management systems to enhance security. Only few work, which tends to be very customIris­Based Biometric Cryptosystems Diplomarbeit zur Erlangung des Diplomgrades an der of the growing interest in biometrics a new field of research has emerged, entitled "biometric cryptosystems

  18. Application of reliability-centered maintenance to boiling water reactor emergency core cooling systems fault-tree analysis

    SciTech Connect

    Choi, Y.A.; Feltus, M.A. [Pennsylvania State Univ., University Park, PA (United States). Nuclear Engineering Dept.

    1995-07-01

    Reliability-centered maintenance (RCM) methods are applied to boiling water reactor plant-specific emergency core cooling system probabilistic risk assessment (PRA) fault trees. The RCM is a technique that is system function-based, for improving a preventive maintenance (PM) program, which is applied on a component basis. Many PM programs are based on time-directed maintenance tasks, while RCM methods focus on component condition-directed maintenance tasks. Stroke time test data for motor-operated valves (MOVs) are used to address three aspects concerning RCM: (a) to determine if MOV stroke time testing was useful as a condition-directed PM task; (b) to determine and compare the plant-specific MOV failure data from a broad RCM philosophy time period compared with a PM period and, also, compared with generic industry MOV failure data; and (c) to determine the effects and impact of the plant-specific MOV failure data on core damage frequency (CDF) and system unavailabilities for these emergency systems. The MOV stroke time test data from four emergency core cooling systems [i.e., high-pressure coolant injection (HPCI), reactor core isolation cooling (RCIC), low-pressure core spray (LPCS), and residual heat removal/low-pressure coolant injection (RHR/LPCI)] were gathered from Philadelphia Electric Company`s Peach Bottom Atomic Power Station Units 2 and 3 between 1980 and 1992. The analyses showed that MOV stroke time testing was not a predictor for eminent failure and should be considered as a go/no-go test. The failure data from the broad RCM philosophy showed an improvement compared with the PM-period failure rates in the emergency core cooling system MOVs. Also, the plant-specific MOV failure rates for both maintenance philosophies were shown to be lower than the generic industry estimates.

  19. Historical comparisons of IRI and early ionograms

    NASA Astrophysics Data System (ADS)

    Rice, Donald; Sojka, Jan J.

    2015-04-01

    The IRI2012 provides ionospheric modeling from 1 January 1958 through the present and near future. However, archives of ionogram films exist dating back to the late 1940s, and are potentially valuable for studying long-term trends and change. IRI is very useful for the analysis and interpretation of the films, so Space Environment Corporation (SEC) has modified IRI2012 to extend its operations back to 1 January 1950. This paper describes results from IRI2012 and observations from the Washington DC ionosonde WA938 (38.7°N, -77.1°E) for 1951 (active post-solar maximum) and 1954 (quiet solar minimum). The comparison shows general agreement between the extended IRI2012 and the ionosonde observations. A nighttime enhancement found in IRI results is observed in some ionograms, with modification by atmospheric waves. A significant discrepancy between IRI and observations was found in nighttime 1954 solar minimum results.

  20. Recent research results in iris biometrics

    NASA Astrophysics Data System (ADS)

    Hollingsworth, Karen; Baker, Sarah; Ring, Sarah; Bowyer, Kevin W.; Flynn, Patrick J.

    2009-05-01

    Many security applications require accurate identification of people, and research has shown that iris biometrics can be a powerful identification tool. However, in order for iris biometrics to be used on larger populations, error rates in the iris biometrics algorithms must be as low as possible. Furthermore, these algorithms need to be tested in a number of different environments and configurations. In order to facilitate such testing, we have collected more than 100,000 iris images for use in iris biometrics research. Using this data, we have developed a number of techniques for improving recognition rates. These techniques include fragile bit masking, signal-level fusion of iris images, and detecting local distortions in iris texture. Additionally, we have shown that large degrees of dilation and long lapses of time between image acquisitions negatively impact performance.

  1. Monte Carlo Calculation of Core Reactivity and Fluxes for the Development of the BNCT Neutron Source at the Kyiv Research Reactor

    NASA Astrophysics Data System (ADS)

    Gritzay, Olena; Kalchenko, Oleksandr; Klimova, Nataliya; Razbudey, Volodymyr; Sanzhur, Andriy; Binney, Stephen

    2005-05-01

    The presented results show our consecutive steps in developing a neutron source with parameters required by Boron Neutron Capture Therapy (BNCT) at the Kyiv Research Reactor (KRR). The main goal of this work was to analyze the influence of installation of different types of uranium converters close to the reactor core on neutron beam characteristics and on level of reactor safety. The general Monte Carlo radiation transport code MCNP, version 4B, has been used for these calculations.

  2. Monte Carlo Calculation of Core Reactivity and Fluxes for the Development of the BNCT Neutron Source at the Kyiv Research Reactor

    SciTech Connect

    Gritzay, Olena; Kalchenko, Oleksandr; Klimova, Nataliya; Razbudey, Volodymyr; Sanzhur, Andriy [Institute for Nuclear Research, Prospekt Nauky 47, Kyiv, 03680 (Ukraine); Binney, Stephen [Oregon State University, Corvallis, OR 97331-5902 (United States)

    2005-05-24

    The presented results show our consecutive steps in developing a neutron source with parameters required by Boron Neutron Capture Therapy (BNCT) at the Kyiv Research Reactor (KRR). The main goal of this work was to analyze the influence of installation of different types of uranium converters close to the reactor core on neutron beam characteristics and on level of reactor safety. The general Monte Carlo radiation transport code MCNP, version 4B, has been used for these calculations.

  3. Neutron transport with the method of characteristics for 3-D full core boiling water reactor applications

    Microsoft Academic Search

    Justin W. Thomas

    2006-01-01

    The Numerical Nuclear Reactor (NNR) is a code suite that is being developed to provide high-fidelity multi-physics capability for the analysis of light water nuclear reactors. The focus of the work here is to extend the capability of the NNR by incorporation of the neutronics module, DeCART, for Boiling Water Reactor (BWR) applications. The DeCART code has been coupled to

  4. Numerical simulation of the power characteristics of twin-core pulse reactor-pumped laser system

    NASA Astrophysics Data System (ADS)

    Gulevich, A. V.; Barzilov, A. P.; Dyachenko, P. P.; Zrodnikov, A. V.; Kukharchuk, O. F.; Kachanov, B. V.; Kolyada, S. G.; Pashin, E. A.

    1996-05-01

    Concept for high-power pulsed reactor-pumped laser system (RPLS) based on the new physical principles (direct nuclear-to-optical conversion) is discussed with reference to ICF feasibility problem. Theoretical problems for substantiation of the neutronic and physical characteristics of the RPLS power model are considered. Results of numerical studies of the expected power characteristics of reactor laser system are discussed.

  5. The best bits in an iris code.

    PubMed

    Hollingsworth, Karen P; Bowyer, Kevin W; Flynn, Patrick J

    2009-06-01

    Iris biometric systems apply filters to iris images to extract information about iris texture. Daugman's approach maps the filter output to a binary iris code. The fractional Hamming distance between two iris codes is computed and decisions about the identity of a person are based on the computed distance. The fractional Hamming distance weights all bits in an iris code equally. However, not all the bits in an iris code are equally useful. Our research is the first to present experiments documenting that some bits are more consistent than others. Different regions of the iris are compared to evaluate their relative consistency, and contrary to some previous research, we find that the middle bands of the iris are more consistent than the inner bands. The inconsistent-bit phenomenon is evident across genders and different filter types. Possible causes of inconsistencies, such as segmentation, alignment issues, and different filters are investigated. The inconsistencies are largely due to the coarse quantization of the phase response. Masking iris code bits corresponding to complex filter responses near the axes of the complex plane improves the separation between the match and nonmatch Hamming distance distributions. PMID:19372603

  6. Monte Carlo estimation of the dose and heating of cobalt adjuster rods irradiated in the CANDU 6 reactor core.

    PubMed

    Gugiu, Daniela; Dumitrache, Ion

    2005-01-01

    The present work is a part of a more complex project related to the replacement of the original stainless steel adjuster rods with cobalt assemblies in the CANDU 6 reactor core. The 60Co produced by 59Co irradiation could be used extensively in medicine and industry. The paper will mainly describe some of the reactor physics and safety requirements that must be carried into practice for the Co adjuster rods. The computations related to the neutronic equivalence of the stainless steel adjusters with the Co adjuster assemblies, as well as the estimations of the activity and heating of the irradiated cobalt rods, are performed using the Monte Carlo codes MCNP5 and MONTEBURNS 2.1. The activity values are used to evaluate the dose at the surface of the device designed to transport the cobalt adjusters. PMID:16604599

  7. MATADOR: a computer code for the analysis of radionuclide behavior during degraded core accidents in light water reactors

    SciTech Connect

    Baybutt, P.; Raghuram, S.; Avci, H.I.

    1985-04-01

    A new computer code called MATADOR (Methods for the Analysis of Transport And Deposition Of Radionuclides) has been developed to replace the CORRAL computer code which was written for the Reactor Safety Study (WASH-1400). This report contains a detailed description of the models used in MATADOR. MATADOR is intended for use in system risk studies to analyze radionuclide transport and deposition in reactor containments. The principal output of the code is information on the timing and magnitude of radionuclide releases to the environment as a result of severely degraded core accidents. MATADOR considers the transport of radionuclides through the containment and their removal by natural deposition and the operation of engineered safety systems such as sprays. The code requires input data on the source term from the primary system, the geometry of the containment, and the thermal-hydraulic conditions in the containment.

  8. Fluidized-bed atomic layer deposition reactor for the synthesis of core-shell nanoparticles

    SciTech Connect

    Didden, Arjen P.; Middelkoop, Joost; Krol, Roel van de, E-mail: roel.vandekrol@helmholtzberlin.de [Delft University of Technology, Faculty of Applied Sciences, Department of Chemical Engineering, P.O. Box 5045, 2600 GA Delft (Netherlands); Besling, Wim F. A. [NXP Semiconductors, High Tech Campus 32, 5656 AE Eindhoven (Netherlands)] [NXP Semiconductors, High Tech Campus 32, 5656 AE Eindhoven (Netherlands); Nanu, Diana E. [Thin Film Factory B.V., Hemma Oddastrjitte 5, 8927 AA Leeuwarden (Netherlands)] [Thin Film Factory B.V., Hemma Oddastrjitte 5, 8927 AA Leeuwarden (Netherlands)

    2014-01-15

    The design of a fluidized bed atomic layer deposition (ALD) reactor is described in detail. The reactor consists of three parts that have all been placed in one protective cabinet: precursor dosing, reactor, and residual gas treatment section. In the precursor dosing section, the chemicals needed for the ALD reaction are injected into the carrier gas using different methods for different precursors. The reactor section is designed in such a way that a homogeneous fluidized bed can be obtained with a constant, actively controlled, reactor pressure. Furthermore, no filters are required inside the reactor chamber, minimizing the risk of pressure increase due to fouling. The residual gas treatment section consists of a decomposition furnace to remove residual precursor and a particle filter and is installed to protect the pump. In order to demonstrate the performance of the reactor, SiO{sub 2} particles have been coated with TiO{sub 2} using tetrakis-dimethylamino titanium (TDMAT) and H{sub 2}O as precursors. Experiments with varying pulse times show that saturated growth can be obtained with TDMAT pulse times larger than 600 s. Analysis of the powder with High-Angle Annular Dark-Field Scanning Transmission Electron Microscopy (HAADF-STEM) and energy dispersive X-ray spectroscopy confirmed that after 50 cycles, all SiO{sub 2} particles were coated with a 1.6 nm homogenous shell of TiO{sub 2}.

  9. Development of Inspection Modality for Shell Weld of Core Support Structure of a Fast Breeder Reactor Using Civa

    NASA Astrophysics Data System (ADS)

    Rao, Chelamchala Babu; Raillon, Raphaële; Sharma, Govind Kumar; Jayakumar, Tammana; Benoist, Philippe; Raj, Baldev

    2010-02-01

    The core support structure is welded with a 40 mm thick base plate of the main vessel, of 500 MWe Prototype Fast Breeder Reactor, Kalpakkam, India, along the circumference. This `shell weld' situated at a distance of about 435 mm away from the weld overlay, is in-accessible to contact mode ultrasonic testing during in-service inspection. An unconventional ultrasonic methodology was developed for this purpose. This inspection modality is validated using the ultrasonic module of CIVA simulation software. There is reasonable agreement with experimental measurements.

  10. Infrared Imaging Surveyor (IRIS) project

    NASA Astrophysics Data System (ADS)

    Shibai, Hiroshi; Murakami, Hiroshi

    1996-06-01

    In this paper we describe the concept and the design of the InfraRed Imaging Surveyor (IRIS), the first Japanese satellite solely dedicated to infrared astronomy. It will follow a successful precursor, the Infrared Telescope in Space (IRTS) onboard the Space Flyer Et (SFU) in 1995. The IRIS has a 70 cm telescope cooled down to 7 K by using superfluid helium assisted by two-state Stirling cycle coolers. The expected hold time of the super-fluid helium is one year. After consumption of the helium, near-infrared observation can be continued by using the mechanical coolers. Two focal plane instruments are planned; the infrared camera (IRC) and the far-infrared surveyor (FIS). The total spectral coverage is 2 to 200 microns. The major scientific objectives are to investigate birth and evolution of galaxies in the early universe by survey of young normal galaxies and starburst galaxies. The orbit is a sun- synchronous orbit, in which the cooled telescope can avoid huge emissions from the Sun and the Earth by pointing the telescope on the great circle perpendicular to the Sun. The IRIS project is expected to start in 1997 and it will be launched by a M-V rocket in 2002.

  11. Stand-off Iris Recognition System

    Microsoft Academic Search

    Frederick W. Wheeler; A. G. Amitha Perera; Gil Abramovich; Bing Yu; Peter H. Tu

    2008-01-01

    The iris is a highly accurate biometric identifier. However widespread adoption is hindered by the difficulty of capturing high-quality iris images with minimal user co-operation. This paper describes a first-generation prototype iris identification system designed for stand-off cooperative access control. This system identifies individuals who stand in front of and face the system after 3.2 seconds on average. Subjects within

  12. Recognising Persons by Their Iris Patterns

    Microsoft Academic Search

    John Daugman

    2004-01-01

    \\u000a Algorithms developed by the author for recognizing persons by their iris patterns have now been tested in many field deployments,\\u000a producing no false matches in millions of iris comparisons. The recognition principle is the failure of a test of statistical\\u000a independence on iris phase structure, as encoded by multi-scale quadrature 2D Gabor wavelets. The combinatorial complexity\\u000a of this phase information

  13. Optimized core design of a supercritical carbon dioxide-cooled fast reactor

    E-print Network

    Handwerk, Christopher S. (Christopher Stanley), 1974-

    2007-01-01

    Spurred by the renewed interest in nuclear power, Gas-cooled Fast Reactors (GFRs) have received increasing attention in the past decade. Motivated by the goals of the Generation-IV International Forum (GIF), a GFR cooled ...

  14. Development of a low enrichment uranium core for the MIT reactor

    E-print Network

    Newton, Thomas Henderson

    2006-01-01

    An investigation has been made into converting the MIT research reactor from using high enrichment uranium (HEU) to low enrichment uranium (LEU) with a newly developed fuel material. The LEU fuel introduces negative ...

  15. Configuration adjustment potential of the Very High Temperature Reactor prismatic cores with advanced actinide fuels 

    E-print Network

    Ames, David E, II

    2006-10-30

    Minor actinides represent the long-term radiotoxicity of nuclear wastes. As one of their potential incineration options, partitioning and transmutation in fission reactors are seriously considered worldwide. If implemented, ...

  16. Preliminary core characterization of a Generation IV lead fast reactor DEMO: Goals, design rationales and options

    Microsoft Academic Search

    Sara Bortot; Carlo Artioli; Giacomo Grasso; Vincenzo Peluso; Marco E. Ricotti

    2010-01-01

    An Italian effort has been initiated under a cooperation between ENEA and CIRTEN for the investigation of a concept for a small size Generation IV Lead-cooled Fast Reactor (LFR) demonstration project (DEMO) which could provide significant support to the European Lead-cooled System (ELSY), by validating lead technology and the overall system behaviour.A demonstration reactor is expected to prove the viability

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

    Microsoft Academic Search

    Johan Carlsson; Kamil Tucek; Hartmut Wider

    2006-01-01

    This paper concerns two independent safety investigations on critical and sub-critical heavy liquid metal cooled fast reactors using simple flow paths. The first investigation applies to locating the steam generators in the risers instead of the down-comers of a simple flow path designed sub-critical reactor of 600 MW{sub th} power. This was compared to a similar design, but with the

  18. Recognising persons by their iris patterns

    NASA Astrophysics Data System (ADS)

    Daugman, John

    2010-04-01

    Iris recognition provides real-time, high confidence identification of persons by analysis of the random patterns that are visible within the iris of an eye from some distance. Because the iris is a protected, internal, organ whose random texture is epigenetic and stable over the lifespan, it can serve as a living password. Recognition decisions are made with confidence levels high enough to support rapid exhaustive searches through national-sized databases. The principle that underlies these algorithms is the failure of an efficient test of statistical independence involving more than 200 degrees-of-freedom, based on phase sequencing each iris pattern with quadrature 2D wavelets. Different persons always pass this test of statistical independence, but images from the same iris almost always fail this test of independence. Database search speeds are around 1 million persons per second per CPU. Data from 200 billion cross-comparisons between different eyes will be presented in this talk, using a database consisting of 632,500 iris images acquired in the United Arab Emirates in a networked national border-crossing security system which performs, every day, about 9 billion iris comparisons using these algorithms. Current research efforts with this technology aim to make it more tolerant of difficult conditions of iris capture, such as "iris on the move," at a distance, and off-axis.

  19. Design analysis of the molten core confinement within the reactor vessel in the case of severe accidents at nuclear power plants equipped with a reactor of the VVER type

    NASA Astrophysics Data System (ADS)

    Zvonaryov, Yu. A.; Budaev, M. A.; Volchek, A. M.; Gorbaev, V. A.; Zagryazkin, V. N.; Kiselyov, N. P.; Kobzar', V. L.; Konobeev, A. V.; Tsurikov, D. F.

    2013-12-01

    The present paper reports the results of the preliminary design estimate of the behavior of the core melt in vessels of reactors of the VVER-600 and VVER-1300 types (a standard optimized and informative nuclear power unit based on VVER technology—VVER TOI) in the case of beyond-design-basis severe accidents. The basic processes determining the state of the core melt in the reactor vessel are analyzed. The concept of molten core confinement within the vessel based on the idea of outside cooling is discussed. Basic assumptions and models, as well as the results of calculation of the interaction between molten materials of the core and the wall of the reactor vessel performed by means of the SOCRAT severe accident code, are presented and discussed. On the basis of the data obtained, the requirements on the operation of the safety systems are determined, upon the fulfillment of which there will appear potential prerequisites for implementing the concept of the confinement of the core melt within the reactor in cases of severe accidents at nuclear power plants equipped with VVER reactors.

  20. A Pin Power Reconstruction Method for CANDU Reactor Cores Based on Coarse-Mesh Finite Difference Calculations

    SciTech Connect

    Lee, Hyung-Seok [Chosun University (Korea, Republic of); Yang, Won Sik [Chosun University (Korea, Republic of); Na, Man Gyun [Chosun University (Korea, Republic of); Choi, Hangbok [Korea Atomic Energy Research Institute (Korea, Republic of)

    2000-04-15

    A reconstruction method has been developed for recovering pin powers from Canada deuterium uranium (CANDU) reactor core calculations performed with a coarse-mesh finite difference diffusion approximation and single-assembly lattice calculations. The homogeneous intranodal distributions of group fluxes are efficiently computed using polynomial shapes constrained to satisfy the nodal information approximated from the node-average fluxes. The group fluxes of individual fuel pins in a heterogeneous fuel bundle are determined using these homogeneous intranodal flux distributions and the form functions obtained from the single-assembly lattice calculations. The pin powers are obtained using these pin fluxes and the pin power cross sections generated by the single-assembly lattice calculation. The accuracy of the reconstruction schemes has been estimated by performing benchmark calculations for partial core representation of a natural uranium CANDU reactor. The results indicate that the reconstruction schemes are quite accurate, yielding maximum pin power errors of less than {approx}3%. The main contribution to the reconstruction error is made by the errors in the node-average fluxes obtained from the coarse-mesh finite difference diffusion calculation; the errors due to the reconstruction schemes are <1%.

  1. Measurements of actinide-fission product yields in Caliban and Prospero metallic core reactor fission neutron fields

    SciTech Connect

    Casoli, P.; Authier, N. [CEA, Centre de Valduc, 21120 Is-sur-Tille (France); Laurec, J.; Bauge, E.; Granier, T. [CEA, Centre DIF, 91297 Arpajon (France)

    2011-07-01

    In the 1970's and early 1980's, an experimental program was performed on the facilities of the CEA Valduc Research Center to measure several actinide-fission product yields. Experiments were, in particular, completed on the Caliban and Prospero metallic core reactors to study fission-neutron-induced reactions on {sup 233}U, {sup 235}U, and {sup 239}Pu. Thick actinide samples were irradiated and the number of nuclei of each fission product was determined by gamma spectrometry. Fission chambers were irradiated simultaneously to measure the numbers of fissions in thin deposits of the same actinides. The masses of the thick samples and the thin deposits were determined by mass spectrometry and alpha spectrometry. The results of these experiments will be fully presented in this paper for the first time. A description of the Caliban and Prospero reactors, their characteristics and performances, and explanations about the experimental approach will also be given in the article. A recent work has been completed to analyze and reinterpret these measurements and particularly to evaluate the associated uncertainties. In this context, calculations have also been carried out with the Monte Carlo transport code Tripoli-4, using the published benchmarked Caliban description and a three-dimensional model of Prospero, to determine the average neutron energy causing fission. Simulation results will be discussed in this paper. Finally, new fission yield measurements will be proposed on Caliban and Prospero reactors to strengthen the results of the first experiments. (authors)

  2. New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions.

    PubMed

    Geslot, B; Vermeeren, L; Filliatre, P; Lopez, A Legrand; Barbot, L; Jammes, C; Bréaud, S; Oriol, L; Villard, J-F

    2011-03-01

    Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 × 10(20) n?cm(2). A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement. PMID:21456734

  3. New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

    NASA Astrophysics Data System (ADS)

    Geslot, B.; Vermeeren, L.; Filliatre, P.; Lopez, A. Legrand; Barbot, L.; Jammes, C.; Bréaud, S.; Oriol, L.; Villard, J.-F.

    2011-03-01

    Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 × 1020 n/cm2. A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

  4. Leveraging community support for Education and Outreach: The IRIS E&O Program

    NASA Astrophysics Data System (ADS)

    Taber, J.; Hubenthal, M.; Wysession, M. E.

    2009-12-01

    The IRIS E&O Program was initiated 10 years ago, some 15 years after the creation of the IRIS Consortium, as IRIS members increasingly recognized the fundamental need to communicate the results of scientific research more effectively and to attract more students to study Earth science. Since then, IRIS E&O has received core funding through successive 5-year cooperative agreements with NSF, based on proposals submitted by IRIS. While a small fraction of the overall Consortium budget, this consistent funding has allowed the development of strong, long-term elements within the E&O Program, including summer internships, IRIS/USGS museum displays, seismographs in schools, IRIS/SSA Distinguished Lecture series, and professional development for middle school and high school teachers. Reliable funding has allowed us to develop expertise in these areas due to the longevity of the programs and the continuous improvement resulting from ongoing evaluations. Support from Consortium members, including volunteering time and expertise, has been critical for the program, as the Consortium has to continually balance the value of E&O products versus equipment and data services for seismology research. The E&O program also provides service to the Consortium, such as PIs being able to count on and leverage IRIS resources when defining the broader impacts of their own research. The reliable base has made it possible to build on the core elements with focused and innovative proposals, allowing, for example, the expansion of our internship program into a full REU site. Developing collaborative proposals with other groups has been a key strategy where IRIS E&O's long-term viability can be combined with expertise from other organizations to develop new products and services. IRIS can offer to continue to reliably deliver and maintain products after the end of a 2-3 year funding cycle, which can greatly increase the reach of the project. Consortium backing has also allowed us to establish an educational fund in honor of the late John Lahr. This fund, which is comprised of individual donations, is being used to provide seismographs to schools along with professional development and ongoing support from the E&O program. We are also developing a plan for attracting larger private and/or foundation funds for new E&O activities, leveraging the reputation of a long-term program.

  5. Iris recognition: an emerging biometric technology

    Microsoft Academic Search

    RICHARD P. WILDES

    1997-01-01

    This paper examines automated iris recognition as a biometrically based technology for personal identification and verification. The motivation for this endeavor stems from the observation that the human iris provides a particularly interesting structure on which to base a technology for noninvasive biometric assessment. In particular the biomedical literature suggests that irises are as distinct as fingerprints or patterns of

  6. Glycosylated constituents of iris fulva and Iris brevicaulis.

    PubMed

    Fang, Rui; Veitch, Nigel C; Kite, Geoffrey C; Howes, Melanie-Jayne R; Porter, Elaine A; Simmonds, Monique S J

    2011-01-01

    The major constituents of leaf extracts of Iris fulva KER GAWL. comprised a known flavone C-glycoside, 5,4'-dihydroxy-7-methoxyflavone-6-C-(6?-O-(E)-p-coumaroyl-?-glucopyranosyl)(1??2?)-?-glucopyranoside (1) and the new monoterpene glycoside, linalyl-6'-O-(3?-hydroxy-3?-methylglutaroyl)-?-D-glucopyranoside (2), both of which were prominent components of Iris brevicaulis RAF. leaf extracts. The structure of a new polyacylated sucrose derivative (3a) obtained from the rhizomes of I. fulva was elucidated as 3-O-(E)-p-coumaroyl-?-D-fructofuranosyl-(2?1')-[2?,4?,6?-tri-O-acetyl-?-D-glucopyranosyl-(1??3')-(2',6'-di-O-acetyl-4'-O-(E)-p-coumaroyl-?-D-glucopyranoside)]. Selective hydrolysis of the 4?-O-acetyl moiety of the terminal ?-glucopyranosyl residue of 3a occurred after several hours in solution giving 3-O-(E)-p-coumaroyl-?-D-fructofuranosyl-(2?1')-[2?,6?-di-O-acetyl-?-D-glucopyranosyl-(1??3')-(2',6'-di-O-acetyl-4'-O-(E)-p-coumaroyl-?-D-glucopyranoside)] (3b), which subsequently underwent further deacetylation. PMID:21212561

  7. Investigations of sloshing fluid motions in pools related to recriticalities in liquid-metal fast breeder reactor core meltdown accidents

    SciTech Connect

    Maschek, W.; Munz, C.D.; Meyer, L. (Kernforschungszentrum Karlsruhe, Inst. fuer Neutronenphysik und Reaktortechnik, Postfach 3640, W-7500 Karlsruhe 1 (DE))

    1992-04-01

    This paper reports that analyses of unprotected loss-of-flow accidents for medium-size cores of current liquid-metal fast breeder reactors have shown that the accident proceeds into a transition phase where further meltdown is accompanied by recriticalities and secondary excursions. Assuming very pessimistic conditions concerning fuel discharge and blockage formation, a neutronically active whole-core pool of molten m material can form. Neutronic or thermohydraulic disturbances may initiate a special motion pattern in these pools, called centralized sloshing, which can lead to energetic power excursions. If such a whole-core pool is formed, its energetic potential must be adequately assessed. This requires sufficiently correct theoretical tools (codes) and proper consideration of the fluid-dynamic and thermo-hydraulic conditions for these pools. A series of experiments has been performed that serves as a benchmark for the SIMMER-II and the AFDM codes in assessing their adequacy in modeling such sloshing motions. Additional phenomenologically oriented experiments provide deeper insight into general motion patterns of sloshing fluids while taking special notice of asymmetries and obstacles that exist in such pools.

  8. Enhanced LMR (liquid metal reactors) core cooling utilizing passive vortex devices

    Microsoft Academic Search

    W. E. Bickford; J. W. Jaeckle; B. J. Webb

    1988-01-01

    Several design options for improved passive circulation flow have been investigated for use in small, modular liquid metal cooled reactors (LMRs). The purpose is to enhance the transition to natural convection cooling following loss of forced circulation flow, reducing thermal transients experienced by the fuel and possibly eliminating the need for emergency pony-motor flow. Design details to minimize pressure drops

  9. Fuel pins and core response under liquid-metal fast breeder reactor transient overpower accident conditions

    Microsoft Academic Search

    N. P. Wilburn; D. E. Smith; R. E. Baars; D. B. Atcheson; B. W. Spencer

    1979-01-01

    Since the earlier liquid-metal fast breeder reactor transient overpower assessments were done (1975), new experimental data and modeling improvements have occurred that indicate later failures and more molten fuel squirted into the channel with a higher propensity for plugging. An initial sweepout still occurs, and an analysis shows that even if coherent instead of the expected stochastic failures occur, the

  10. CORE-TEMPERATURE EXCURSIONS FOLLOWING A PIPING FAILURE IN THE PLUTONIUM RECYCLE TEST REACTOR

    Microsoft Academic Search

    A. W. Jr. Lemmon; C. A. Alexander; L. E. Hulbert; R. B. Jr. Filbert

    1959-01-01

    An evaluation of the temperature excursion and its possible consequences ; arising from loss of coolant from the Plutonium Recycle Test Reactor (PRTR) was ; made for four different postulated ruptures in the primary heavy water coolant ; system. As a basis for the evaluation, a series of computations was made. These ; were based on incremental heat and mass

  11. An application of the new way to prevent core melting in pressure tube reactors (CANDU type)

    Microsoft Academic Search

    Stefan Mehedinteanu

    2001-01-01

    The pressure tube reactors, especially CANDU type, have a calandria low pressure vessel (near to atmospheric pressure) immersed into a concrete vault filled with water. The accident analysis done by ELFIN-HTCELL code for the channel heat up and by fluid flow PHOENICS code as applied for moderator cooling system efficacy, showed that even the moderator cooling system operates, in some

  12. Fuel and core testing plan for a target fueled isotope production reactor

    Microsoft Academic Search

    Richard Lee Coats; James J. Dahl; Parma Edward J. Jr

    2010-01-01

    In recent years there has been an unstable supply of the critical diagnostic medical isotope 99Tc. Several concepts and designs have been proposed to produce 99Mo the parent nuclide of 99Tc, at a commercial scale sufficient to stabilize the world supply. This work lays out a testing and experiment plan for a proposed 2 MW open pool reactor fueled by

  13. In vivo OCT microangiography of rodent iris

    PubMed Central

    Choi, Woo June; Zhi, Zhongwei; Wang, Ruikang K.

    2014-01-01

    We report on the functional optical coherence tomography (OCT) imaging of iris tissue morphology and microcirculation in living small animals. Anterior segments of healthy mouse and rat eyes are imaged with high-speed spectral domain OCT (SD-OCT) utilizing ultra-high sensitive optical microangiography (UHS-OMAG) imaging protocol. 3D iris microvasculature is produced by the use of an algorithm that calculates absolute differences between the amplitudes of the OCT inter-frames. We demonstrate that the UHS-OMAG is capable of delineating iris microvascular beds in the mouse and rat with capillary-level resolution. Furthermore, the fast imaging speed enables dynamic imaging of iris micro-vascular response during drug-induced pupil dilation. We believe that this OCT angiographic approach has a great potential for in situ and in vivo monitoring of the microcirculation within iris tissue beds in rodent disease models that have microvascular involvement. PMID:24979017

  14. WWER-1000 core and reflector parameters investigation in the LR-0 reactor

    SciTech Connect

    Zaritsky, S. M.; Alekseev, N. I.; Bolshagin, S. N. [RRC Kurchatov Inst., 1 Kurchatov Sq., Moscow, 123182 (Russian Federation); Riazanov, D. K.; Lichadeev, V. V. [Research Inst. of Atomic Reactors, Dimitrovgrad 10, 433510 (Russian Federation); Ocmera, B. [Nuclear Research Inst., Rez, 25068 (Czech Republic); Cvachovec, F. [Univ. of Defense, 65 Kounicova st., Brno, 61200 (Czech Republic)

    2006-07-01

    Measurements and calculations carried out in the core and reflector of WWER-1000 mock-up are discussed: - the determination of the pin-to-pin power distribution in the core by means of gamma-scanning of fuel pins and pin-to-pin calculations with Monte Carlo code MCU-REA and diffusion codes MOBY-DICK (with WIMS-D4 cell constants preparation) and RADAR - the fast neutron spectra measurements by proton recoil method inside the experimental channel in the core and inside the channel in the baffle, and corresponding calculations in P{sub 3}S{sub 8} approximation of discrete ordinates method with code DORT and BUGLE-96 library - the neutron spectra evaluations (adjustment) in the same channels in energy region 0.5 eV-18 MeV based on the activation and solid state track detectors measurements. (authors)

  15. Thermohydraulics in a high-temperature gas-cooled reactor primary loop during early phases of unrestricted core-heatup accidents

    SciTech Connect

    Kroeger, P.G.; Colman, J.; Hsu, C.J.

    1983-01-01

    In High Temperature Gas Cooled Reactor (HTGR) siting considerations, the Unrestricted Core Heatup Accidents (UCHA) are considered as accidents of highest consequence, corresponding to core meltdown accidents in light water reactors. Initiation of such accidents can be, for instance, due to station blackout, resulting in scram and loss of all main loop forced circulation, with none of the core auxiliary cooling system loops being started. The result is a slow but continuing core heatup, extending over days. During the initial phases of such UCHA scenarios, the primary loop remains pressurized, with the system pressure slowly increasing until the relief valve setpoint is reached. The major objectives of the work described here were to determine times to depressurization as well as approximate loop component temperatures up to depressurization.

  16. A review of the SIMMER-2 analyses of liquid-metal-cooled fast breeder reactor core-disruptive accident fuel escape

    NASA Astrophysics Data System (ADS)

    Devault, G. P.; Bell, C. R.

    Early fuel removal from the active core of a liquid-metal-cooled fast breeder reactor undergoing a core-disruptive accident may reduce the potential for large energetics resulting from recriticalities. A review of analyses with the SIMMER-II computer program of the effectiveness of possible fuel escape paths is presented. Where possible, how SIMMER-II compares with or is validated against experiments that simulated the escape paths also is discussed.

  17. Review of the SIMMER-II analyses of liquid-metal-cooled fast breeder reactor core-disruptive accident fuel escape

    SciTech Connect

    DeVault, G.P.; Bell, C.R.

    1985-01-01

    Early fuel removal from the active core of a liquid-metal-cooled fast breeder reactor undergoing a core-disruptive accident may reduce the potential for large energetics resulting from recriticalities. This paper presents a review of analyses with the SIMMER-II computer program of the effectiveness of possible fuel escape paths. Where possible, how SIMMER-II compares with or is validated against experiments that simulated the escape paths also is discussed.

  18. Core-shell structured microcapsular-like Ru@SiO2 reactor for efficient generation of CO(x)-free hydrogen through ammonia decomposition.

    PubMed

    Li, Yanxing; Yao, Lianghong; Song, Yanyan; Liu, Shunqiang; Zhao, Jing; Ji, Weijie; Au, Chak-Tong

    2010-08-01

    The core-shell structured microcapsular-like Ru@SiO(2) reactor is proved to be the most efficient material known to date for CO(x)-free hydrogen production via ammonia decomposition for fuel cells application. The very active Ru core particles can retain good stability even at high temperatures (up to 650 degrees C) thanks to the protection of the inert SiO(2) nano-shell. PMID:20544106

  19. Coarse-grained parallel genetic algorithm applied to a nuclear reactor core design optimization problem

    Microsoft Academic Search

    Cláudio M. N. A. Pereira; Celso M. F. Lapa

    2003-01-01

    This work extends the research related to genetic algorithms (GA) in core design optimization problems, which basic investigations were presented in previous work. Here we explore the use of the Island Genetic Algorithm (IGA), a coarse-grained parallel GA model, comparing its performance to that obtained by the application of a traditional non-parallel GA. The optimization problem consists on adjusting several

  20. Analysis of proposed gamma-ray detection system for the monitoring of core water inventory in a pressurized water reactor

    SciTech Connect

    Markoff, D.M.

    1987-12-01

    An initial study has been performed of the feasibility of employing an axial array of gamma detectors located outside the pressure vessel to monitor the coolant in a PWR. A one-dimensional transport analysis model is developed for the LOFT research reactor and for a mock-PWR geometry. The gamma detector response to coolant voiding in the core and downcomer has been determined for both geometries. The effects of various conditions (for example, time after shutdown, materials in the transport path, and the relative void fraction in different water regions) on the detector response are studied. The calculational results have been validated by a favorable comparison with LOFT experimental data. Within the limitations and approximations considered in the analysis, the results indicate that the gamma-ray detection scheme is able to unambiguously respond to changes in the coolant inventory within any vessel water region.

  1. Real-Time Image Restoration for Iris Recognition Systems

    Microsoft Academic Search

    Byung Jun Kang; Kang Ryoung Park

    2007-01-01

    In the field of biometrics, it has been reported that iris recognition techniques have shown high levels of accuracy because unique patterns of the human iris, which has very many degrees of freedom, are used. However, because conventional iris cameras have small depth-of-field (DOF) areas, input iris images can easily be blurred, which can lead to lower recognition performance, since

  2. IRIS TOXICOLOGICAL REVIEW AND SUMMARY DOCUMENTS FOR 2-HEXANONE

    EPA Science Inventory

    EPA will conduct an assessment of the noncancer health effects of 2-hexanone. The IRIS program will prepare an IRIS assessment for 2-hexanone. The IRIS assessment for 2-hexanone will consist of a Toxicological Review and an IRIS Summary. The Toxicological Review is a critical ...

  3. Analysis of cracked core spray injection line piping from the Quad Cities Units 1 and 2 boiling water reactors

    SciTech Connect

    Diercks, D.R.

    1983-12-01

    Elbow assemblies and adjacent piping from the loops A and B core spray injection lines of Quad Cities Units 1 and 2 Boiling Water Reactors have been examined in order to determine the nature and causes of coolant leakages and flaw indications detected during hydrostatic tests and subsequent ultrasonic inspections. The elbow assemblies were found to contain multiple intergranular cracks in the weld heat-affected zones. The cracking was predominantly axial in orientation in the forged elbow and wedge components, whereas mixed axial and circumferential cracking was seen in the wrought piping pieces. In at least two instances, axial cracks completely penetrated the circumferential weld joining adjacent components. Based upon the observations made in the present study, the failures were attributed to intergranular stress corrosion cracking caused by the weld-induced sensitized microstructure and residual stresses present; dissolved oxygen in the reactor coolant apparently served as the corrosive species. The predominantly axial orientation of the cracks present in the forged components is believed to be related to the banded microstructure present in these components. The metallographic studies reported are supplemented by x-radiography, chemical analysis and mechanical test results, determinations of the degree of sensitization present, and measurements of weld metal delta ferrite content.

  4. Accident source terms for boiling water reactors with high burnup cores.

    SciTech Connect

    Gauntt, Randall O.; Powers, Dana Auburn; Leonard, Mark Thomas

    2007-11-01

    The primary objective of this report is to provide the technical basis for development of recommendations for updates to the NUREG-1465 Source Term for BWRs that will extend its applicability to accidents involving high burnup (HBU) cores. However, a secondary objective is to re-examine the fundamental characteristics of the prescription for fission product release to containment described by NUREG-1465. This secondary objective is motivated by an interest to understand the extent to which research into the release and behaviors of radionuclides under accident conditions has altered best-estimate calculations of the integral response of BWRs to severe core damage sequences and the resulting radiological source terms to containment. This report, therefore, documents specific results of fission product source term analyses that will form the basis for the HBU supplement to NUREG-1465. However, commentary is also provided on observed differences between the composite results of the source term calculations performed here and those reflected NUREG-1465 itself.

  5. Theoretical and experimental studies of gaseous laser pumped by a twin-core fast burst reactor

    NASA Astrophysics Data System (ADS)

    Barzilov, Alexander P.; Bokhovko, Mikhail V.; Gulevich, Andrey V.; Dyachenko, Peter P.; Kachanov, Boris V.; Kononov, Victor N.; Kukharchuk, Oleg F.; Pashin, Evgeny A.; Regushevsky, Victor I.; Zrodnikov, Anatoly V.

    1997-04-01

    Experimental setup configuration and lasing experiments results on fission fragments pumping of gas lasers are presented. The IPPE's fast burst reactor BARS-6 has been used as a neutron source for nuclear pumping of lasers. Experimental results on nuclear pumping of «master oscillator-amplifier» pulse laser system and energy parameters of 1.73 ?m 5d-6p Xel transition of Ar-Xe laser and amplifier are presented. Neutron characteristics of system were compared with computed ones. It was shown that experimental and calculated results are in a good agreement.

  6. Abstract--Identity verification using iris biometrics is degraded if portions of the iris texture have a distorted

    E-print Network

    Bowyer, Kevin W.

    and with specular highlights through improved segmentation of the iris region. Our approach assumes that some local corresponding regions of local distortion in the iris code derived from the image. We introduce an approach to detect such regions of local distortion in the iris code through analysis of the iris code matching

  7. Thin gauge nickel-iron cores--Specification for use in high-power pulse transformers and saturable reactors with reset

    Microsoft Academic Search

    P. C. HOOKE; D. Driver; R. Major

    1974-01-01

    To specify a core for use as a power pulse transformer or pulse reactor it is necessary to develop a test method closely related to the operating conditions in a resetting line modulator. The important magnetic pulse properties for both applications are given, and the apparatus used to measure them described.

  8. Large Core Code Evaluation Working Group Benchmark Problem Four: neutronics and burnup analysis of a large heterogeneous fast reactor. Part 1. Analysis of benchmark results. [LMFBR

    Microsoft Academic Search

    C. L. Cowan; R. Protsik; J. W. Lewellen

    1984-01-01

    The Large Core Code Evaluation Working Group Benchmark Problem Four was specified to provide a stringent test of the current methods which are used in the nuclear design and analyses process. The benchmark specifications provided a base for performing detailed burnup calculations over the first two irradiation cycles for a large heterogeneous fast reactor. Particular emphasis was placed on the

  9. Three-dimensional thermal-hydraulic analysis of wire-wrapped rods in liquid-metal fast breeder reactor core assemblies. [FATHOM

    Microsoft Academic Search

    M. C. Chuang; M. D. Carelli; C. W. Bach; J. S. Killimayer

    1977-01-01

    A study is presented to determine the detailed coolant velocity and temperature profile around the entire rod circumference in liquid-metal fast breeder reactor (LMFBR) core assemblies as well as the detailed radial and circumferential temperature profile in the rod. The digital computer code FATHOM-360 developed to perform the above calculations is described. Fuel, radial blanket, and control assembly rods (both

  10. Neutronics methods, models, and applications at the Idaho National Engineering Laboratory for the advanced neutron source reactor three-element core design

    Microsoft Academic Search

    C. A. Wemple; B. G. Schnitzler; J. M. Ryskamp

    1995-01-01

    A summary of the methods and models used to perform neutronics analyses on the Advanced Neutron Source reactor three-element core design is presented. The applications of the neutral particle Monte Carlo code MCNP are detailed, as well as the expansion of the static role of MCNP to analysis of fuel cycle depletion calculations. Results to date of these applications are

  11. RESULTS OF TENSILE TESTS PERFORMED ON MATERIALS EXPOSED IN THE HOMOGENEOUS REACTOR EXPERIMENT NO. 2 BLANKET AND LOW-FLUX-CORE REGION

    Microsoft Academic Search

    Prislinger

    1962-01-01

    Tensile specimens of various alloys of zirconium and titanium and a ; variety of austenitic stainless steels and Incoloy were exposed in the blanket ; and low-flux core regions of the Homogeneous Reactor Experiment (HRE) No. 2. The ; purpose of this exposure was to determine the effect of HRE-2 environment on ; mechanical properties and to provide corrosion data.

  12. Use of Axially Graded Burnable Boron for Hot-Spot Temperature Reduction in a Pressurized Water Reactor Core

    SciTech Connect

    Segev, M.; Galperin, A.; Schwageraus, E. [Ben Gurion University of the Negev (Israel)

    2000-07-15

    Shortly after the loading of a pressurized water reactor (PWR) core, the axial power distribution in fresh fuel has already attained the characteristic, almost flat shape, typical of burned fuel. At beginning of cycle (BOC), however, the axial distribution is centrally peaked. In assemblies hosting uniform burnable boron rods, this BOC peaking is even more pronounced. A reduction in the axial peaking is today often achieved by shortening the burnable boron rods by some 30 cm at each edge.It is shown that a two-zone grading of the boron rod leads, in a representative PWR cycle, to a reduction of the hot-spot temperature of {approx}70 deg. C, compared with the nongraded case. However, with a proper three-zone grading of the boron rod, an additional 20 deg. C may be cut off the hot-spot temperature. Further, with a slightly skewed application of this three-zone grading, an additional 50 deg. C may be cut off.The representative PWR cycle studied was cycle 11 of the Indian Point 2 station, with a simplification in the number of fuel types and in the burnup distribution. The analysis was based on a complete three-dimensional burnup calculation. The code system was ELCOS, with BOXER as an assembly code for the generation of burnup-dependent cross sections and SILWER as a three-dimensional core code with thermal-hydraulic feedback.

  13. Fuel Summary for Peach Bottom Unit 1 High-Temperature Gas-Cooled Reactor Cores 1 and 2

    SciTech Connect

    Karel I. Kingrey

    2003-04-01

    This fuel summary report contains background and summary information for the Peach Bottom Unit 1, High-Temperature, Gas-Cooled Reactor Cores 1 and 2. This report contains detailed information about the fuel in the two cores, the Peach Bottom Unit 1 operating history, nuclear parameters, physical and chemical characteristics, and shipping and storage canister related data. The data in this document have been compiled from a large number of sources and are not qualified beyond the qualification of the source documents. This report is intended to provide an overview of the existing data pertaining to spent fuel management and point to pertinent reference source documents. For design applications, the original source documentation must be used. While all referenced sources are available as records or controlled documents at the Idaho National Engineering and Environmental Laboratory (INEEL), some of the sources were marked as informal or draft reports. This is noted where applicable. In some instances, source documents are not consistent. Where they are known, this document identifies those instances and provides clarification where possible. However, as stated above, this document has not been independently qualified and such clarifications are only included for information purposes. Some of the information in this summary is available in multiple source documents. An effort has been made to clearly identify at least one record document as the source for the information included in this report.

  14. Simulation of in-core neutron noise measurements for axial void profile reconstruction in boiling water reactors

    SciTech Connect

    Dykin, V.; Pazsit, I. [Chalmers Univ. of Technology, Div. of Nuclear Engineering, Dept. of Applied Physics, SE-412 96 Gothenburg (Sweden)

    2012-07-01

    A possibility to reconstruct the axial void profile from the simulated in-core neutron noise which is caused by density fluctuations in a Boiling Water Reactor (BWR) heated channel is considered. For this purpose, a self-contained model of the two-phase flow regime is constructed which has quantitatively and qualitatively similar properties to those observed in real BWRs. The model is subsequently used to simulate the signals of neutron detectors induced by the corresponding perturbations in the flow density. The bubbles are generated randomly in both space and time using Monte-Carlo techniques. The axial distribution of the bubble production is chosen such that the mean axial void fraction and void velocity follow the actual values of BWRs. The induced neutron noise signals are calculated and then processed by the standard signal analysis methods such as Auto-Power Spectral Density (APSD) and Cross-Power Spectral Density (CPSD). Two methods for axial void and velocity profiles reconstruction are discussed: the first one is based on the change of the break frequency of the neutron auto-power spectrum with axial core elevation, while the second refers to the estimation of transit times of propagating steam fluctuations between different axial detector positions. This paper summarizes the principles of the model and presents a numerical testing of the qualitative applicability to estimate the required parameters for the reconstruction of the void fraction profile from the neutron noise measurements. (authors)

  15. SUMER-IRIS Observations of AR11875

    NASA Astrophysics Data System (ADS)

    Schmit, Donald; Innes, Davina

    2014-05-01

    We present results of the first joint observing campaign of IRIS and SOHO/SUMER. While the IRIS datasets provide information on the chromosphere and transition region, SUMER provides complementary diagnostics on the corona. On 2013-10-24, we observed an active region, AR11875, and the surrounding plage for approximately 4 hours using rapid-cadence observing programs. These datasets include spectra from a small C -class flare which occurs in conjunction with an Ellerman-bomb type event. Our analysis focusses on how the high spatial resolution and slit jaw imaging capabilities of IRIS shed light on the unresolved structure of transient events in the SUMER catalog.

  16. Development and experimental validation of a calculation scheme for nuclear heating evaluation in the core of the OSIRIS material testing reactor

    SciTech Connect

    Malouch, F. [Saclay Center CEA, DEN/DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette Cedex (France)

    2011-07-01

    The control of the temperature in material samples irradiated in a material testing reactor requires the knowledge of the nuclear heating caused by the energy deposition by neutrons and photons interacting in the irradiation device structures. Thus, a neutron-photonic three-dimensional calculation scheme has been developed to evaluate the nuclear heating in experimental devices irradiated in the core of the OSIRIS MTR reactor (CEA/Saclay Center). The aim is to obtain a predictive tool for the nuclear heating estimation in irradiation devices. This calculation scheme is mainly based on the TRIPOLI-4 three-dimensional continuous-energy Monte Carlo transport code, developed by CEA (Saclay Center). An experimental validation has been carried out on the basis of nuclear heating measurements performed in the OSIRIS core. After an overview of the experimental devices irradiated in the OSIRIS reactor, we present the calculation scheme and the first results of the experimental validation. (authors)

  17. Particle image velocimetry measurements in a representative gas-cooled prismatic reactor core model for the estimation of bypass flow

    NASA Astrophysics Data System (ADS)

    Conder, Thomas E.

    Core bypass flow is considered one of the largest contributors to uncertainty in fuel temperature within the Modular High Temperature Gas-cooled Reactor (MHTGR). It refers to the coolant that navigates through the interstitial regions between the graphite fuel blocks instead of traveling through the designated coolant channels. These flows are of concern because they reduce the desired flow rates in the coolant channels, and thereby have significant influence on the maximum fuel element and coolant exit temperatures. Thus, accurate prediction of the bypass flow is important because it directly impacts core temperature, influencing the life and efficiency of the reactor. An experiment was conducted at Idaho National Laboratory to quantify the flow in the coolant channels in relation to the interstitial gaps between fuel blocks in a representative MHTGR core. Particle Image Velocimetry (PIV) was used to measure the flow fields within a simplified model, which comprised of a stacked junction of six partial fuel blocks with nine coolant tubes, separated by a 6mm gap width. The model had three sections: The upper plenum, upper block, and lower block. Model components were fabricated from clear, fused quartz where optical access was needed for the PIV measurements. Measurements were taken in three streamwise locations: in the upper plenum and in the midsection of the large and small fuel blocks. A laser light sheet was oriented parallel to the flow, while velocity fields were measured at millimeter intervals across the width of the model, totaling 3,276 PIV measurement locations. Inlet conditions were varied to incorporate laminar, transition, and turbulent flows in the coolant channels---all which produced laminar flow in the gap and non-uniform, turbulent flow in the upper plenum. The images were analyzed to create vector maps, and the data was exported for processing and compilation. The bypass flow was estimated by calculating the flow rates through the coolant channels and gap, then forming a ratio of the results. It was found that the gap consumed about 6.9-15.8% of the total flow for a channel Reynolds number between 1,700 and 4,600, where the flow distribution amid the coolant channels varied less than 4.6%.

  18. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ...cooling systems for light-water nuclear power reactors. 50.46 Section 50...systems for light-water nuclear power reactors. (a)(1)(i) Each... (2) The Director of Nuclear Reactor Regulation may impose...

  19. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ...cooling systems for light-water nuclear power reactors. 50.46 Section 50...systems for light-water nuclear power reactors. (a)(1)(i) Each... (2) The Director of Nuclear Reactor Regulation may impose...

  20. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ...cooling systems for light-water nuclear power reactors. 50.46 Section 50...systems for light-water nuclear power reactors. (a)(1)(i) Each... (2) The Director of Nuclear Reactor Regulation may impose...

  1. Porcine cadaver iris model for iris heating during corneal surgery with a femtosecond laser

    NASA Astrophysics Data System (ADS)

    Sun, Hui; Fan, Zhongwei; Wang, Jiang; Yan, Ying; Juhasz, Tibor; Kurtz, Ron

    2015-03-01

    Multiple femtosecond lasers have now been cleared for use for ophthalmic surgery, including for creation of corneal flaps in LASIK surgery. Preliminary study indicated that during typical surgical use, laser energy may pass beyond the cornea with potential effects on the iris. As a model for laser exposure of the iris during femtosecond corneal surgery, we simulated the temperature rise in porcine cadaver iris during direct illumination by the femtosecond laser. Additionally, ex-vivo iris heating due to femtosecond laser irradiation was measured with an infrared thermal camera (Fluke corp. Everett, WA) as a validation of the simulation.

  2. Three-Dimensional Full Core Power Calculations for Pressurized Water Reactors

    SciTech Connect

    Evans, Thomas M [ORNL] [ORNL; Davidson, Gregory G [ORNL] [ORNL; Slaybaugh, Rachel N [ORNL] [ORNL

    2010-01-01

    We have implemented a new multilevel parallel decomposition in the Denovo discrete ordinates radiation transport code. In concert with Krylov subspace iterative solvers, the multilevel decomposition allows concurrency over energy in addition to space-angle. The original space-angle partitioning in Denovo placed an eective limit on the scalability of the transport solver that was highly dependent on the problem size. The added phase-space concurrency combined with the high-performance Krylov solvers has enabled weak scaling to 100K cores on the Jaguar XT5 supercomputer. Furthermore, the multilevel decomposition provides enough concurrency to scale to exascale computing and beyond.

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

    SciTech Connect

    Williams, W.C.

    2002-08-01

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

  4. Results of recent reactor-material tests on dispersal of oxide fuel from a disrupted core

    SciTech Connect

    Spencer, B.W.; Wilson, R.J.; Vetter, D.L.; Erickson, E.G.; Dewey, G.

    1985-01-01

    The results of experimental investigations and related analyses are reported addressing the dispersal of molten oxide fuel from a disrupted core via various available pathways for the CRBR system. These investigations included the GAPFLOW tests in which pressure-driven and gravity drainage tests were performed using dispersal pathways mocking up the intersubassembly gaps, the CAMEL C6 and C7 tests in which molten fuel entered sodium-filled control assembly ducts under prototypic thermal-hydraulic conditions, and the Lower Internals Drainage (LID) tests in which molten fuel drained downward through simulated below-core structure (orifice plate stacks) as the bottom of control assembly ducts. The results of SHOTGUN tests addressing basic freezing of molten UO/sub 2/ and UO/sub 2//metal mixtures flowing through circular tubes are also reported. Test results have invariably shown the existance of stable UO/sub 2/ crusts on the inside surfaces of the flow paths. Appreciable removal of fuel was indicated prior to freezing-induced immobilization. Application of heat transfer models based upon the presence of stable, insulating fuel crusts tends to overpredict the removal process.

  5. IRIS Launch Animation - Duration: 108 seconds.

    NASA Video Gallery

    This animation demonstrates the launch and deployment of NASA's Interface Region Imaging Spectrograph (IRIS) mission satellite via a Pegasus rocket. The launch is scheduled for June 26, 2013 from V...

  6. [Iridoschisis, a special form of iris atrophy].

    PubMed

    Agard, E; Malcles, A; El Chehab, H; Ract-Madoux, G; Swalduz, B; Aptel, F; Denis, P; Dot, C

    2013-04-01

    Iridoschisis is a rare degenerative disease characterized by the separation of the anterior iris stroma from the posterior layer. The anterior layer splits into strands, and the free ends float freely in the anterior chamber. We report the case of a 57-year-old man, in whom we incidentally discovered isolated unilateral iris atrophy. The patient had no history of the common causes of atrophy (herpes, pigment dispersion, ocular trauma, etc.). During follow-up, the atrophy gradually worsened, with an increase in the number and bilaterality of the lesions. Ultrasound biomicroscopy (UBM) and optical coherence tomography (OCT) of anterior chamber showed thinning of the anterior iris and cleavage of the iris into two layers, an imaging result which, to our knowledge, has not yet been reported in the literature. Familiarity with iridoschisis is important, due to its frequent association with glaucoma, so that appropriate screening can be carried out at the time of diagnosis and on follow-up. PMID:23261208

  7. ZPR-6 assembly 7 high {sup 240}Pu core experiments : a fast reactor core with mixed (Pu,U)-oxide fuel and a centeral high{sup 240}Pu zone.

    SciTech Connect

    Lell, R. M.; Morman, J. A.; Schaefer, R.W.; McKnight, R.D.; Nuclear Engineering Division

    2009-02-23

    ZPR-6 Assembly 7 (ZPR-6/7) encompasses a series of experiments performed at the ZPR-6 facility at Argonne National Laboratory in 1970 and 1971 as part of the Demonstration Reactor Benchmark Program (Reference 1). Assembly 7 simulated a large sodium-cooled LMFBR with mixed oxide fuel, depleted uranium radial and axial blankets, and a core H/D near unity. ZPR-6/7 was designed to test fast reactor physics data and methods, so configurations in the Assembly 7 program were as simple as possible in terms of geometry and composition. ZPR-6/7 had a very uniform core assembled from small plates of depleted uranium, sodium, iron oxide, U{sub 3}O{sub 8} and Pu-U-Mo alloy loaded into stainless steel drawers. The steel drawers were placed in square stainless steel tubes in the two halves of a split table machine. ZPR-6/7 had a simple, symmetric core unit cell whose neutronic characteristics were dominated by plutonium and {sup 238}U. The core was surrounded by thick radial and axial regions of depleted uranium to simulate radial and axial blankets and to isolate the core from the surrounding room. The ZPR-6/7 program encompassed 139 separate core loadings which include the initial approach to critical and all subsequent core loading changes required to perform specific experiments and measurements. In this context a loading refers to a particular configuration of fueled drawers, radial blanket drawers and experimental equipment (if present) in the matrix of steel tubes. Two principal core configurations were established. The uniform core (Loadings 1-84) had a relatively uniform core composition. The high {sup 240}Pu core (Loadings 85-139) was a variant on the uniform core. The plutonium in the Pu-U-Mo fuel plates in the uniform core contains 11% {sup 240}Pu. In the high {sup 240}Pu core, all Pu-U-Mo plates in the inner core region (central 61 matrix locations per half of the split table machine) were replaced by Pu-U-Mo plates containing 27% {sup 240}Pu in the plutonium component to construct a central core zone with a composition closer to that in an LMFBR core with high burnup. The high {sup 240}Pu configuration was constructed for two reasons. First, the composition of the high {sup 240}Pu zone more closely matched the composition of LMFBR cores anticipated in design work in 1970. Second, comparison of measurements in the ZPR-6/7 uniform core with corresponding measurements in the high {sup 240}Pu zone provided an assessment of some of the effects of long-term {sup 240}Pu buildup in LMFBR cores. The uniform core version of ZPR-6/7 is evaluated in ZPR-LMFR-EXP-001. This document only addresses measurements in the high {sup 240}Pu core version of ZPR-6/7. Many types of measurements were performed as part of the ZPR-6/7 program. Measurements of criticality, sodium void worth, control rod worth and reaction rate distributions in the high {sup 240}Pu core configuration are evaluated here. For each category of measurements, the uncertainties are evaluated, and benchmark model data are provided.

  8. Improved iris localization by using wide and narrow field of view cameras for iris recognition

    NASA Astrophysics Data System (ADS)

    Kim, Yeong Gon; Shin, Kwang Yong; Park, Kang Ryoung

    2013-10-01

    Biometrics is a method of identifying individuals by their physiological or behavioral characteristics. Among other biometric identifiers, iris recognition has been widely used for various applications that require a high level of security. When a conventional iris recognition camera is used, the size and position of the iris region in a captured image vary according to the X, Y positions of a user's eye and the Z distance between a user and the camera. Therefore, the searching area of the iris detection algorithm is increased, which can inevitably decrease both the detection speed and accuracy. To solve these problems, we propose a new method of iris localization that uses wide field of view (WFOV) and narrow field of view (NFOV) cameras. Our study is new as compared to previous studies in the following four ways. First, the device used in our research acquires three images, one each of the face and both irises, using one WFOV and two NFOV cameras simultaneously. The relation between the WFOV and NFOV cameras is determined by simple geometric transformation without complex calibration. Second, the Z distance (between a user's eye and the iris camera) is estimated based on the iris size in the WFOV image and anthropometric data of the size of the human iris. Third, the accuracy of the geometric transformation between the WFOV and NFOV cameras is enhanced by using multiple matrices of the transformation according to the Z distance. Fourth, the searching region for iris localization in the NFOV image is significantly reduced based on the detected iris region in the WFOV image and the matrix of geometric transformation corresponding to the estimated Z distance. Experimental results showed that the performance of the proposed iris localization method is better than that of conventional methods in terms of accuracy and processing time.

  9. New Neutron Simulation Capabilities Provided by the Sandia Pulse Reactor-III (SPR-III) and the Upgraded Annular Core Pulse Reactor (ACPR)

    Microsoft Academic Search

    Larry M. Choate; Theodore R. Schmidt

    1978-01-01

    The paper briefly describes the nuclear reactor facilities at Sandia Laboratories which are used for simulating nuclear weapon produced neutron environments. These reactor facilities are used principally in support of continuing R and D programs for the Department of Energy\\/Office of Military Application (DOE\\/OMA) in studying the effects of radiation on nuclear weapon systems and components. As such, the reactors

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

    SciTech Connect

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

    2006-07-31

    Gas-cooled fast reactor (GFR) designs are being developed to meet Gen IV goals of sustainability, economics, safety and reliability, and proliferation resistance and physical protection as part of an International Generation IV Nuclear Energy System Research Initiative effort. Different organizations are involved in the development of a variety of GFR design concepts. The current analysis has focused on the evaluation of low-pressure drop, pin-core designs with favorable passive cooling properties. Initial evaluation of the passive cooling safety case for the GFR during depressurized decay heat removal accidents with concurrent loss of electric power have resulted in requirements for a reduction of core power density to the 100 w/cc level and a low core pressure drop of 0.5 bars. Additional design constraints and the implementation of their constraints are evaluated in this study to enhance and passive cooling properties of the reactor. Passive cooling is made easier by a flat radial distribution of the decay heat. One goal of this study was to evaluate the radial power distribution and determine to what extent it can be flattened, since the decay heat is nearly proportional to the fission power at shutdown. In line with this investigation of the radial power profile, an assessment was also made of the control rod configuration. The layout provided a large number of control rod locations with a fixed area provided for control rods. The number of control rods was consistent with other fast reactor designs. The adequacy of the available control rod locations was evaluated. Future studies will be needed to optimize the control rod designs and evaluate the shutdown system. The case for low pressure drop core can be improved by the minimization of pressure drop sources such as the number of required fuel spacers in the subassembly design and by the details of the fuel pin design. The fuel pin design is determined by a number of neutronic, thermal-hydraulic (gas dynamics) and fuel performance considerations. For the purposes of this study, the starting point is the fuel pin design established by the CEA-ANL/US I-NERI collaboration project for the selected 2400 MWt large rector option. Structural mechanics factors are now included in the design assessment. In particular, thermal bowing establishes a bound on the minimum of fuel pin spacers required in each fuel subassembly to prevent the local flow channel restrictions and pin-to-pin mechanical interaction. There are also fabrication limitations on the maximum length of SiC fuel pin cladding which can be manufactured. This geometric limitation effects the minimum ceramic clad thickness which can be produced. This ties into the fuel pin heat transfer and temperature thresholds. All these additional design factors were included in the current iteration on the subassembly design to produce a lower core pressure drop. A more detailed definition of the fuel pin/subassembly design is proposed here to meet these limitations. This subassembly design was then evaluated under low pressure natural convection conditions to assess its acceptability for the decay heat removal accidents. A number of integrated decay heat removal (DHR) loop plus core calculations were performed to scope the thermal-hydraulic response of the subassembly design to the accidents of interest. It is evident that there is a large sensitivity to the guard containment back pressure for these designs. The implication of this conclusion and possible design modifications to reduce this sensitivity will be explored under the auspices of the International GENIV GFR collaborative R&D plan. Chapter 2 describes the core reference design for the 2,400 MWt GFR being evaluated. The methodology, modeling, and codes used in the analysis of the fuel pin structural behavior are described in Chapter 3. Chapter 4 provides the result of the thermal-hydraulic study of the assembly design for the accidents of interest. An evaluation of the performance and control rod reactivity control is also presented in Chapter 2.

  11. Some features of the effect the pH value and the physicochemical properties of boric acid have on mass transfer in a VVER reactor's core

    NASA Astrophysics Data System (ADS)

    Gavrilov, A. V.; Kritskii, V. G.; Rodionov, Yu. A.; Berezina, I. G.

    2013-07-01

    Certain features of the effect of boric acid in the reactor coolant of nuclear power installations equipped with a VVER-440 reactor on mass transfer in the reactor core are considered. It is determined that formation of boric acid polyborate complexes begins under field conditions at a temperature of 300°C when the boric acid concentration is equal to around 0.065 mol/L (4 g/L). Operations for decontaminating the reactor coolant system entail a growth of corrosion product concentration in the coolant, which gives rise to formation of iron borates in the zones where subcooled boiling of coolant takes place and to the effect of axial offset anomalies. A model for simulating variation of pressure drop in a VVER-440 reactor's core that has invariable parameters during the entire fuel campaign is developed by additionally taking into account the concentrations of boric acid polyborate complexes and the quantity of corrosion products (Fe, Ni) represented by the ratio of their solubilities.

  12. Experimental and Numerical Observations of Hydrate Reformation during Depressurization in a Core-Scale Reactor

    SciTech Connect

    Seol, Yongkoo; Myshakin, Evgeniy

    2011-01-01

    Gas hydrate has been predicted to reform around a wellbore during depressurization-based gas production from gas hydrate-bearing reservoirs. This process has an adverse effect on gas production rates and it requires time and sometimes special measures to resume gas flow to producing wells. Due to lack of applicable field data, laboratory scale experiments remain a valuable source of information to study hydrate reformation. In this work, we report laboratory experiments and complementary numerical simulations executed to investigate the hydrate reformation phenomenon. Gas production from a pressure vessel filled with hydrate-bearing sand was induced by depressurization with and without heat flux through the boundaries. Hydrate decomposition was monitored with a medical X-ray CT scanner and pressure and temperature measurements. CT images of the hydrate-bearing sample were processed to provide 3-dimensional data of heterogeneous porosity and phase saturations suitable for numerical simulations. In the experiments, gas hydrate reformation was observed only in the case of no-heat supply from surroundings, a finding consistent with numerical simulation. By allowing gas production on either side of the core, numerical simulations showed that initial hydrate distribution patterns affect gas distribution and flow inside the sample. This is a direct consequence of the heterogeneous pore network resulting in varying hydraulic properties of the hydrate-bearing sediment.

  13. The response of ex-core neutron detectors to large- and small-break loss-of-coolant accidents in pressurized water reactors

    SciTech Connect

    Okyere, E.W. (City Univ. of New York, Staten Island, NY (US)); Baratta, A.J.; Jester, W.A. (Pennsylvania State Univ., University Park, PA (United States). Dept. of Nuclear Engineering)

    1991-12-01

    This paper reports on a variety of water level measurement systems that are proposed to resolve the problem of reactor vessel level measurement. Two such systems, the heated thermocouple and the multiple differential pressure cell system, are used commercially. A third system based on ex-core neutron detectors was tested at the Pennsylvania State University Breazeale nuclear reactor facility and at the Idaho National Engineering Laboratory Loss-of-Fluid Test Facility. Results of these tests show that such a system is sensitive to both large- and small-break loss-of-coolant accidents and to voiding in the upper plenum of the vessel.

  14. Advanced MOX Core Design Study of Sodium Cooled Reactors in Current Feasibility Study on Commercialized Fast Reactor Cycle Systems in Japan

    SciTech Connect

    Mizuno, T.; Niwa, H. [Japan Nuclear Cycle development institute, O-arai Engineering Center, 4002 Narita-cho, O-arai-machi, Higashi-ibaraki-gun, Ibaraki-ken, 311-1393 (Japan)

    2002-07-01

    The Sodium cooled MOX core design studies are performed with the target burnup of 150 GWd/t and measures against the recriticality issues in core disruptive accidents (CDAs). Four types of core are comparatively studied in viewpoints of core performance and reliability. Result shows that all the types of core satisfy the target and that the homogeneous core with axial blanket partial elimination subassembly is the most superior concept in case the effectiveness of measures against recriticality issues by the axial blanket partial elimination is assured. (authors)

  15. Methodology for Identification of the Coolant Thermalhydraulic Regimes in the Core of Nuclear Reactors

    SciTech Connect

    Sharaevsky, L.G.; Sharaevskaya, E.I.; Domashev, E.D.; Arkhypov, A.P.; Kolochko, V.N. [Ministry of Fuel and Energy of Ukraine, 30, Khreshchatyk str., Kyiv 01601 (Ukraine)

    2002-07-01

    The paper deals with one of the acute for the nuclear energy problem of accident regimes of NPPs recognition diagnostics using noise signal diagnostics methodology. The methodology intends transformation of the random noise signals of the main technological parameters at the exit of a nuclear facility (neutron flow, dynamic pressure etc.) which contain the important information about the technical status of the equipment. The effective algorithms for identification of random processes wore developed. After proper transformation its were considered as multidimensional random vectors. Automatic classification of these vectors in the developed algorithms is realized on the basis of the probability function in particular Bayes classifier and decision functions. Till now there no mathematical models for thermalhydraulic regimes of fuel assemblies recognition on the acoustic and neutron noises parameters in the core of nuclear facilities. The two mathematical models for analysis of the random processes submitted to the automatic classification is proposed, i.e. statistical (using Bayes classifier of acoustic spectral density diagnosis signals) and geometrical (on the basis of formation in the featured space of dividing hyper-plane). The theoretical basis of the bubble boiling regimes in the fuel assemblies is formulated as identification of these regimes on the basis of random parameters of auto spectral density of acoustic noise (ASD) measured in the fuel assemblies (dynamic pressure in the upper plenum in the paper). The elaborated algorithms allow recognize realistic status of the fuel assemblies. For verification of the proposed mathematical models the analysis of experimental measurements was carried out. The research of the boiling onset and definition of the local values of the flow parameters in the seven-beam fuel assembly (length of 1.3 m, diameter of 6 mm) have shown the correct identification of the bubble boiling regimes. The experimental measurements on real WWER core assemblies were analysed as well. On the basis of model of Bayes classifier for bubble structure of two-phase flow in fuel assemblies of WWER-440 (intends usage of 28 dimensional accidental realizations of ASD of neutron noise) the reliable identification of the pointed regimes of fuel assemblies in WWERs up to 98% was obtained. On the basis of geometrical mathematical model of identification at essentially more limited volume of teaching sampling the recognition of ASD realizations of the neutron noise of the same both dimensions and quantity of the reliability of correct identification of these parameters was up to 92%. The recognition of the pointed thermalhydraulic parameters was carried out on the basis of experimental research of ASD of acoustic noise parameters of the experimental fuel assembly with electrically heated imitators using the two recognition models - statistical and geometrical. It confirmed high efficiency of the algorithms developed. The average reliability of identification of the first vapor bubbles activation regime at the heat transfer surface was not low then 90%. (authors)

  16. Bow resistant structural member for fuel assemblies in non-control rod locations of a nuclear reactor core

    SciTech Connect

    Wilson, J.F.; Gjertsen, R.K.; Ferrari, H.M.

    1987-08-04

    This patent describes a fuel assembly for use at non-control rod locations of a nuclear reactor core, the fuel assembly including top and bottom nozzles and longitudinal structural members extending between and attached to the nozzles for forming the assembly into an integral unitary structure. At least certain of the structural members includes an elongated hollow cladding tube extending between the top and bottom nozzles and means secured to opposite ends of the tube for hermetically sealing the tube and attaching it to the top and bottom nozzles. The improvement comprises: (a) a quantity of irradiation-induced creep resistant material disposed within the tube; and (b) pretensioning means positioned within the tube for applying a predetermined axially-directed compressive load to the creep resistant material and reacting the load so as to axially preload the tube in a state of pretension having a magnitude sufficient to substantially counteract an axial load typically transmitted through the unitary structure of the fuel assembly. This greatly reduces the compressive stress in the tube of the structural member.

  17. Review of cystic and solid tumors of the iris

    PubMed Central

    Shields, Carol L.; Shields, Patrick W.; Manalac, Janet; Jumroendararasame, Chaisiri; Shields, Jerry A.

    2013-01-01

    Iris tumors are broadly classified into cystic or solid lesions. The cystic lesions arise from iris pigment epithelium (IPE) or iris stroma. IPE cysts classically remain stable without need for intervention. Iris stromal cyst, especially those in newborns, usually requires therapy of aspiration, possibly with alcohol-induced sclerosis, or surgical resection. The solid tumors included melanocytic and nonmelanocytic lesions. The melanocytic iris tumors include freckle, nevus (including melanocytoma), Lisch nodule, and melanoma. Information from a tertiary referral center revealed that transformation of suspicious iris nevus to melanoma occurred in 4% by 10 years and 11% by 20 years. Risk factors for transformation of iris nevus to melanoma can be remembered using the ABCDEF guide as follows: A=age young (<40 years), B=blood (hyphema) in anterior chamber, C=clock hour of mass inferiorly, D=diffuse configuration, E=ectropion, F=feathery margins. The most powerful factors are diffuse growth pattern and hyphema. Tumor seeding into the anterior chamber angle and onto the iris stroma are also important. The nonmelanocytic iris tumors are relatively uncommon and included categories of choristomatous, vascular, fibrous, neural, myogenic, epithelial, xanthomatous, metastatic, lymphoid, leukemic, secondary, and non-neoplastic simulators. Overall, the most common diagnoses in a clinical series include nevus, IPE cyst, and melanoma. In summary, iris tumors comprise a wide spectrum including mostly iris nevus, IPE cyst, and iris melanoma. Risk factors estimating transformation of iris nevus to melanoma can be remembered by the ABCDEF guide. PMID:24379549

  18. Region-based SIFT approach to iris recognition

    NASA Astrophysics Data System (ADS)

    Belcher, Craig; Du, Yingzi

    2009-01-01

    Traditional iris recognition systems transfer iris images to polar (or log-polar) coordinates and have performed very well on data that tends to have a centered gaze. The patterns of an iris are part of a 3-D structure that is captured as a two-dimensional (2-D) image and cooperative iris recognition systems are capable of correctly matching these 2-D representations of iris features. However, when the gaze of an eye changes with respect to the camera lens, many times the size, shape, and detail of iris patterns will change as well and cannot be matched to enrolled images using traditional methods. Additionally, the transformation of off-angle eyes to polar coordinates becomes much more challenging and noncooperative iris algorithms will require a different approach. The direct application of the scale-invariant feature transform (SIFT) method would not work well for iris recognition because it does not take advantage of the characteristics of iris patterns. We propose the region-based SIFT approach to iris recognition. This new method does not require polar transformation, affine transformation or highly accurate segmentation to perform iris recognition and is scale invariant. This method was tested on the iris challenge evaluation (ICE), WVU and IUPUI noncooperative databases and results show that the method is capable of cooperative and noncooperative iris recognition.

  19. QUALITY-BASED FUSION FOR MULTICHANNEL IRIS RECOGNITION Mayank Vatsa1

    E-print Network

    Ross, Arun Abraham

    the recognition accuracy using color iris images character- ized by three spectral channels - Red, Green and Blue the use of color iris images in conjunction with their NIR counterparts. Index Terms-- Color iris the feasibility Color Iris Image Grayscale Iris Image Red Channel Green Channel Blue Channel Fig. 1. A color iris

  20. IRI and GPS Variations Over Ilorin Nigeria

    NASA Astrophysics Data System (ADS)

    Okonkwo, Perpetua; Okoh, Daniel

    Abstract Diurnal and day-to-day variations of Vertical Total Electron Content (VTEC) over an equatorial region (Ilorin, Nigeria; Geographic 8.500N, 4.550E; Geomagnetic 10.600N, 78.410E) is presented in this paper using data from the IRI model and from the AFRL-SCINDA (Air Force Research Laboratory - Scintillation Network Decision Aid) GPS receiver installed at the Ilorin station. A comparison between VTEC data from the two sources is also presented since a major concern in the work is to use available GPS-TEC data for year 2010 to evaluate the performance of the IRI model in TEC prediction over the region, and to therefore inform a proposed use of the IRI model in TEC modeling over the African region. Our results show generally good comparisons between the IRI TEC predictions and the GPS TEC measurements, results from the comparisons on diurnal basis were, as expected, better than those on day-to-day basis. The work also indicated that the lower TEC thresholds of the IRI predictions for the days observed occurred at around 04:00 UT while for the GPS measurements they occurred at around 05:00 UT.

  1. Results of analyses performed on concrete cores removed from floors and D-ring walls of the TMI2 reactor building

    Microsoft Academic Search

    C. V. McIsaac; C. M. Davis; J. T. Horan; D. G. Keefer

    1984-01-01

    The March 28, 1979 loss-of-coolant accident at Three Mile Island Unit 2 (TMI-2) exposed about 7200 m² of concrete surfaces within the Reactor building to liquid and vapor-phase contaminants. The majority of those surfaces are protected by coatings of epoxy-based, nuclear grade paints. during September 1983, seventeen high quality cores were extracted from the concrete floors and D-ring walls ar

  2. Design of a low enrichment, enhanced fast flux core for the Massachusetts Institute of Technology Research Reactor

    E-print Network

    Ellis, Tyler Shawn

    2009-01-01

    Worldwide, there is limited test reactor capacity to perform the required irradiation experiments on advanced fast reactor materials and fuel designs. This is particularly true in the U.S., which no longer has an operating ...

  3. Thermal hydraulic limits analysis for the MIT Research Reactor low enrichment uranium core conversion using statistical propagation of parametric uncertainties

    E-print Network

    Chiang, Keng-Yen

    2012-01-01

    The MIT Research Reactor (MITR) is evaluating the conversion from highly enriched uranium (HEU) to low enrichment uranium (LEU) fuel. In addition to the fuel element re-design from 15 to 18 plates per element, a reactor ...

  4. Thirty Years Supporting Portable Arrays: The IRIS Passcal Instrument Center

    NASA Astrophysics Data System (ADS)

    Beaudoin, B. C.; Anderson, K. R.; Bilek, S. L.; Woodward, R.

    2014-12-01

    Thirty years have passed since establishment of the IRIS Program for the Array Seismic Studies of the Continental Lithosphere (PASSCAL). PASSCAL was part of a coordinated plan proposed to the National Science Foundation (NSF) defining the instrumentation, data collection and management structure to support a wide range of research in seismology. The PASSCAL program has surpassed the early goal of 6000 data acquisition channels with a current inventory of instrumentation capable of imaging from the near surface to the inner core. Here we present the evolution of the PASSCAL program from instrument depot to full service community resource. PASSCAL has supported close to 1100 PI driven seismic experiments since its inception. Instruments from PASSCAL have covered the globe and have contributed over 7400 SEED stations and 242 assembled data sets to the IRIS Data Management Center in Seattle. Since the combination in 1998 of the Stanford and Lamont instrument centers into the single PASSCAL Instrument Center (PIC) at New Mexico Tech, the facility has grown in scope by adding the EarthScope Array Operations Facility in 2005, the incorporation of the EarthScope Flexible Array, and a Polar support group in 2006. The polar support group enhances portable seismic experiments in extremely harsh polar environments and also extends to special projects such as the Greenland Ice Sheet Monitoring Network (GLISN) and the recent development effort for Geophysical Earth Observatory for Ice Covered Environments (GEOICE). Through these support efforts the PIC has established itself as a resource for field practices, engineered solutions for autonomous seismic stations, and a pioneer in successful seismic recording in polar environments. We are on the cusp of a new generation of instrumentation driven in part by the academic community's desire to record unaliased wavefields in multiple frequency bands and industry's interest in utilizing lower frequency data. As part of the recently funded IRIS proposal to NSF for support of Seismological Facilities for the Advancement of Geoscience and EarthScope (SAGE), IRIS is developing plans for this new instrumentation that will ensure that the PASSCAL program continues to provide state-of-the-art observing capabilities into the coming decades.

  5. ISO Technical Specification for the Ionosphere -IRI Recent Activities

    NASA Astrophysics Data System (ADS)

    Bilitza, Dieter; Reinisch, Bodo; Tamara, Gulyaeva

    ISO Technical Specification TS 16457 recommends the International Reference Ionosphere (IRI) for the specification of ionospheric densities and temperatures. We review the latest develop-ments towards improving the IRI model and the newest version of the model IRI-2010. IRI-2010 includes several important improvements and additions. This presentation introduces these changes and discusses their benefits. The changes affect primarily the density profiles in the bottomside ionosphere and the density and height of the F2 peak, the point of highest density in the ionosphere. An important new addition to the model is the inclusion of auroral boundaries and their movement with magnetic activity. We will also discuss the status of other ongoing IRI activities and some of the recent applications of the IRI model. The homepage for the IRI project is at http://IRI.gsfc.nasa.gov/.

  6. Sector iris hemangioma in association with diffuse choroidal hemangioma.

    PubMed

    Shields, Carol L; Atalay, Hatice Tuba; Wuthisiri, Wadakarn; Levin, Alex V; Lally, Sara E; Shields, Jerry A

    2015-02-01

    Two patients referred for iris lesions were found to have sector hemangioma of the iris stroma in contiguity with diffuse choroidal hemangioma. Neither patient had other manifestations of Sturge-Weber syndrome. PMID:25727597

  7. MELPROG-POW/MOD1: A two-dimensional, mechanistic code for analysis of reactor core melt progression and vessel attack under severe accident conditions

    SciTech Connect

    Dosanjh, S.S. (ed.)

    1989-05-01

    The US Nuclear regulatory Commission has made the development of mechanistic models for severe accident progression a major priority. The purpose of these models is to provide detailed, best-estimate, coupled analyses of all the major phenomena involved in the reactor vessel and coolant system in the course of the accident. To meet this objective, the MELPROG computer code is being developed. This report describes the two-dimensional, pressurized water reactor (PWR) version of the MELPROG computer code, MELPROG/PWR-MOD1. Preliminary BWR work is described in this report. MELPROG is coupled to the TRAC-PF-1 RCS thermal-hydraulics code to provide an integrated analysis of the behavior of core, vessel, and reactor coolant systems during severe accidents. MELPROG treats core degradation and loss of geometry, debris formation, core melting, attack on supporting structures, slumping, melt/water interactions and vessel failure. The key element in MELPROG is the use of detailed modeling for the entire damage progression and failure sequence. Emphasis is also placed on the rates of hydrogen, steam and fission product formation, and transport to containment during the accident.

  8. Thermal-fluid assessment of the design options for reactor vessel cooling in a prismatic core VHTR

    Microsoft Academic Search

    Min-Hwan Kim; Nam-il Tak; Hong-Sik Lim

    2010-01-01

    The design of the reactor pressure vessel is an important issue in the VHTR design due to its high operating temperature. The extensive experience base in Light Water Reactor makes SA508\\/533 steel emerge as a strong candidate for the VHTR reactor vessel but requires maintaining the vessel temperature below the ASME code limit. To meet the temperature requirement, three types

  9. Jets and Bombs: Characterizing IRIS Spectra

    NASA Astrophysics Data System (ADS)

    Schmit, Donald; Innes, Davina

    2014-06-01

    For almost two decades, SUMER has provided an unique perspective on explosive events in the lower solar atmosphere. One of the hallmark observations during this tenure is the identification of quiet sun bi-directional jets in the lower transition region. We investigate these events through two distinct avenues of study: a MHD model for reconnection and the new datasets of the Interface Region Imaging Spectrograph (IRIS). Based on forward modeling optically thin spectral profiles, we find the spectral signatures of reconnection can vary dramatically based on viewing angle and altitude. We look to the IRIS data to provide a more complete context of the chromospheric and coronal environment during these dynamic events. During a joint IRIS-SUMER observing campaign, we observed spectra of multiple jets, a small C flare, and an Ellerman bomb event. We discuss the questions that arise from the inspection of these new data.

  10. Iris-Corinna Schwarz (PhD, MAPS) Department of Linguistics iris@ling.su.se Phonetics Lab/Babylab

    E-print Network

    Iris-Corinna Schwarz (PhD, MAPS) Department of Linguistics iris@ling.su.se Phonetics Lab/Babylab w current Assistant professor at the Phonetics lab/Babylab, Department of Linguistics, Stockholm University

  11. Iris recognition system by using support vector machines

    Microsoft Academic Search

    Hasimah Ali; Momoh J. E. Salami; Wahyudi

    2008-01-01

    In recent years, with the increasing demands of security in our networked society, biometric systems for user verification are becoming more popular. Iris recognition system is a new technology for user verification. In this paper, the CASIA iris database is used for individual userpsilas verification by using support vector machines (SVMs) which based on the analysis of iris code as

  12. Implementation of various approaches for iris image normalization

    Microsoft Academic Search

    Namrata P. Joshi; Roopal K. Lamba; Devang U. Shah; Bhargav V. Ghadia

    2011-01-01

    A biometric system provides automatic identification of an individual, based on a unique feature or characteristic possessed by the individual. Iris recognition is regarded as the most reliable and accurate biometric identification system available. Here, the segmented iris region is processed to allow comparisons with existing database (Normalization and Enhancement). The normalization process will produce iris regions, which have the

  13. Reactor physics of natriun-cooled fast breeder reactors

    Microsoft Academic Search

    M. Rajamaeki

    1980-01-01

    Reactor properties of liquid metal fast breeder reactors and pertinent computational methods with their applications and results are presented. Designs of reactor core and fuel are outlined. Main characteristics and layout of the homogenous and the heterogenous core are described. Reactor safety and the possibility of a hypothetical core descriptive accident is discussed. The influence of the neutron flux on

  14. Comparison of Two IRI plasmasphere Extensions with GPS-TEC Observations

    NASA Technical Reports Server (NTRS)

    Gulyaeva, T. L.; Gallagher, Dennis L.

    2006-01-01

    Comparisons of two model results with Global Positioning System GPS-TEC measurements have been carried out for different latitudinal, solar activity, magnetic activity, diurnal and seasonal conditions. The models evaluated are the Global Core Plasma Model (GCPM-2000) and the IRI extension with Russian plasmasphere model (IRI*).Data of 23 observatories providing GPS-TEC and ionosonde data have been used. It is shown that IRI* plasmasphere electron density is greater than GCPM results by an order of magnitude at 6370 km altitude (one Earth's radius) with this excess growing to 2-3 orders of magnitude towards the GPS satellite altitude of 20000 km. Another source of model and GPS-TEC differences is a way of selection of the F2 layer peak parameters driving the models either with ITU-R (former CCIR) maps or ionosonde observations. Plasmasphere amendment to IRI improves accuracy of TEC model predictions because the plasmasphere contribution to the total TEC varies from 10% by daytime under quiet magnetic conditions to more than 50% by night under stormy conditions.

  15. Iris unwrapping using the Bresenham circle algorithm for real-time iris recognition

    NASA Astrophysics Data System (ADS)

    Carothers, Matthew T.; Ngo, Hau T.; Rakvic, Ryan N.; Broussard, Randy P.

    2015-02-01

    An efficient parallel architecture design for the iris unwrapping process in a real-time iris recognition system using the Bresenham Circle Algorithm is presented in this paper. Based on the characteristics of the model parameters this algorithm was chosen over the widely used polar conversion technique as the iris unwrapping model. The architecture design is parallelized to increase the throughput of the system and is suitable for processing an inputted image size of 320 × 240 pixels in real-time using Field Programmable Gate Array (FPGA) technology. Quartus software is used to implement, verify, and analyze the design's performance using the VHSIC Hardware Description Language. The system's predicted processing time is faster than the modern iris unwrapping technique used today?.

  16. INDUSTRIAL RESEARCH AND DEVELOPMENT INFORMATION SYSTEM (IRIS)

    EPA Science Inventory

    The National Science Foundation's (NSF) Industrial Research and Development Information System (IRIS) links an online interface to a historical database with more than 2,500 statistical tables containing all industrial research and development (R&D) data published by NSF since 19...

  17. IRIS Resolving Unresolved Structure - Duration: 11 seconds.

    NASA Video Gallery

    NASA’s IRIS, which is able to look at a low layer of the sun’s atmosphere in unprecedented resolution, reveals details in the bright loops seen over the sun’s limb that have never been witnessed be...

  18. Iris Compression for Cryptographically Secure Person Identification

    Microsoft Academic Search

    Daniel Schonberg; Darko Kirovski

    2004-01-01

    Abstract In this paper we propose EyeCerts, a biometric system for identification of people which achieves o - line verification of certified, cryptographically secure documents An EyeCert is a printed d ocument which certifies the association of a given text with a biometric feature - a compressed version of a human iris in this work As a central component of

  19. The Road to IRIS data products

    NASA Astrophysics Data System (ADS)

    Hurlburt, N. E.; Title, A. M.; De Pontieu, B.; Lemen, J. R.; Wuelser, J.; Tarbell, T. D.; Wolfson, C. J.; Schrijver, C. J.; Golub, L.; DeLuca, E. E.; Kankelborg, C. C.; Hansteen, V. H.; Carlsson, M.; Bush, R. I.

    2013-12-01

    The Interface Region Imaging Spectrograph generates a complex set of data products that the IRIS team has strived to deliver to the community in forms that are easy to find and use. We review the results of these efforts and invite the community to explore the data and tools. All standard IRIS data products are based on calibrated images are corrected for a variety of instrumental effects. The resulting products are incorporated into the Heliophysics Event Knowledgebase (HEK) as annotated data sets accessible through the HEK Coverage Registry (HCR). Annotations include descriptions of the data products themselves (pointing, field of view, cadence...) as well as references to coordinated observations from the Hinode mission and other observatories, and to solar events identified in the HEK Event Registry (HER). IRIS data products are available at the LMSAL and Stanford (JSOC) data centers in Palo Alto and the Hinode Data Center in Oslo. Portals that can help users to select data products include the LMSAL iSolsearch, the Virtual Solar Observatory and Helioviewer. Supporting analysis software is available in the IRIS branch of SolarSoft.

  20. Pigment Melanin: Pattern for Iris Recognition

    Microsoft Academic Search

    S. Mahdi Hosseini; Babak Nadjar Araabi; Hamid Soltanian-Zadeh

    2010-01-01

    Recognition of iris based on visible light (VL) imaging is a difficult problem because of the light reflection from the cornea. Nonetheless, pigment melanin provides a rich feature source in VL, which is unavailable in near-infrared (NIR) imaging. This is due to the biological spectroscopy of eumelanin, a chemical not stimulated in NIR. In this case, a plausible solution to

  1. IRI/LDEO Introduction to Climate Data

    NSDL National Science Digital Library

    A collection of the datasets from the IRI/LDEO Climate Data Library, which contain important information about our planet Earth. Datasets include: Topography, ENSO (El Nino-Southern Oscillation) Monitor, Historical Temperature and Precipitation, and Ocean Climatology, including ocean temperature, salinity, and nutrients, including dissolved oxygen, nitrate, phosphate, and silicate. Figures are very easy to manipulate and the parameters are explained in detail.

  2. Iris mammillations in two female siblings with congenital adrenal hyperplasia

    PubMed Central

    Peyman, Mohammadreza; Ong, Ming Jew; Iqbal, Tajunisah; Subrayan, Visvaraja

    2010-01-01

    Iris mammillations are dark brown, smooth, mound- or dome-shaped protuberances that are typically found on the anterior iris surface and are presumed to be congenital in origin. This congenital anomaly is usually unilateral and can be hereditary or sporadic. Lisch nodules in neurofibromatosis, tapioca melanoma of the iris, inflammatory iris granulomata and Cogan–Reese syndrome should be considered in the differential diagnosis. In this case report, the authors present a case of a bilateral iris mammillations in two siblings with congenital adrenal hyperplasia (CAH). To our knowledge, this is the first case where bilateral iris mammilations have been found to be associated with a systemic condition. Iris mammillations can be considered as one of the clinical signs in CAH in view of the pathogenesis discussed. Detailed ocular examination in CAH may reveal an increased incidence. PMID:22802477

  3. The features of neutronic calculations for fast reactors with hybrid cores on the basis of BFS-62-3A critical assembly experiments

    SciTech Connect

    Mitenkova, E. F.; Novikov, N. V. [Nuclear Safety Inst. of Russian Academy of Sciences, B. Tulskaya 52, Moscow, 115119 (Russian Federation); Blokhin, A. I. [State Scientific Center of Russian Federation, Inst. of Physics and Power Engineering Named after A.I. Leypunsky, Bondarenko Square 1, Obninsk, Kaluga Region, 249030 (Russian Federation)

    2012-07-01

    The different (U-Pu) fuel compositions are considered for next generation of sodium fast breeder reactors. The considerable discrepancies in axial and radial neutron spectra for hybrid reactor systems compared to the cores with UO{sub 2} fuel cause increasing uncertainty of generating the group nuclear constants in those reactor systems. The calculation results of BFS-62-3A critical assembly which is considered as full-scale model of BN-600 hybrid core with steel reflector specify quite different spectra in local areas. For those systems the MCNP 5 calculations demonstrate significant sensitivity of effective multiplication factor K{sub eff} and spectral indices to nuclear data libraries. For {sup 235}U, {sup 238}U, {sup 239}Pu the results of calculated radial fission rate distributions against the reconstructed ones are analyzed. Comparative analysis of spectral indices, neutron spectra and radial fission rate distributions are performed using the different versions of ENDF/B, JENDL-3.3, JENDL-4, JEFF-3.1.1 libraries and BROND-3 for Fe, Cr isotopes. For analyzing the fission rate sensitivity to the plutonium presence in the fuel {sup 239}Pu is substituted for {sup 235}U (enrichment 90%) in the FA areas containing the plutonium. For {sup 235}U, {sup 238}U, {sup 239}Pu radial fission rate distributions the explanation of pick values discrepancies is based on the group fission constants analyses and possible underestimation of some features at the experimental data recovery method (Westcott factors, temperature dependence). (authors)

  4. Computer Models for IRIS Control System Transient Analysis

    SciTech Connect

    Gary D. Storrick; Bojan Petrovic; Luca Oriani

    2007-01-31

    This report presents results of the Westinghouse work performed under Task 3 of this Financial Assistance Award and it satisfies a Level 2 Milestone for the project. Task 3 of the collaborative effort between ORNL, Brazil and Westinghouse for the International Nuclear Energy Research Initiative entitled “Development of Advanced Instrumentation and Control for an Integrated Primary System Reactor” focuses on developing computer models for transient analysis. This report summarizes the work performed under Task 3 on developing control system models. The present state of the IRIS plant design – such as the lack of a detailed secondary system or I&C system designs – makes finalizing models impossible at this time. However, this did not prevent making considerable progress. Westinghouse has several working models in use to further the IRIS design. We expect to continue modifying the models to incorporate the latest design information until the final IRIS unit becomes operational. Section 1.2 outlines the scope of this report. Section 2 describes the approaches we are using for non-safety transient models. It describes the need for non-safety transient analysis and the model characteristics needed to support those analyses. Section 3 presents the RELAP5 model. This is the highest-fidelity model used for benchmark evaluations. However, it is prohibitively slow for routine evaluations and additional lower-fidelity models have been developed. Section 4 discusses the current Matlab/Simulink model. This is a low-fidelity, high-speed model used to quickly evaluate and compare competing control and protection concepts. Section 5 describes the Modelica models developed by POLIMI and Westinghouse. The object-oriented Modelica language provides convenient mechanisms for developing models at several levels of detail. We have used this to develop a high-fidelity model for detailed analyses and a faster-running simplified model to help speed the I&C development process. Section 6 describes an ACSL model that Westinghouse started but suspended developing for the moment. ACSL is an old simulation language that Westinghouse used on many projects. It may (or may not) offer some advantages during the later stages of detailed plant design and analysis, but supporting the ACSL model does not appear to be necessary at this time. Section 7 summarizes our expectations for future development.

  5. Morphology and neurochemistry of rabbit iris innervation.

    PubMed

    He, Jiucheng; Bazan, Haydee E P

    2015-06-01

    The aim of this study was to map the entire nerve architecture and sensory neuropeptide content of the rabbit iris. Irises from New Zealand rabbits were stained with antibodies against neuronal-class ?III-tubulin, calcitonin gene-related peptide (CGRP) and substance P (SP), and whole-mount images were acquired to build a two-dimensional view of the iridal nerve architecture. After taking images in time-lapse mode, we observed thick nerves running in the iris stroma close to the anterior epithelia, forming four to five stromal nerve rings from the iris periphery to the pupillary margin and sub-branches that connected with each other, constituting the stromal nerve plexus. In the anterior side, fine divisions derivated from the stromal nerves, forming a nerve network-like structure to innervate the superficial anterior border layer, with the pupillary margin having the densest innervation. In the posterior side, the nerve bundles ran along with the pupil dilator muscle in a radial pattern. The morphology of the iris nerves on both sides changed with pupil size. To obtain the relative content of the neuropeptides in the iris, the specimens were double stained with ?III-tubulin and CGRP or SP antibodies. Relative nerve fiber densities for each fiber population were assessed quantitatively by computer-assisted analysis. On the anterior side, CGRP-positive nerve fibers constituted about 61%, while SP-positive nerves constitute about 30.5%, of the total nerve content, which was expressed as ?III tubulin-positive fibers. In addition, in the anterior stroma of the collarette region, there were non-neuronal cells that were positive for SP. On the posterior side, CGRP-positive nerve fibers were about 69% of total nerve content, while SP constituted only up to 20%. Similarly, in the trigeminal ganglia (TG), the number of CGRP-positive neurons significantly outnumbered those that were positive for SP. Also, all the SP-positive neurons were labeled with CGRP. This is the first study to provide a two-dimensional whole mount and a cross-sectional view of the entire iris nerve architecture. Considering the anatomical location, the high expression of CGRP and SP suggests that these neuropeptides may play a role in the pathogenesis of anterior uveitis, glaucoma, cataracts and chronic ocular pain. PMID:25752697

  6. Procedure of recovery of pin-by-pin fields of energy release in the core of VVER-type reactor for the BIPR-8 code

    NASA Astrophysics Data System (ADS)

    Gordienko, P. V.; Kotsarev, A. V.; Lizorkin, M. P.

    2014-12-01

    The procedure of recovery of pin-by-pin energy-release fields for the BIPR-8 code and the algorithm of the BIPR-8 code which is used in nodal computation of the reactor core and on which the recovery of pin-by-pin fields of energy release is based are briefly described. The description and results of the verification using the module of recovery of pin-by-pin energy-release fields and the TVS-M program are given.

  7. Neutronics methods, models, and applications at the Idaho National Engineering Laboratory for the advanced neutron source reactor three-element core design

    SciTech Connect

    Wemple, C.A.; Schnitzler, B.G.; Ryskamp, J.M. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

    1995-08-01

    A summary of the methods and models used to perform neutronics analyses on the Advanced Neutron Source reactor three-element core design is presented. The applications of the neutral particle Monte Carlo code MCNP are detailed, as well as the expansion of the static role of MCNP to analysis of fuel cycle depletion calculations. Results to date of these applications are presented also. A summary of the calculations not yet performed is also given to provide a {open_quotes}to-do{close_quotes} list if the project is resurrected.

  8. Procedure of recovery of pin-by-pin fields of energy release in the core of VVER-type reactor for the BIPR-8 code

    SciTech Connect

    Gordienko, P. V., E-mail: gorpavel@vver.kiae.ru; Kotsarev, A. V.; Lizorkin, M. P. [National Research Center Kurchatov Institute (Russian Federation)

    2014-12-15

    The procedure of recovery of pin-by-pin energy-release fields for the BIPR-8 code and the algorithm of the BIPR-8 code which is used in nodal computation of the reactor core and on which the recovery of pin-by-pin fields of energy release is based are briefly described. The description and results of the verification using the module of recovery of pin-by-pin energy-release fields and the TVS-M program are given.

  9. Simulation of (16)O (n, p) (16)N reaction rate and nitrogen-16 inventory in a high performance light water reactor with one pass core.

    PubMed

    Kebwaro, Jeremiah Monari; Zhao, Yaolin; He, Chaohui

    2014-12-01

    The rate of activation of the isotope (16)O to (16)N in a typical HPLWR one pass concept was calculated using MCNP code. A mathematical model was used to track the inventory of the radioisotope (16)N in a unit mass of coolant traversing the system. The water leaving the moderator channels has the highest activity in the circuit, but due to interaction with fresh coolant at the lower plenum, the activity is downscaled. The calculated core exit activity is higher than values reported in literature for commercial boiling water reactors. PMID:25084129

  10. An Iris Recognition System to Enhance E-security Environment Based on Wavelet Theory

    Microsoft Academic Search

    Jafar M. H. Ali; Aboul Ella

    In this paper, efficient biometric security techniques for iris recognition system with high performance and high confidence are described. The system is based on an empirical analysis of the iris image and it is split in several steps using local image properties. The system steps are capturing iris patterns; determine the location of the iris boundaries; converting the iris boundary

  11. Determination of total serum insulin (IRI) in insulin-treated diabetic patients

    Microsoft Academic Search

    Lise G. Heding

    1972-01-01

    Summary  A routine method is described for the determination of total IRI (imraunoreactive insulin) in insulintreated diabetics. The method involves an easy acid ethanol extraction, whereby antibody-bound IRI is dissociated and separated, together with the free IRI from the serum proteins and the antibodies. The recovery of IRI in this procedure is about 80%. After the separation, the isolated total IRI

  12. Non-local equilibrium two-phase flow model with phase change in porous media and its application to reflooding of a severely damaged reactor core

    NASA Astrophysics Data System (ADS)

    Bachrata, A.; Fichot, F.; Quintard, M.; Repetto, G.; Fleurot, J.

    2012-05-01

    A generalized non local-equilibrium, three-equation model was developed for the macroscopic description of two-phase flow heat and mass transfer in porous media subjected to phase change. Six pore-scale closure problems were proposed to determine all the effective transport coefficients for representative unit cells. An improved model is presented in this paper with the perspective of application to intense boiling phenomena. The objective of this paper is to present application of this model to the simulation of reflooding of severely damaged nuclear reactor cores. In case of accident at a nuclear power plant, water sources may not be available for a long period of time and the core heats up due to the residual power. Any attempt to inject water during core degradation can lead to quenching and further fragmentation of the core material. The fragmentation of fuel rods and melting of reactor core materials may result in the formation of a "debris bed". The typical particle size in a debris bed might reach few millimeters (characteristic length-scale: 1 to 5 mm), which corresponds to a high permeability porous medium. The proposed two-phase flow model is implemented in the ICARECATHARE code, developed by IRSN to study severe accident scenarios in pressurized water reactors. Currently, the French IRSN has set up two experimental facilities to study debris bed reflooding, PEARL and PRELUDE, with the objective to validate safety models. The PRELUDE program studies the complex two phase flow of water and steam in a porous medium (diameter 180 mm, height 200 mm), initially heated to a high temperature (400°C or 700°C). The series of PRELUDE experiments achieved in 2010 constitute a significant complement to the database of high temperature bottom reflood experimental data. They provide relevant data to understand the progression of the quench front and the intensity of heat transfer. Modeling accurately these experiments required improvements to the reflooding model, especially in terms of the existence of various saturation regimes. The improved two-phase flow model shows a good agreement with PRELUDE experimental results.

  13. Evaluation of the thermal-hydraulic operating limits of the HEU-LEU transition cores for the MIT Research Reactor

    E-print Network

    Wang, Yunzhi (Yunzhi Diana)

    2009-01-01

    The MIT Research Reactor (MITR) is in the process of conducting a design study to convert from High Enrichment Uranium (HEU) fuel to Low Enrichment Uranium (LEU) fuel. The currently selected LEU fuel design contains 18 ...

  14. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    Except as provided in paragraph (a)(1)(ii) of this section, the evaluation model must include sufficient supporting justification to show that the analytical technique realistically describes the behavior of the reactor system during a loss-of-coolant...

  15. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    Except as provided in paragraph (a)(1)(ii) of this section, the evaluation model must include sufficient supporting justification to show that the analytical technique realistically describes the behavior of the reactor system during a loss-of-coolant...

  16. TITAN : an advanced three dimensional coupled neutronicthermal-hydraulics code for light water nuclear reactor core analysis

    E-print Network

    Griggs, D. P.

    1984-01-01

    The accurate analysis of nuclear reactor transients frequently requires that neutronics, thermal-hydraulics and feedback be included. A number of coupled neutronics/thermal-hydraulics codes have been developed for this ...

  17. Estimate of radiation release from MIT reactor with low enriched uranium (LEU) core during maximum hypothetical accident

    E-print Network

    Plumer, Kevin E. (Kevin Edward)

    2011-01-01

    In accordance with a 1986 NRC ruling, the MIT Research Reactor (MITR) is planning on converting from the use of highly enriched uranium (HEU) to low enriched uranium (LEU) for fuel. A component of the conversion analysis ...

  18. Nuclear safety analyses and core design calculations to convert the Texas A & M University Nuclear Science Center reactor to low enrichment uranium fuel. Final report

    SciTech Connect

    Parish, T.A.

    1995-03-02

    This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclear safety analysis capabilities established and maintained at the NSCR.

  19. Video-Based Noncooperative Iris Image Segmentation

    Microsoft Academic Search

    Yingzi Du; Emrah Arslanturk; Zhi Zhou; Craig Belcher

    2011-01-01

    In this paper, we propose a video-based noncooper- ative iris 1 image segmentation scheme that incorporates a qual- ity filter to quickly eliminate images without an eye, employs a coarse-to-fine segmentation scheme to improve the overall effi- ciency, uses a direct least squares fitting of ellipses method to model the deformed pupil and limbic boundaries, and develops a window gradient-based

  20. The Interface Region Imaging Spectrograph (IRIS)

    E-print Network

    De Pontieu, B; Lemen, J; Kushner, G D; Akin, D J; Allard, B; Berger, T; Boerner, P; Cheung, M; Chou, C; Drake, J F; Duncan, D W; Freeland, S; Heyman, G F; Hoffman, C; Hurlburt, N E; Lindgren, R W; Mathur, D; Rehse, R; Sabolish, D; Seguin, R; Schrijver, C J; Tarbell, T D; Wuelser, J -P; Wolfson, C J; Yanari, C; Mudge, J; Nguyen-Phuc, N; Timmons, R; van Bezooijen, R; Weingrod, I; Brookner, R; Butcher, G; Dougherty, B; Eder, J; Knagenhjelm, V; Larsen, S; Mansir, D; Phan, L; Boyle, P; Cheimets, P N; DeLuca, E E; Golub, L; Gates, R; Hertz, E; McKillop, S; Park, S; Perry, T; Podgorski, W A; Reeves, K; Saar, S; Testa, P; Tian, H; Weber, M; Dunn, C; Eccles, S; Jaeggli, S A; Kankelborg, C C; Mashburn, K; Pust, N; Springer, L; Carvalho, R; Kleint, L; Marmie, J; Mazmanian, E; Pereira, T M D; Sawyer, S; Strong, J; Worden, S P; Carlsson, M; Hansteen, V H; Leenaarts, J; Wiesmann, M; Aloise, J; Chu, K -C; Bush, R I; Scherrer, P H; Brekke, P; Martinez-Sykora, J; Lites, B W; McIntosh, S W; Uitenbroek, H; Okamoto, T J; Gummin, M A; Auker, G; Jerram, P; Pool, P; Waltham, N

    2014-01-01

    The Interface Region Imaging Spectrograph (IRIS) small explorer spacecraft provides simultaneous spectra and images of the photosphere, chromosphere, transition region, and corona with 0.33-0.4 arcsec spatial resolution, 2 s temporal resolution and 1 km/s velocity resolution over a field-of-view of up to 175 arcsec x 175 arcsec. IRIS was launched into a Sun-synchronous orbit on 27 June 2013 using a Pegasus-XL rocket and consists of a 19-cm UV telescope that feeds a slit-based dual-bandpass imaging spectrograph. IRIS obtains spectra in passbands from 1332-1358, 1389-1407 and 2783-2834 Angstrom including bright spectral lines formed in the chromosphere (Mg II h 2803 Angstrom and Mg II k 2796 Angstrom) and transition region (C II 1334/1335 Angstrom and Si IV 1394/1403 Angstrom). Slit-jaw images in four different passbands (C II 1330, Si IV 1400, Mg II k 2796 and Mg II wing 2830 Angstrom) can be taken simultaneously with spectral rasters that sample regions up to 130 arcsec x 175 arcsec at a variety of spatial sa...

  1. Infrasound product resources at the IRIS DMC

    NASA Astrophysics Data System (ADS)

    Bahavar, M.; Trabant, C.; Hutko, A. R.; Karstens, R.

    2012-12-01

    In 2011 infrasound sensors were installed at some existing USArray Transportable Array (TA) sites and became a standard component of all new sites. Currently there are over 400 sites with infrasound sensors with an average spacing of 70 kilometers. To promote and facilitate the use of these data the IRIS Data Management Center has developed two new data products: an infrasound reference event database and an infrasound signal detector. The TA Infrasound Reference Event Database (TAIRED) is a user-supported database that contains information on events of interest for which there are associated USArray microbarograph recordings. This database is initially populated with a few events from observations on the USArray infrasound data, event bulletins, news on explosions, meteorological events and rocket launches. As a user-supported resource, we ask users to submit events of interest to be included in the database or submit their alternate solutions to the existing events. The second data product is an infrasound signal detector that regularly scans the USArray broadband infrasound data (BDF channel sampled at 40 Hz) and produces detections that highlight time intervals containing potential signals of interest. The detection product includes two components, standard signal-to-noise ratio based detections and spectral power based detections. No attempt is made to categorize detections or associate them to events. These data products join the growing collection of products produced and managed at the IRIS DMC, for the complete list please visit http://www.iris.edu/dms/products/.

  2. IRIS: Industrial Research and Development Information System

    NSDL National Science Digital Library

    2001-01-01

    The National Science Foundation's Industrial Research and Development Information System (IRIS) houses a database of all of the statistics produced and published from the 1953-1998 cycles of the annual Survey of Industrial Research and Development (R&D). The statistics would be useful to workers in economics or anyone interested in learning about how funds are allocated among research areas. NSF states that the results of the survey are used by government agencies, corporations, and research organizations to determine productivity factors, formulate tax policy, and to investigate company performance. The statistics available in IRIS describe national estimates of the total expenditures on R&D performed within the United States by industrial firms, given in dollar amounts. Tabulations from the survey contain R&D statistics by industry, size of company, source of funds, character of R&D, R&D as a percentage of net sales, and R&D contracted to outside organizations and performed outside of the United States. They also contain estimates of the sales and total employment of R&D-performing companies, employment of R&D scientists and engineers, and statistics by state. Users have a variety of options for searching and browsing the Excel tables and Word documentation in IRIS -- by year, topic, or by measure -- and the resulting tables can display all years combined or just selected years.

  3. A novel portable iris recognition system and usability evaluation

    Microsoft Academic Search

    Youngkyoon Jang; Byung Jun Kang; Kang Ryoung Park

    2010-01-01

    We propose a new portable iris recognition system. Because existing portable iris systems use customized embedded processing\\u000a units, they are limited in ability to expand to other applications, and they have low processing power. To overcome such problems,\\u000a we propose a new portable iris recognition system consisting of a conventional ultra-mobile personal computer (UMPC), a small\\u000a universal serial bus (USB)

  4. Availability of the emergency core cooling system of a CANDU pressurized heavy-water reactor following a small loss-of-coolant accident

    SciTech Connect

    Al-kusayer, T.A.

    1985-06-01

    The availability of the emergency core cooling system (ECCS) as a long-term safety backup system following a small loss-of-coolant accident (LOCA) has been analyzed for the Pickering NGS Unit A, a Canada deuterium uranium (CANDU) type of pressurized heavy-water reactor (PHWR). Fault tree analysis methodology was used to assess the unavailability of the ECCS. The PREP and KITT computer codes were used to estimate the failure probabilities. From these computations, the unavailability of the ECCS to supply sufficient coolant to the core is estimated as 3.63 X 10/sup -3/. This figure is higher than the failure probability target 3 X 10/sup -3/ that is specified by the Canadian Atomic Energy Control Board for the safety systems of CANDU PHWRs. It has been found that human error might make a very important contribution to ECCS unavailability, especially if the human error rates have been assigned the upper bound values in the fault tree calculations. That should be the case, therefore, for any fault analysis and reliability assessment of nuclear generating stations. Unlike the case for light water reactors, the ECCS in a CANDU PHWR is not the last defense against the LOCA, because of the availability of quite a large amount of D/sub 2/O moderator in the calandria around the pressure tubes.

  5. Updating of ASME Nuclear Code Case N-201 to Accommodate the Needs of Metallic Core Support Structures for High Temperature Gas Cooled Reactors Currently in Development

    SciTech Connect

    Mit Basol; John F. Kielb; John F. MuHooly; Kobus Smit

    2007-05-02

    On September 29, 2005, ASME Standards Technology, LLC (ASME ST-LLC) executed a multi-year, cooperative agreement with the United States DOE for the Generation IV Reactor Materials project. The project's objective is to update and expand appropriate materials, construction, and design codes for application in future Generation IV nuclear reactor systems that operate at elevated temperatures. Task 4 was embarked upon in recognition of the large quantity of ongoing reactor designs utilizing high temperature technology. Since Code Case N-201 had not seen a significant revision (except for a minor revision in September, 2006 to change the SA-336 forging reference for 304SS and 316SS to SA-965 in Tables 1.2(a) and 1.2(b), and some minor editorial changes) since December 1994, identifying recommended updates to support the current high temperature Core Support Structure (CSS) designs and potential new designs was important. As anticipated, the Task 4 effort identified a number of Code Case N-201 issues. Items requiring further consideration range from addressing apparent inconsistencies in definitions and certain material properties between CC-N-201 and Subsection NH, to inclusion of additional materials to provide the designer more flexibility of design. Task 4 developed a design parameter survey that requested input from the CSS designers of ongoing high temperature gas cooled reactor metallic core support designs. The responses to the survey provided Task 4 valuable input to identify the design operating parameters and future needs of the CSS designers. Types of materials, metal temperature, time of exposure, design pressure, design life, and fluence levels were included in the Task 4 survey responses. The results of the survey are included in this report. This research proves that additional work must be done to update Code Case N-201. Task 4 activities provide the framework for the Code Case N-201 update and future work to provide input on materials. Candidate materials such as Type 321 and Type 347 austenitic stainless steels, Modified 9Cr-1Mo steel for core support structure construction, and Alloy 718 for Threaded Structural Fasteners were among the recommended materials for inclusion in the Code Case. This Task 4 Report identifies the need to address design life beyond 3 x 105 hours, especially in consideration of 60-year design life. A proposed update to the latest Code Case N-201 revision (i.e., Code Case N-201-5) including the items resolved in this report is included as Appendix A.

  6. Iris recognition as a biometric method after cataract surgery

    PubMed Central

    Roizenblatt, Roberto; Schor, Paulo; Dante, Fabio; Roizenblatt, Jaime; Belfort, Rubens

    2004-01-01

    Background Biometric methods are security technologies, which use human characteristics for personal identification. Iris recognition systems use iris textures as unique identifiers. This paper presents an analysis of the verification of iris identities after intra-ocular procedures, when individuals were enrolled before the surgery. Methods Fifty-five eyes from fifty-five patients had their irises enrolled before a cataract surgery was performed. They had their irises verified three times before and three times after the procedure, and the Hamming (mathematical) distance of each identification trial was determined, in a controlled ideal biometric environment. The mathematical difference between the iris code before and after the surgery was also compared to a subjective evaluation of the iris anatomy alteration by an experienced surgeon. Results A correlation between visible subjective iris texture alteration and mathematical difference was verified. We found only six cases in which the eye was no more recognizable, but these eyes were later reenrolled. The main anatomical changes that were found in the new impostor eyes are described. Conclusions Cataract surgeries change iris textures in such a way that iris recognition systems, which perform mathematical comparisons of textural biometric features, are able to detect these changes and sometimes even discard a pre-enrolled iris considering it an impostor. In our study, re-enrollment proved to be a feasible procedure. PMID:14748929

  7. IRIS Launch, Deploy and Beauty Pass Animation - Duration: 108 seconds.

    NASA Video Gallery

    Understanding the interface between the photosphere and corona remains a fundamental challenge in solar and heliospheric science. The Interface Region Imaging Spectrograph (IRIS) mission opens a wi...

  8. Iris recognition in the presence of ocular disease

    PubMed Central

    Aslam, Tariq Mehmood; Tan, Shi Zhuan; Dhillon, Baljean

    2009-01-01

    Iris recognition systems are among the most accurate of all biometric technologies with immense potential for use in worldwide security applications. This study examined the effect of eye pathology on iris recognition and in particular whether eye disease could cause iris recognition systems to fail. The experiment involved a prospective cohort of 54 patients with anterior segment eye disease who were seen at the acute referral unit of the Princess Alexandra Eye Pavilion in Edinburgh. Iris camera images were obtained from patients before treatment was commenced and again at follow-up appointments after treatment had been given. The principal outcome measure was that of mathematical difference in the iris recognition templates obtained from patients' eyes before and after treatment of the eye disease. Results showed that the performance of iris recognition was remarkably resilient to most ophthalmic disease states, including corneal oedema, iridotomies (laser puncture of iris) and conjunctivitis. Problems were, however, encountered in some patients with acute inflammation of the iris (iritis/anterior uveitis). The effects of a subject developing anterior uveitis may cause current recognition systems to fail. Those developing and deploying iris recognition should be aware of the potential problems that this could cause to this key biometric technology. PMID:19324690

  9. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2008

    SciTech Connect

    Primm, Trent [ORNL; Chandler, David [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Sease, John D [ORNL; Jolly, Brian C [ORNL

    2009-03-01

    This report documents progress made during FY 2008 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Scoping experiments with various manufacturing methods for forming the LEU alloy profile are presented.

  10. A three-dimensional neutronics-thermohydraulics simulation of core disruptive accident in sodium-cooled fast reactor

    Microsoft Academic Search

    Hidemasa Yamano; Yoshiharu Tobita; Satoshi Fujita

    2009-01-01

    The SIMMER-III code is a two-dimensional, multi-velocity-field, multi-phase, multi-component, Eulerian, fluid-dynamics code coupled with a fuel-pin model and a space- and energy-dependent neutron transport kinetics model. Since the three-dimensional representation of the core enables realistic distribution of the materials constituting the core, including control rods, SIMMER-IV has been developed as a direct extension of SIMMER-III to three dimensions with retaining

  11. Calculation and comparison of xenon and samarium reactivities of the HEU, LEU core in the low power research reactor.

    PubMed

    Dawahra, S; Khattab, K; Saba, G

    2015-07-01

    Comparative studies for the conversion of the fuel from HEU to LEU in the Miniature Neutron Source Reactor (MNSR) have been performed using the MCNP4C and GETERA codes. The precise calculations of (135)Xe and (149)Sm concentrations and reactivities were carried out and compared during the MNSR operation time and after shutdown for the existing HEU fuel (UAl4-Al, 90% enriched) and the potential LEU fuels (U3Si2-Al, U3Si-Al, U9Mo-Al, 19.75% enriched and UO2, 12.6% enriched) in this paper using the MCNP4C and GETERA codes. It was found that the (135)Xe and (149)Sm reactivities did not reach their equilibrium reactivities during the daily operating time of the reactor. The (149)Sm reactivities could be neglected compared to (135)Xe reactivities during the reactor operating time and after shutdown. The calculations for the UAl4-Al produced the highest (135)Xe reactivity in all the studied fuel group during the reactor operation (0.39mk) and after the reactor shutdown (0.735mk), It followed by U3Si-Al (0.34mk, 0.653mk), U3Si2-Al (0.33mk, 0.634mk), U9Mo-Al (0.3mk, 0.568mk) and UO2 (0.24mk, 0.448mk) fuels, respectively. Finally, the results showed that the UO2 was the best candidate for fuel conversion to LEU in the MNSR since it gave the lowest (135)Xe reactivity during the reactor operation and after shutdown. PMID:25816783

  12. Formation of a nuclear reactor's molten core bath in a crucible-type corium catcher for a nuclear power station equipped with VVER reactors

    NASA Astrophysics Data System (ADS)

    Beshta, S. V.; Vitol', S. A.; Granovskii, V. S.; Kalyago, E. K.; Kovtunova, S. V.; Krushinov, E. V.; Sulatskaya, M. B.; Sulatskii, A. A.; Khabenskii, V. B.; Al'Myashev, V. I.; Gusarov, V. V.

    2011-05-01

    Results from a calculation study on analyzing the formation of a melt bath in a crucible-type catcher for the conditions of a severe accident at a nuclear power station equipped with VVER-1000 reactors are presented. It is shown that the heat loads exerted on the water-cooled walls of the corium catcher shell are limited to a permissible level at which the necessary margins to nucleate boiling crisis and to destruction are ensured under the conditions of thermal and mechanical loading of the shell. An important role of sacrificial material in the efficient operation of the corium catcher is pointed out.

  13. Neutronics and Thermal Hydraulics Study for Using a Low-Enriched Uranium Core in the Advanced Test Reactor -- 2008 Final Report

    SciTech Connect

    G. S. Chang; M. A. Lillo; R. G. Ambrosek

    2008-06-01

    The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuel cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis was performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff versus effective full power days (EFPDs) between the HEU and the LEU cores. The MCNP ATR 1/8th core model was used to optimize the U 235 loading in the LEU core, such that the differences in K-eff and heat flux profiles between the HEU and LEU cores were minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the ATR reference HEU case study. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm (20 mil). In this work, the proposed LEU (U-10Mo) core conversion case with nominal fuel meat thickness of 0.330 mm (13 mil) and U-235 enrichment of 19.7 wt% is used to optimize the radial heat flux profile by varying the fuel meat thickness from 0.191 mm (7.0 mil) to 0.330 mm (13.0 mil) at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). A 0.8g of Boron-10, a burnable absorber, was added in the inner and outer plates to reduce the initial excess reactivity, and the peak to average ratio of the inner/outer heat flux more effectively. Because the B-10 (n,a) reaction will produce Helium-4 (He-4), which might degrade the LEU foil type fuel performance, an alternative absorber option is proposed. The proposed LEU case study will have 6.918 g of Cadmium (Cd) mixed with the LEU at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19) as a burnable absorber to achieve peak to average ratios similar to those for the ATR reference HEU case study.

  14. Reactor safety method

    DOEpatents

    Vachon, Lawrence J. (Clairton, PA)

    1980-03-11

    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.

  15. MATADOR: a computer code for the analysis of radionuclide behavior during degraded core accidents in light water reactors

    Microsoft Academic Search

    P. Baybutt; S. Raghuram; H. I. Avci

    1985-01-01

    A new computer code called MATADOR (Methods for the Analysis of Transport And Deposition Of Radionuclides) has been developed to replace the CORRAL computer code which was written for the Reactor Safety Study (WASH-1400). This report contains a detailed description of the models used in MATADOR. MATADOR is intended for use in system risk studies to analyze radionuclide transport and

  16. IRI Task Force Activity at ICTP: Proposed improvements for the IRI region below the F peak

    Microsoft Academic Search

    S. M. Radicella; D. Bilitza; B. W. Reinisch; J. O. Adeniyi; M. E. Mosert Gonzalez; B. Zolesi; M. L. Zhang; S. R. Zhang

    1998-01-01

    The Aeronomy and Radiopropagation Laboratory of the International Center for Theoretical Physics (ICTP) in Trieste, Italy has hosted special IRI Task Force Activities (TFAs) annually since 1994. This article reviews the format and results of the 1994, 1995, and 1996 TFAs. The prime focus of these TFAs has been the F1 region and the bottomside F2 region. Each meeting has

  17. Real-Time Iris Localization for Iris Recognition in Cellular Phone

    Microsoft Academic Search

    Dal Ho Cho; Kang Ryoung Park; Dae Woong Rhee

    2005-01-01

    With the increasing need of guaranteeing the security in case of using bank transaction service by using cellular phone, it is required to apply biometrics for the security of cellular phone. Especially, iris recognition is good for cellular phone security because of its reliability and accuracy compared to other biometrics such as face, fingerprint and voice recognition. In this paper,

  18. An Ocularist's Approach to Human Iris Synthesis Aaron Lefohn

    E-print Network

    Utah, University of

    An Ocularist's Approach to Human Iris Synthesis Aaron Lefohn University of Utah Richard Caruso Eye difficult. One such challenge is presented by the human eye, which exhibits intricate detail in the iris industries. In Section 2 we introduce the terminology for describing the anatomy of human eyes and irises

  19. IRIS TOXICOLOGICAL REVIEW AND SUMMARY DOCUMENTS FOR BORON

    EPA Science Inventory

    Boron was last updated on IRIS in the late 1980's . Since that time new studies have provided developmental data in three species. The goal of this updated IRIS file is to re-assess the RfD based on new research and updated methodology, specifically application of the benchmark a...

  20. IRIS Toxicological Review of Trimethylbenzenes (Revised External Review Draft)

    EPA Science Inventory

    In August 2013, EPA submitted a revised draft IRIS assessment of trimethylbenzenes to the agency's Science Advisory Board (SAB) and posted this draft on the IRIS website. EPA had previously released a draft of the assessment for public comment, held a public meeting about the dr...

  1. IRIS Toxicological Review of Ammonia (2013, Revised External Review Draft)

    EPA Science Inventory

    In August 2013, EPA submitted a revised draft IRIS assessment of ammonia to the agency's Science Advisory Board (SAB) and posted this draft on the IRIS website. EPA had previously released a draft of the assessment for public comment, held a public meeting about the draft, and ...

  2. The IRIS network site at the Wilcox Solar Observatory

    NASA Technical Reports Server (NTRS)

    Hoeksema, J. T.; Scherrer, P. H.

    1991-01-01

    The site for the International Research on the Interior of the Sun (IRIS) instrument housed at the Wilcox Solar Observatory at Stanford University (near San Francisco, USA) is described together with the instrument operation procedure. The IRIS instrument, which measures global oscillations of the sun, operates continuously every clear day since it was installed in August 1987.

  3. IRIS TOXICOLOGICAL REVIEW OF METHYL ETHYL KETONE (2003 Final)

    EPA Science Inventory

    EPA is announcing the release of the final report, "Toxicological Review of Methyl Ethyl Ketone: in support of the Integrated Risk Information System (IRIS)". The updated Summary for Methyl Ethyl Ketone and accompanying Quickview have also been added to the IRIS Database. ...

  4. Engineering performance of IRIS2 infrared imaging camera and spectrograph

    Microsoft Academic Search

    Vladimir Churilov; John Dawson; Greg A. Smith; Lew Waller; John D. Whittard; Roger Haynes; Allan Lankshear; Stuart D. Ryder; Chris G. Tinney

    2004-01-01

    IRIS2, the infrared imager and spectrograph for the Cassegrain focus of the Anglo Australian Telescope, has been in service since October 2001. IRIS2 incorporated many novel features, including multiple cryogenic multislit masks, a dual chambered vacuum vessel (the smaller chamber used to reduce thermal cycle time required to change sets of multislit masks), encoded cryogenic wheel drives with controlled backlash,

  5. A real-time focusing algorithm for iris recognition camera

    Microsoft Academic Search

    Kang Ryoung Park; Jaihie Kim

    2005-01-01

    For fast iris recognition, it is very important to capture the user's focused eye image at fast speed. Previous researchers have used the focusing method which has been applied to general landscape scenes without considering the characteristics of the iris image. So, they take much focusing time, especially in the case of the user with glasses. To overcome such problems,

  6. Eclipse-Free-Time Assessment Tool for IRIS

    NASA Technical Reports Server (NTRS)

    Eagle, David

    2012-01-01

    IRIS_EFT is a scientific simulation that can be used to perform an Eclipse-Free- Time (EFT) assessment of IRIS (Infrared Imaging Surveyor) mission orbits. EFT is defined to be those time intervals longer than one day during which the IRIS spacecraft is not in the Earth s shadow. Program IRIS_EFT implements a special perturbation of orbital motion to numerically integrate Cowell's form of the system of differential equations. Shadow conditions are predicted by embedding this integrator within Brent s method for finding the root of a nonlinear equation. The IRIS_EFT software models the effects of the following types of orbit perturbations on the long-term evolution and shadow characteristics of IRIS mission orbits. (1) Non-spherical Earth gravity, (2) Atmospheric drag, (3) Point-mass gravity of the Sun, and (4) Point-mass gravity of the Moon. The objective of this effort was to create an in-house computer program that would perform eclipse-free-time analysis. of candidate IRIS spacecraft mission orbits in an accurate and timely fashion. The software is a suite of Fortran subroutines and data files organized as a "computational" engine that is used to accurately predict the long-term orbit evolution of IRIS mission orbits while searching for Earth shadow conditions.

  7. Comparison of IRI2001 With TOPEX TEC Measurements

    Microsoft Academic Search

    G. Jee; R. W. Schunk; L. Scherliess

    2003-01-01

    The International Reference Ionosphere (IRI) is an international joint project of the Committee on Space Research (COSPAR) and the International Union of Radio Science (URSI). As one of the most comprehensive empirical models of the ionosphere, the IRI provides the electron density, electron temperature, ion temperature, and ion composition in the altitude range from about 50 km to 2000 km,

  8. Aggregates of the pentacenequinone derivative as reactors for the preparation of Ag@Cu2O core-shell NPs: an active photocatalyst for Suzuki and Suzuki type coupling reactions.

    PubMed

    Sharma, Kamaldeep; Kumar, Manoj; Bhalla, Vandana

    2015-07-23

    Aggregates of the pentacenequinone derivative 1 act as reactors and stabilizers for rapid and facile preparation of Ag@Cu2O core-shell nanoparticles (NPs) in aqueous medium. The in situ generated Ag@Cu2O core-shell hybrid materials enabled efficient visible light harvesting to catalyse the palladium free Suzuki-Miyaura and Suzuki type cross coupling reactions at room temperature. PMID:26151737

  9. DESIGN STUDY FOR A LOW-ENRICHED URANIUM CORE FOR THE HIGH FLUX ISOTOPE REACTOR, ANNUAL REPORT FOR FY 2010

    SciTech Connect

    Cook, David Howard [ORNL; Freels, James D [ORNL; Ilas, Germina [ORNL; Jolly, Brian C [ORNL; Miller, James Henry [ORNL; Primm, Trent [ORNL; Renfro, David G [ORNL; Sease, John D [ORNL; Pinkston, Daniel [ORNL

    2011-02-01

    This report documents progress made during FY 2010 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current level. Studies are reported of support to a thermal hydraulic test loop design, the implementation of finite element, thermal hydraulic analysis capability, and infrastructure tasks at HFIR to upgrade the facility for operation at 100 MW. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. Continuing development in the definition of the fuel fabrication process is described.

  10. A MELCOR Application to Two Light Water Reactor Nuclear Power Plant Core Melt Scenarios with Assumed Cavity Flooding Action

    Microsoft Academic Search

    Francisco Martin-Fuertes; Juan Manuel Martin-Valdepenas; Jose Mira; Maria Jesus Sanchez

    2003-01-01

    The MELCOR 1.8.4 code Bottom Head package has been applied to simulate two reactor cavity flooding scenarios for when the corium material relocates to the lower-plenum region in postulated severe accidents. The applications were preceded by a review of two main physical models, which highly impacted the results. A model comparison to available bibliography models was done, which allowed some

  11. Possible design improvements and a high power density fuel design for integral type small modular pressurized water reactors

    Microsoft Academic Search

    Ayd?n Karahan

    2010-01-01

    The study evaluates potential weaknesses and possible improvements for integral type small modular pressurized water reactor designs. By taking International Reactor Innovative and Secure (IRIS) as the reference design and keeping the power output as the same, a new fuel and reactor design were proposed. The proposed design relocates the primary coolant pumps and the pressurizer outside the reactor pressure

  12. Intraocular lens implantation for patients with coloboma of the iris

    PubMed Central

    LI, JUANJUAN; LI, YAN; HU, ZHULIN; KONG, LEI

    2014-01-01

    The aim of this study was to analyze the techniques for intraocular lens (IOL) implantation in patients with coloboma of the iris. A retrospective cohort study was used to analyze the degree of iris coloboma and the characteristics of the crystalline lens in 56 patients with iris coloboma. The patients with a lesser degree of coloboma of the iris and an intact lens capsule were treated by iris suture and IOL implantation into the posterior chamber. Patients with an iris coloboma confined to one quadrant, severe iris atrophy and significant lens capsule coloboma were treated with an annular suture at the edge of the pupil and IOL implantation into the anterior chamber. Patients with a greater degree of iris coloboma and an intact lens capsule were treated with an artificial iris and IOL implantation. The patients were followed up for between five months and five years after surgery. Data relating to vision, photophobia, IOL location, postoperative complications and treatment were also obtained at follow-up. The vision of the patients was improved to varying degrees following the surgery, with the exception of those with amblyopia or serious corneal scars. The photophobia of the patients had also improved. The patients’ levels of satisfaction and comfort were deemed to be satisfactory. Early postoperative complications included hyphema, increased intraocular pressure and uveitis. However, serious complications such as corneal decompensation and IOL dislocation were not observed. Various techniques for IOL implantation were selected based on the degree of iris and lens capsule coloboma; these techniques were capable of improving the vision and photophobia of the patients. PMID:24926350

  13. GLOBAL MANTLE STRUCTURE 2006 IRIS 5-YEAR PROPOSAL Using IRIS Digital Data to Determine the Radial Attenuation Structure of the

    E-print Network

    Wysession, Michael E.

    GLOBAL MANTLE STRUCTURE 2006 IRIS 5-YEAR PROPOSAL 152 Using IRIS Digital Data to Determine the Radial Attenuation Structure of the Lower Mantle Jesse Fisher Lawrence · Washington University model of shear quality factor (Qµ) with high sensitivity to the lower mantle. The data were obtained

  14. This paper explores the possibility of using multispectral iris information to enhance the recognition performance of an iris biometric system.

    E-print Network

    Ross, Arun Abraham

    that, based on the color of the eye, different components of the iris are highlighted at multiple. The components of the iris that are revealed in multiple spectral channels/wavelengths based on the color components that may exhibit different reflectance characteristics thereby justifying a multispectral analysis

  15. MEMS DM development at Iris AO, Inc.

    NASA Astrophysics Data System (ADS)

    Helmbrecht, Michael A.; He, Min; Kempf, Carl J.; Besse, Marc

    2011-03-01

    Iris AO is actively developing piston-tip-tilt (PTT) segmented MEMS deformable mirrors (DM) and adaptive optics (AO) controllers for these DMs. This paper discusses ongoing research at Iris AO that has advanced the state-of-the-art of these devices and systems over the past year. Improvements made to open-loop operation and mirror fabrication enables mirrors to open-loop flatten to 4 nm rms. Additional testing of an anti snap-in technology was conducted and demonstrates that the technology can withstand 100 million snap-in events without failure. Deformable mirrors with dielectric coatings are shown that are capable of handling 630 W/cm2 of incident laser power. Over a localized region on the segment, the dielectric coatings can withstand 100kW/cm2 incident laser power for 30 minutes. Results from the first-ever batch of PTT489 DMs that were shipped to pilot customers are reported. Optimizations made to the open-loop PTT controller are shown to have latencies of 157.5 ?s and synchronous array update rates of nearly 6.5 kHz. Finally, plans for the design and fabrication of the next-generation PTT939 DM are presented.

  16. Pigment Melanin: Pattern for Iris Recognition

    E-print Network

    Hosseini, Mahdi S; Soltanian-Zadeh, Hamid

    2009-01-01

    Recognition of iris based on Visible Light (VL) imaging is a difficult problem because of the light reflection from the cornea. Nonetheless, pigment melanin provides a rich feature source in VL, unavailable in Near-Infrared (NIR) imaging. This is due to biological spectroscopy of eumelanin, a chemical not stimulated in NIR. In this case, a plausible solution to observe such patterns may be provided by an adaptive procedure using a variational technique on the image histogram. To describe the patterns, a shape analysis method is used to derive feature-code for each subject. An important question is how much the melanin patterns, extracted from VL, are independent of iris texture in NIR. With this question in mind, the present investigation proposes fusion of features extracted from NIR and VL to boost the recognition performance. We have collected our own database (UTIRIS) consisting of both NIR and VL images of 158 eyes of 79 individuals. This investigation demonstrates that the proposed algorithm is highly s...

  17. Design of single-winding energy-storage reactors for dc-to-dc converters using air-gapped magnetic-core structures

    NASA Technical Reports Server (NTRS)

    Ohri, A. K.; Wilson, T. G.; Owen, H. A., Jr.

    1977-01-01

    A procedure is presented for designing air-gapped energy-storage reactors for nine different dc-to-dc converters resulting from combinations of three single-winding power stages for voltage stepup, current stepup and voltage stepup/current stepup and three controllers with control laws that impose constant-frequency, constant transistor on-time and constant transistor off-time operation. The analysis, based on the energy-transfer requirement of the reactor, leads to a simple relationship for the required minimum volume of the air gap. Determination of this minimum air gap volume then permits the selection of either an air gap or a cross-sectional core area. Having picked one parameter, the minimum value of the other immediately leads to selection of the physical magnetic structure. Other analytically derived equations are used to obtain values for the required turns, the inductance, and the maximum rms winding current. The design procedure is applicable to a wide range of magnetic material characteristics and physical configurations for the air-gapped magnetic structure.

  18. A “Generation III+” nuclear reactor for space needs

    Microsoft Academic Search

    A. Cammi; E. Finzi; C. Lombardi; M. E. Ricotti; L. Santini

    2009-01-01

    Nuclear power has been extensively used in space applications but in very few cases it entailed the use of nuclear fission reactors. The paper deals with a feasibility study of a space nuclear reactor for planet settlements, based on a “Generation III+” technology, specifically the integral type PWR one, currently under study in several projects (e.g. IRIS, CAREM, IMR, SMART)

  19. EVALUATION OF CORE PHYSICS ANALYSIS METHODS FOR CONVERSION OF THE INL ADVANCED TEST REACTOR TO LOW-ENRICHMENT FUEL

    SciTech Connect

    Mark DeHart; Gray S. Chang

    2012-04-01

    Computational neutronics studies to support the possible conversion of the ATR to LEU are underway. Simultaneously, INL is engaged in a physics methods upgrade project to put into place modern computational neutronics tools for future support of ATR fuel cycle and experiment analysis. A number of experimental measurements have been performed in the ATRC in support of the methods upgrade project, and are being used to validate the new core physics methods. The current computational neutronics work is focused on performance of scoping calculations for the ATR core loaded with a candidate LEU fuel design. This will serve as independent confirmation of analyses that have been performed previously, and will evaluate some of the new computational methods for analysis of a candidate LEU fuel for ATR.

  20. Evaluation of core physics analysis methods for conversion of the INL advanced test reactor to low-enrichment fuel

    SciTech Connect

    DeHart, M. D.; Chang, G. S. [Idaho National Laboratory, 2525 Fremont Street, Idaho Falls, ID 83415-3870 (United States)

    2012-07-01

    Computational neutronics studies to support the possible conversion of the ATR to LEU are underway. Simultaneously, INL is engaged in a physics methods upgrade project to put into place modern computational neutronics tools for future support of ATR fuel cycle and experiment analysis. A number of experimental measurements have been performed in the ATRC in support of the methods upgrade project, and are being used to validate the new core physics methods. The current computational neutronics work is focused on performance of scoping calculations for the ATR core loaded with a candidate LEU fuel design. This will serve as independent confirmation of analyses that have been performed previously, and will evaluate some of the new computational methods for analysis of a candidate LEU fuel for ATR. (authors)

  1. Design Study for a Low-enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2007

    SciTech Connect

    Primm, Trent [ORNL; Ellis, Ronald James [ORNL; Gehin, Jess C [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Sease, John D [ORNL

    2007-11-01

    This report documents progress made during fiscal year 2007 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low enriched uranium fuel (LEU). Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. A high volume fraction U/Mo-in-Al fuel could attain the same neutron flux performance as with the current, HEU fuel but materials considerations appear to preclude production and irradiation of such a fuel. A diffusion barrier would be required if Al is to be retained as the interstitial medium and the additional volume required for this barrier would degrade performance. Attaining the high volume fraction (55 wt. %) of U/Mo assumed in the computational study while maintaining the current fuel plate acceptance level at the fuel manufacturer is unlikely, i.e. no increase in the percentage of plates rejected for non-compliance with the fuel specification. Substitution of a zirconium alloy for Al would significantly increase the weight of the fuel element, the cost of the fuel element, and introduce an as-yet untried manufacturing process. A monolithic U-10Mo foil is the choice of LEU fuel for HFIR. Preliminary calculations indicate that with a modest increase in reactor power, the flux performance of the reactor can be maintained at the current level. A linearly-graded, radial fuel thickness profile is preferred to the arched profile currently used in HEU fuel because the LEU fuel media is a metal alloy foil rather than a powder. Developments in analysis capability and nuclear data processing techniques are underway with the goal of verifying the preliminary calculations of LEU flux performance. A conceptual study of the operational cost of an LEU fuel fabrication facility yielded the conclusion that the annual fuel cost to the HFIR would increase significantly from the current, HEU fuel cycle. Though manufacturing can be accomplished with existing technology, several engineering proof-of-principle tests would be required. The RERTR program is currently conducting a series of generic fuel qualification tests at the Advanced Test Reactor. A review of these tests and a review of the safety basis for the current, HEU fuel cycle led to the identification of a set of HFIR-specific fuel qualification tests. Much additional study is required to formulate a HFIR-specific fuel qualification plan from this set. However, one such test - creating a graded fuel profile across a flat foil - has been initiated with promising results.

  2. Depth-charge static and time-dependence perturbation/sensitivity system for nuclear reactor core analysis. [LMFBR

    SciTech Connect

    White, J.R.

    1981-09-01

    This report provides the background theory, user input, and sample problems required for the efficient application of the DEPTH-CHARGE system - a code block for both static and time-dependence perturbation theory and data sensitivity analyses. The DEPTH-CHARGE system is of modular construction and has been implemented within the VENTURE-BURNER computational system at Oak Ridge National Labortary. The DEPTH-CHARGE system provides, for the first time, a complete generalized first-order perturbation/sensitivity theory capability for both static and time-dependent analysis of realistic multidimensional reactor models.

  3. Tear-duct detector for identifying left versus right iris images

    Microsoft Academic Search

    Ramzi Abiantun; Marios Savvides

    2008-01-01

    In this paper, we present different pattern recognition approaches for automatically detecting tear ducts in iris acquired eye images for enhancing iris recognition and detecting mislabeling in datasets. Detecting the tear duct in an image will tell an iris recognition system whether the presented eye image is that of a left or a right eye. This will enable the iris

  4. Real-time iris detection on rotated faces

    NASA Astrophysics Data System (ADS)

    Perez, Claudio A.; Lazcano, Vanel A.; Estevez, Pablo A.; Held, Claudio M.

    2003-10-01

    Real-time face and iris detection on video sequences has been used to study the eye function and in diverse applications such as drowsiness detection, virtual keyboard interfaces, face recognition and multimedia retrieval. A non-invasive real time iris detection method was developed and consists of three stages: coarse face detection, fine face detection and iris detection. Anthropometric templates are used in these three stages. Elliptical templates are used to locate the coarse face center. A set of anthropometric templates which are probabilistic maps for the eyebrows, nose and mouth are used to perform the fine face detection. Face rotations are considered by rotating the anthropometric templates in fixed angles in steps of 10 degrees. Iris position is then determined within the eye region using another template with concentric semi-circles to compute a line integral in the boundary iris-sclera. The position with the maximum value indicates the iris center. The new method was applied on 10 video sequences, with a total of 6470 frames, from different people rotating their faces in the coronal axis. Results of correct face detection on 8 video sequences was 100%, one reached 99.9% and one 98.2%. Results on correct iris detection are above 96% in 9 of the video sequences and one reached 77.8%. The method was implemented in real-time (30 frames per second) with a PC 1.8 GHz.

  5. Lead-Cooled Fast Reactor (LFR) Design: Safety, Neutronics, Thermal Hydraulics, Structural Mechanics, Fuel, Core, and Plant Design

    SciTech Connect

    Smith, C

    2010-02-22

    The idea of developing fast spectrum reactors with molten lead (or lead alloy) as a coolant is not a new one. Although initially considered in the West in the 1950s, such technology was not pursued to completion because of anticipated difficulties associated with the corrosive nature of these coolant materials. However, in the Soviet Union, such technology was actively pursued during the same time frame (1950s through the 1980s) for the specialized role of submarine propulsion. More recently, there has been a renewal of interest in the West for such technology, both for critical systems as well as for Accelerator Driven Subcritical (ADS) systems. Meanwhile, interest in the former Soviet Union, primarily Russia, has remained strong and has expanded well beyond the original limited mission of submarine propulsion. This section reviews the past and current status of LFR development.

  6. Modeling and analysis of core debris recriticality during hypothetical severe accidents in the Advanced Neutron Source Reactor

    SciTech Connect

    Taleyarkhan, R.P.; Kim, S.H.; Slater, C.O.; Moses, D.L.; Simpson, D.B.; Georgevich, V.

    1993-05-01

    This paper discusses salient aspects of severe-accident-related recriticality modeling and analysis in the Advanced Neutron Source (ANS) reactor. The development of an analytical capability using the KENO V.A-SCALE system is described including evaluation of suitable nuclear cross-section sets to account for the effects of system geometry, mixture temperature, material dispersion and other thermal-hydraulic conditions. Benchmarking and validation efforts conducted with KENO V.A-SCALE and other neutronic codes against critical experiment data are described. Potential deviations and biases resulting from use of the 16-group Hansen-Roach library are shown. A comprehensive test matrix of calculations to evaluate the threat of a recriticality event in the ANS is described. Strong dependencies on geometry, material constituents, and thermal-hydraulic conditions are described. The introduction of designed mitigative features is described.

  7. Iris Data Classification Using Quantum Neural Networks

    SciTech Connect

    Sahni, Vishal; Patvardhan, C. [Faculty of Engineering, Dayalbagh Educational Institute (Deemed University), Dayalbagh, Agra - 282 005 (India)

    2006-11-15

    Quantum computing is a novel paradigm that promises to be the future of computing. The performance of quantum algorithms has proved to be stunning. ANN within the context of classical computation has been used for approximation and classification tasks with some success. This paper presents an idea of quantum neural networks along with the training algorithm and its convergence property. It synergizes the unique properties of quantum bits or qubits with the various techniques in vogue in neural networks. An example application of Fisher's Iris data set, a benchmark classification problem has also been presented. The results obtained amply demonstrate the classification capabilities of the quantum neuron and give an idea of their promising capabilities.

  8. In vitro propagation of Iris pallida.

    PubMed

    Gozu, Y; Yokoyama, M; Nakamura, M; Namba, R; Yomogida, K; Yanagi, M; Nakamura, S

    1993-11-01

    Plantlets were regenerated from callus of Iris pallida, an important perfume plant. Only the leaf base attached to the rhizome had the ability to generate yellow-colored callus on LS medium supplemented with 1 mg/l 2,4-D and 0.1 mg/l KT in the dark. Yellow calli grew with partial differentiation into white tissue, probably embryogenic, during subculture on the same medium with a 16-h photoperiod. Only yellow-colored calli with the white tissue could differentiate into plantlets after transfer to kinetin- or gibberellin- supplemented LS medium. Regenerated plantlets which grew on the medium without growth regulators were transferred to the soil. After 2 years of cultivation in soil, the regenerated plants flowered and formed rhizomes. The components of the essential oil in the rhizome of regenerated plants were essentially the same as those in natural plants. PMID:24196175

  9. EBT reactor analysis

    Microsoft Academic Search

    N. A. Uckan; E. F. Jaeger; R. T. Santoro; D. A. Spong; T. Uckan; L. W. Owen; J. M. Barnes; J. B. McBride

    1983-01-01

    This report summarizes the results of a recent ELMO Bumpy Torus (EBT) reactor study that includes ring and core plasma properties with consistent treatment of coupled ring-core stability criteria and power balance requirements. The principal finding is that constraints imposed by these coupling and other physics and technology considerations permit a broad operating window for reactor design optimization. Within this

  10. Process, Techniques, and Successes in Welding the Dry Shielded Canister Welds of the TMI-2 Reactor Core Debris

    SciTech Connect

    Zirker, Laurence R; Rankin, Richard Allen; Ferrell, Larry Joseph

    2002-02-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) is operated by Bechtel-BWXT Idaho LLC (BBWI), which recently completed a very successful $100 million Three-Mile Island-2 (TMI-2) program for the Department of Energy (DOE). This complex and challenging program used an integrated multidisciplinary team approach that loaded, welded, and transported an unprecedented 25 dry shielded canisters (DSC) in seven months, and did so ahead of schedule. The program moved over 340 canisters of TMI-2 core debris that had been in wet storage into a dry storage facility at the INEEL. The main thrust of this paper is relating the innovations, techniques, approaches, and lessons learned associated to welding of the DSC's. This paper shows the synergism of elements to meet program success and shares these lessons learned that will facilitate success with welding of dry shielded canisters in other DOE complex dry storage programs.

  11. Processes, Techniques, and Successes in Welding the Dry Shielded Canisters of the TMI-2 Reactor Core Debris

    SciTech Connect

    Zirker, L.R.; Rankin, R.A.; Ferrell, L.J.

    2002-01-29

    The Idaho National Engineering and Environmental Laboratory (INEEL) is operated by Bechtel-BWXT Idaho LLC (BBWI), which recently completed a very successful $100 million Three-Mile Island-2 (TMI-2) program for the Department of Energy (DOE). This complex and challenging program used an integrated multidisciplinary team approach that loaded, welded, and transported an unprecedented 25 dry shielded canisters (DSC) in seven months, and did so ahead of schedule. The program moved over 340 canisters of TMI-2 core debris that had been in wet storage into a dry storage facility at the INEEL. The main thrust of this paper is relating the innovations, techniques, approaches, and lessons learned associated to welding of the DSC's. This paper shows the synergism of elements to meet program success and shares these lessons learned that will facilitate success with welding of dry shielded canisters in other DOE complex dry storage programs.

  12. Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

    Microsoft Academic Search

    F. Faghihi; S. M. Mirvakili

    2009-01-01

    Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity (?ex), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and

  13. IRIS Toxicological Review for Acrylamide (Interagency Science Discussion Draft)

    EPA Science Inventory

    EPA is releasing the draft report, Toxicological Review for Acrylamide , that was distributed to Federal agencies and White House Offices for comment during the Science Discussion step of the IRIS Assessment Development Process...

  14. Site Help & Tools - Tools & Databases| IRIS | US EPA

    Science.gov Websites

    System (IRIS), Chemical Carcinogen Research Information System (CCRIS), Toxicology Literature search (TOXLINE), data on mutagenicity studies (GENE-TOX) Developmental and...

  15. IRIS Toxicological Review of Formaldehyde (Inhalation) (External Review Draft 2010)

    EPA Science Inventory

    UPDATE EPA is currently revising its Integrated Risk Information System (IRIS) assessment of formaldehyde to address the 2011 NAS peer review recommendations. This assessment addresses both noncancer and cancer human health effects that are relevant to assessing ...

  16. Adapting the deep burn in-core fuel management strategy for the gas turbine – modular helium reactor to a uranium–thorium fuel

    Microsoft Academic Search

    Alberto Talamo; Waclaw Gudowski

    2005-01-01

    In 1966, Philadelphia Electric has put into operation the Peach Bottom I nuclear reactor, it was the first high temperature gas reactor (HTGR); the pioneering of the helium-cooled and graphite-moderated power reactors continued with the Fort St. Vrain and THTR reactors, which operated until 1989. The experience on HTGRs lead General Atomics to design the gas turbine – modular helium

  17. A bibliography of IRIS-related publications, 2000-2011

    NASA Astrophysics Data System (ADS)

    Muco, B.

    2012-12-01

    Citations and acknowledgements in scientific journals can be an indicator of the role an organization has on the research of that field. Since its formation and incorporation in May 1984, the IRIS Consortium (Incorporated Research Institutions for Seismology) is mentioned more and more as a valuable source of data, instruments and programs in the literature of earth sciences. As a large organization with more than 100 member domestic institutes and about 40 international affiliates, obviously IRIS has a direct impact on the earth sciences through all its programs, projects, workshops, symposia, and news¬letters and as a lively forum for exchanging ideas. In order to maintain support from National Science Foundation (NSF) and the research community, it is important to document the continued use of IRIS facilities in basic research programs. IRIS maintains a database of articles that are based on the use of IRIS facilities or which reference use of IRIS data and resources. Articles in this database have been either been provided to IRIS by the authors or selected through an annual search of a number of prominent journals. A text version of the full bibliographic database is available on the IRIS website and a version in EndNote format is also provided. To provide a more complete bibliography and a consistent evaluation of temporal tends in publications, a special annual search began in 2000 which focused on a subset of key seismology and Earth science journals: Bulletin of Seismological Society of America, Journal of Geophysical Research, Seismological Research Letters, Geophysical Research Letters, Earth and Planetary Science Letters, Physics of the Earth and Planetary Interiors, Tectonophysics, Geophysical Journal International, Nature, Science, Geology and EOS. Using different search engines as Scirus, ScienceDirect, GeoRef, OCLC First Search, EASI Search, NASA Abstract Service etc. for online journals and publishers' databases, we searched for key words (IRIS, GSN, DMS, PASSCAL, USArray etc) in titles, abstracts and text. Most of the selections found by this method were confirmed by reading through online texts or original journals. This bibliography of peer-reviewed articles (excluding abstracts) identified in these key journals for 2000-2011 includes approximately 1800 entries. As for American Geophysical Union (AGU) transaction, the bibliography of IRIS-related abstracts for the abovementioned period includes approximately 1400 abstracts. This study is a clear indicator of making intensive use by the seismological community of the resources that IRIS provides and of the paramount importance this organization has in advancement of seismological research worldwide.

  18. BRCA1-IRIS inactivation sensitizes ovarian tumors to cisplatin.

    PubMed

    Paul, B T; Blanchard, Z; Ridgway, M; ElShamy, W M

    2015-06-01

    Ovarian cancer is the first in mortalities among gynecologic cancers in the United States, often due to late diagnosis and/or acquired platinum-resistant recurrences. This study investigates whether BRCA1-IRIS is a novel treatment target for ovarian cancers and in platinum-resistant recurrences. Here we show that more than half of the ovarian cancer samples analyzed showed BRCA1-IRIS and survivin overexpression and lacked nuclear FOXO3a expression. Normal ovarian epithelial cells overexpressing BRCA1-IRIS formed metastasis in mice when injected in the peritoneal cavity, whereas aggressive ovarian cancer cell lines failed to form tumors or metastases in mice when BRCA1-IRIS was silenced in them. We show that BRCA1-IRIS activates two autocrine signaling loops, brain-derived neurotrophic factor/tyrosine kinase B receptor (BDNF/TrkB) and neuregulin 1 (NRG1)/ErbB2. These loops are involved in anoikis resistance and metastasis promotion. These loops operate in several ovarian cancer cell lines, and BRCA1-IRIS silencing or inactivation using a novel inhibitory peptide renders both non-functional and promoted cell death. In a mouse xenograft model, BRCA1-IRIS inactivation using this novel inhibitory peptide resulted in significant reduction in ovarian tumor growth. More importantly, this treatment sensitized ovarian tumors to low cisplatin concentrations. Taken together, these data strongly suggest that BRCA1-IRIS and/or BDNF/TrkB and NRG1/ErbB2 could serve as rational therapeutic targets for advanced ovarian cancers. PMID:25132263

  19. A Real-Time Focusing Algorithm for Iris Recognition Camera

    Microsoft Academic Search

    Kang Ryoung Park; Jaihie Kim

    2004-01-01

    \\u000a For fast iris recognition, it is very important to capture user’s focused eye image at fast speed. In previous researches\\u000a and systems, they use the focusing method which has been used for general landscape scene without considering the characteristics\\u000a of iris image. So, they take much focusing time, especially in case of the user with glasses. To overcome such problems,

  20. Iris Recognition in Mobile Phone Based on Adaptive Gabor Filter

    Microsoft Academic Search

    Dae Sik Jeong; Hyun-ae Park; Kang Ryoung Park; Jaihie Kim

    2006-01-01

    \\u000a As the security of personal information is becoming more important in mobile phones, we apply iris recognition technology\\u000a to mobile device. Different from conventional iris recognition system used for access control, user puts the mobile phone\\u000a by hands in this case. So, optical and motion blurring happens, frequently. In addition, most users have tendencies to use\\u000a the mobile phone in