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1

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

E-print Network

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

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

2003-01-01

2

Fissioning Plasma Core Reactor  

NASA Technical Reports Server (NTRS)

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.

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

2002-01-01

3

The IRIS General Plant Arrangement  

SciTech Connect

IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the major reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long life core and enhanced safety to address the requirements defined by the US DOE for Generation IV reactors. Bechtel, with Westinghouse consultation, has performed a layout study of the IRIS plant and this paper will discuss the results of this design effort. (authors)

Robertson, J.; Love, J.; Morgan, R. [Bechtel Power Company (United States); Conway, L.E. [Westinghouse Electric Company (United States)

2002-07-01

4

Maintenance Cycle Extension in the IRIS Advanced Light Water Reactor Plant Design  

SciTech Connect

New nuclear power generation in the United States will be realized only if the economic performance can be made competitive with other methods of electrical power generation. The economic performance of a nuclear power plant can be significantly improved by increasing the time spent on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Maintenance includes planned actions (surveillances) and unplanned actions (corrective maintenance) to respond to component degradation or failure. A methodology is described that can be used to resolve, in the design phase, maintenance-related operating cycle length barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the International Reactor, Innovative and Secure (IRIS) design. IRIS is an advanced light water nuclear power plant that is being designed to maximize this on-line generating time by increasing the operating cycle length. This is consequently a maintenance strategy paper using the IRIS plant as the example.Potential IRIS operating cycle length maintenance-related barriers, determined by modification of an earlier operating pressurized water reactor (PWR) plant cycle length analysis to account for differences between the design of IRIS and this operating PWR, are presented. The proposed methodology to resolve these maintenance-related barriers by the design process is described. The results of applying the methodology to two potential IRIS cycle length barriers, relief valve testing and emergency heat removal system testing, are presented.

Galvin, Mark R. [United States Navy (United States); Todreas, Neil E. [Massachusetts Institute of Technology (United States); Conway, Larry E. [Westinghouse Science and Technology (United States)

2003-09-15

5

Reactor core isolation cooling system  

DOEpatents

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.

Cooke, F.E.

1992-12-08

6

Further development around the Hoger Onderwijs reactor of IRI in Delft  

SciTech Connect

The Interfacultair Reactor Instituut (IRI) was founded in 1958, and its reactor first reached criticality in 1963. Until 1987, IRI was an interuniversity institute, owned and directed by the combined universities. Since then it constitutes part of the Delft University of Technology but continues its role as an interuniversity institute. The main facility is the Hoger Onderwijsreactor (HOR), a 2-MW swimming-pool reactor operated 24 h/day, 5 day/week. In the 5-yr working plan of 1988-1993, much attention is being paid to development and construction of new experimental facilities connected to the reactor. A double-stacked mirror neutron guide, a reactor coupled source of variable energy positrons, and an irradiation facility for activation analysis of large samples have been installed. Completion of a neutron reflectometer suitable for application to solids as well as liquids is foreseen for 1993. Further plans for facility development will focus on the construction of a small beam hall and a three- or fourfold stacked mirror neutron guide to provide neutron beams to that hall. The IRI research program will be continued along the lines discussed on earlier occasions but with increasing emphasis on research using neutron beams and positron techniques and nuclear technology. Major new research activities are focused on plant uptake of long-lived fission products and on the behavior of natural nuclides in large-scale industrial processes.

Bruin, M. de (Interfaculty Reactor Inst., Delft (Netherlands))

1992-01-01

7

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

SciTech Connect

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

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

2011-05-31

8

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

SciTech Connect

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)

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

2004-07-01

9

Lateral restraint assembly for reactor core  

DOEpatents

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.

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

1986-01-01

10

Wire core reactor for NTP  

NASA Technical Reports Server (NTRS)

The development of the wire core system for Nuclear Thermal Propulsion (NTP) that took place from 1963 to 1965 is discussed. A wire core consists of a fuel wire with spacer wires. It's an annular flow core having a central control rod. There are actually four of these, with beryllium solid reflectors on both ends and all the way around. Much of the information on the concept is given in viewgraph form. Viewgraphs are presented on design details of the wire core, the engine design, engine weight vs. thrust, a technique used to fabricate the wire fuel element, and axial temperature distribution.

Harty, R. B.

1991-01-01

11

Fabricating the Solid Core Heatpipe Reactor  

NASA Astrophysics Data System (ADS)

The solid core heatpipe nuclear reactor has the potential to be the most dependable concept for the nuclear space power system. The design of the conversion system employed permits multiple failure modes instead of the single failure mode of other concepts. Regardless of the material used for the reactor, either stainless steel, high-temperature alloys, Nb1Zr, Tantalum Alloys or MoRe Alloys, making the solid core by machining holes in a large diameter billet is not satisfactory. This is because the large diameter billet will have large grains that are detrimental to the performance of the reactor due to grain boundary diffusion. The ideal fabrication method for the solid core is by hot isostatic pressure diffusion bonding, (HIPing). By this technique, wrought fine-grained tubes of the alloy chosen are assembled into the final shape with solid cusps and seal welded so that there is a vacuum in between all surfaces to be diffusion bonded. This welded structure is then HIPed for diffusion bonding. A solid core made of Type 321 stainless steel has been satisfactorily produced by Advanced Methods and Materials and is undergoing evaluation by NASA Marshall Space Flight Center.

Ring, Peter J.; Sayre, Edwin D.; Houts, Mike

2006-01-01

12

Search for sterile neutrinos at reactors with a small core  

E-print Network

The sensitivity to the sterile neutrino mixing at very short baseline reactor neutrino experiments is investigated. If the reactor core is relatively large as in the case of commercial reactors, then the sensitivity is lost for $\\Delta m^2 \\gtrsim$ 1 eV$^2$ due to smearing of the reactor core size. If the reactor core is small as in the case of the experimental fast neutron reactor Joyo, the ILL research reactor or the Osiris reactor, on the other hand, then sensitivity to $\\sin^22\\theta_{14}$ can be as good as 0.03 for $\\Delta m^2 \\sim$ several eV$^2$ because of its small size.

Osamu Yasuda

2011-10-12

13

Bowing of core assemblies in advanced liquid metal fast reactors  

Microsoft Academic Search

Two alternative core restraint concepts are considered for a conceptual design of a 900 MWth liquid metal fast reactor core with a heterogeneous layout. The two concepts, known as limited free bowing and free flowering, are evaluated based on core bowing criteria that emphasize the echancement of inherent reactor safety. The core reactivity change during a postulated loss of flow

S. A. Kamal; Y. Orechwa

1986-01-01

14

Multilevel transport solution of LWR reactor cores  

SciTech Connect

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

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

2008-09-01

15

Gas-core reactor power transient analysis  

NASA Technical Reports Server (NTRS)

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.

Kascak, A. F.

1972-01-01

16

Reactor pulse repeatability studies at the annular core research reactor  

SciTech Connect

The Annular Core Research Reactor (ACRR) at Sandia National Laboratories is a water-moderated pool-type reactor designed for testing many types of objects in the pulse and steady-state mode of operations. Personnel at Sandia began working to improve the repeatability of pulse operations for experimenters in the facility. The ACRR has a unique UO{sub 2}-BeO fuel that makes the task of producing repeatable pulses difficult with the current operating procedure. The ACRR produces a significant quantity of photoneutrons through the {sup 9}Be({gamma}, n){sup 8}Be reaction in the fuel elements. The photoneutrons are the result of the gammas produced during fission and in fission product decay, so their production is very much dependent on the reactor power history and changes throughout the day/week of experiments in the facility. Because the photoneutrons interfere with the delayed-critical measurements required for accurate pulse reactivity prediction, a new operating procedure was created. The photoneutron effects at delayed critical are minimized when using the modified procedure. In addition, the pulse element removal time is standardized for all pulse operations with the modified procedure, and this produces less variation in reactivity removal times. (authors)

DePriest, K.R. [Applied Nuclear Technologies, Sandia National Laboratories, Mail Stop 1146, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States); Trinh, T.Q. [Nuclear Facility Operations, Sandia National Laboratories, Mail Stop 0614, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States); Luker, S. M. [Applied Nuclear Technologies, Sandia National Laboratories, Mail Stop 1146, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States)

2011-07-01

17

Multimegawatt NEP with vapor core reactor MHD  

NASA Astrophysics Data System (ADS)

Efforts at the Innovative Nuclear Space Power and Propulsion Institute have assessed the feasibility of combining gaseous or vapor core reactors with magnetohydrodynamic power generators to provide extremely high quality, high density, and low specific mass electrical power for space applications. Innovative shielding strategies are employed to maintain an effective but relatively low mass shield, which is the most dominating part of multi-megawatt space power systems. The fission driven magnetohydrodynamic generator produces tens of kilowatt DC power at specific mass of less than 0.5 kg/kW for the total power system. The MHD output with minor conditioning is coupled to magnetoplasmadynamic thruster to achieve an overall NEP system specific mass of less than 1.0 kg/kW for power levels above 20 MWe. Few other concepts would allow comparable ensuing payload savings and flexible mission abort options for manned flights to Mars for example. .

Smith, Blair; Knight, Travis; Anghaie, Samim

2002-01-01

18

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

Microsoft Academic Search

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

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

2011-01-01

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

Advanced operational strategy for the IRIS reactor: Load follow through mechanical shim (MSHIM)  

Microsoft Academic Search

The renaissance of nuclear power brings more attention to advanced reactor designs and their improved performance and flexibility, including their enhanced load follow capability. Reactor control strategy used to perform transients including power changes has impact on the overall control system design. In particular, as the power change is performed within a load follow maneuver, several modifications occur in the

Fausto Franceschini; Bojan Petrovic

2008-01-01

21

Advanced High Temperature Reactor Neutronic Core Design  

SciTech Connect

The AHTR is a 3400 MW(t) FHR class reactor design concept intended to serve as a central generating station type power plant. While significant technology development and demonstration remains, the basic design concept appears sound and tolerant of much of the remaining performance uncertainty. No fundamental impediments have been identified that would prevent widespread deployment of the concept. This paper focuses on the preliminary neutronic design studies performed at ORNL during the fiscal year 2011. After a brief presentation of the AHTR design concept, the paper summarizes several neutronic studies performed at ORNL during 2011. An optimization study for the AHTR core is first presented. The temperature and void coefficients of reactivity are then analyzed for a few configurations of interest. A discussion of the limiting factors due to the fast neutron fluence follows. The neutronic studies conclude with a discussion of the control and shutdown options. The studies presented confirm that sound neutronic alternatives exist for the design of the AHTR to maintain full passive safety features and reasonable operation conditions.

Ilas, Dan [ORNL] [ORNL; Holcomb, David Eugene [ORNL] [ORNL; Varma, Venugopal Koikal [ORNL] [ORNL

2012-01-01

22

Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)  

NASA Technical Reports Server (NTRS)

Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

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

1977-01-01

23

Solid0Core Heat-Pipe Nuclear Batterly Type Reactor  

SciTech Connect

This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

Ehud Greenspan

2008-09-30

24

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

Microsoft Academic Search

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

Antonio Cammi; Marco E. Ricotti; Alessia Vitulo

2004-01-01

25

Fuel, Core Design and Subchannel Analysis of a Superfast Reactor  

Microsoft Academic Search

A compact supercritical water-cooled fast reactor (superfast reactor) core with a power of 700MWe is designed by using a three-dimensional neutronics thermal-hydraulic coupled method. The core consists of 126 seed assemblies and 73 blanket assemblies. In the seed assemblies, 251 fuel rods, consisting of MOX pellets, stainless steel (SUS304) cladding, and fission gas plenum are arranged into a tight triangle

Liangzhi CAO; Yoshiaki OKA; Yuki ISHIWATARI; Zhi SHANG

2008-01-01

26

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

Federal Register 2010, 2011, 2012, 2013

...Emergency Core Cooling Systems for Boiling-Water Reactors AGENCY: Nuclear Regulatory...Emergency Core Cooling Systems for Boiling- Water Reactors.'' This guide describes...cooling systems (ECCSs) for boiling-water reactors (BWRs). DATES:...

2012-06-15

27

Bowing of core assemblies in advanced liquid metal fast reactors  

SciTech Connect

Two alternative core restraint concepts are considered for a conceptual design of a 900 MWth liquid metal fast reactor core with a heterogeneous layout. The two concepts, known as limited free bowing and free flowering, are evaluated based on core bowing criteria that emphasize the echancement of inherent reactor safety. The core reactivity change during a postulated loss of flow transient is calculated in terms of the lateral displacements and displacement-reactivity-worths of the individual assemblies. The NUBOW-3D computer code is utilized to determine the assembly deformations and interassembly forces that arise when the assemblies are subjected to temperature gradients and irradiation induced creep and swelling during the reactor operation. The assembly ducts are made of the ferritic steel HT-9 and remain in the reactor core for four-years at full power condition. Whereas both restraint systems meet the bowing criteria, a properly designed limited free bowing system appears to be more advantageous than a free flowering system from the point of view of enhancing the reactor inherent safety.

Kamal, S.A.; Orechwa, Y.

1986-01-01

28

Core Physics and Kinetics Calculations for the Fissioning Plasma Core Reactor  

NASA Technical Reports Server (NTRS)

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.

Butler, C.; Albright, D.

2007-01-01

29

CONTROL RODS FOR NUCLEAR REACTOR CORES  

Microsoft Academic Search

A reactor control rod is designed which has increased effectiveness as ; compared with the width of the aperture in the pressure vessel through which it ; passes. The control rod carries six fins, three on each side, and two of the ; fins are fixed while the other, being adjustable, is capable of movement from ; between the fixed

Bell

1961-01-01

30

Unsteady Characteristics of Three-Core Molten Salt Reactor  

NASA Astrophysics Data System (ADS)

Numerical analysis has been performed for load-following capability of a 465 MWth Three-Core Molten Salt Reactor (MSR). “Reactor-slaved-to-turbine control technique” is adopted for reactor control. As for this control technique, a turbine is controlled by a speed regulator of a generator, and subsequently the reactor is controlled so as to follow the turbine output. In this study, the turbine power is rapidly changed in a range of 50-150% of the rated power. Then transient characteristics of fuel salt and graphite temperatures, neutron fluxes, delayed neutron precursors, and reactor output are calculated. The analysis result shows that the reactor output is capable of following the turbine power in the range of the turbine output of 50-150%.

Yamamoto, Takahisa; Mitachi, Koshi; Nishio, Masatoshi

31

Dimensional changes in elements of the BN600 reactor core  

Microsoft Academic Search

Increasing the radiation stability of the structural materials of the core is of great importance for the safe and efficient operation of fast reactors. At the beginning of the operation of the BN-600 reactor, the fuel assembly boxes and the fuel cladding tubes were made from anstenitic 08Khl6NllM3 and 08Khl6N15M3B (t~I-847) steel, respectively, the components of the rods and the

S. E. Astashov; E. A. Kozmanov; A. N. Ogorodov; G. A. Sergeev; V. V. Chuev; A. G. Sheinkman; L. M. Zabud'ko; O. S. Korostin; E. A. Rogov

1993-01-01

32

Thermal barrier and support for nuclear reactor fuel core  

DOEpatents

A thermal barrier/core support for the fuel core of a nuclear reactor having a metallic cylinder secured to the reactor vessel liner and surrounded by fibrous insulation material. A top cap is secured to the upper end of the metallic cylinder that locates and orients a cover block and post seat. Under normal operating conditions, the metallic cylinder supports the entire load exerted by its associated fuel core post. Disposed within the metallic cylinder is a column of ceramic material, the height of which is less than that of the metallic cylinder, and thus is not normally load bearing. In the event of a temperature excursion beyond the design limits of the metallic cylinder and resulting in deformation of the cylinder, the ceramic column will abut the top cap to support the fuel core post.

Betts, Jr., William S. (Del Mar, CA); Pickering, J. Larry (Del Mar, CA); Black, William E. (San Diego, CA)

1987-01-01

33

Gas core reactors for actinide transmutation and breeder applications  

NASA Technical Reports Server (NTRS)

This work consists of design power plant studies for four types of reactor systems: uranium plasma core breeder, uranium plasma core actinide transmuter, UF6 breeder and UF6 actinide transmuter. The plasma core systems can be coupled to MHD generators to obtain high efficiency electrical power generation. A 1074 MWt UF6 breeder reactor was designed with a breeding ratio of 1.002 to guard against diversion of fuel. Using molten salt technology and a superheated steam cycle, an efficiency of 39.2% was obtained for the plant and the U233 inventory in the core and heat exchangers was limited to 105 Kg. It was found that the UF6 reactor can produce high fluxes (10 to the 14th power n/sq cm-sec) necessary for efficient burnup of actinide. However, the buildup of fissile isotopes posed severe heat transfer problems. Therefore, the flux in the actinide region must be decreased with time. Consequently, only beginning-of-life conditions were considered for the power plant design. A 577 MWt UF6 actinide transmutation reactor power plant was designed to operate with 39.3% efficiency and 102 Kg of U233 in the core and heat exchanger for beginning-of-life conditions.

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

1978-01-01

34

A vectorized heat transfer model for solid reactor cores  

SciTech Connect

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

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

1990-01-01

35

Benchmark Evaluation of the NRAD Reactor LEU Core Startup Measurements  

SciTech Connect

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.

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

2011-09-01

36

Dynamic analysis of gas-core reactor system  

NASA Technical Reports Server (NTRS)

A heat transfer analysis was incorporated into a previously developed model CODYN to obtain a model of open-cycle gaseous core reactor dynamics which can predict the heat flux at the cavity wall. The resulting model was used to study the sensitivity of the model to the value of the reactivity coefficients and to determine the system response for twenty specified perturbations. In addition, the model was used to study the effectiveness of several control systems in controlling the reactor. It was concluded that control drums located in the moderator region capable of inserting reactivity quickly provided the best control.

Turner, K. H., Jr.

1973-01-01

37

System Study: Reactor Core Isolation Cooling 1998–2012  

SciTech Connect

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.

T. E. Wierman

2013-10-01

38

One pass core design of a super fast reactor  

SciTech Connect

One pass core design for Supercritical-pressure light water-cooled fast reactor (Super FR) is proposed. The whole core is cooled with upward flow in one through flow pattern like PWR. Compared with the previous two pass core design; this new flow pattern can significantly simplify the core concept. Upper core structure, coolant flow scheme as well as refueling procedure are as simple as in PWR. In one pass core design, supercritical-pressure water is at approximately 25.0 MPa and enters the core at 280 C. degrees and is heated up in one through flow pattern upwardly to the average outlet temperature of 500 C. degrees. Great density change in vertical direction can cause significant axial power offset during the cycle. Meanwhile, Pu accumulated in the UO{sub 2} fuel blanket assemblies also introduces great power increase during cycle, which requires large amount of flow for heat removal and makes the outlet temperature of blanket low at the beginning of equilibrium cycle (BOEC). To deal with these issues, some MOX fuel is applied in the bottom region of the blanket assembly. This can help to mitigate the power change in blanket due to Pu accumulation and to increase the outlet temperature of the blanket during cycle. Neutron transport and thermohydraulics coupled calculation shows that this design can satisfy the requirement in the Super FR principle for both 500 C. degrees outlet temperature and negative coolant void reactivity. (authors)

Liu, Qingjie; Oka, Yoshiaki [Cooperative Major in Nuclear Energy, Waseda University, Tokyo 169-8555 (Japan)

2013-07-01

39

78 FR 64027 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors  

Federal Register 2010, 2011, 2012, 2013

...the U.S. Advanced Pressurized-Water Reactor, U.S. Evolutionary Power...Cooling Systems for Pressurized-Water Reactors.'' Draft Regulatory Guide...Emergency Core Cooling Systems for New Boiling-Water Reactors'' (proposed new RG...

2013-10-25

40

Experimental Criticality Benchmarks for SNAP 10A/2 Reactor Cores  

SciTech Connect

This report describes computational benchmark models for nuclear criticality derived from descriptions of the Systems for Nuclear Auxiliary Power (SNAP) Critical Assembly (SCA)-4B experimental criticality program conducted by Atomics International during the early 1960's. The selected experimental configurations consist of fueled SNAP 10A/2-type reactor cores subject to varied conditions of water immersion and reflection under experimental control to measure neutron multiplication. SNAP 10A/2-type reactor cores are compact volumes fueled and moderated with the hydride of highly enriched uranium-zirconium alloy. Specifications for the materials and geometry needed to describe a given experimental configuration for a model using MCNP5 are provided. The material and geometry specifications are adequate to permit user development of input for alternative nuclear safety codes, such as KENO. A total of 73 distinct experimental configurations are described.

Krass, A.W.

2005-12-19

41

Post impact behavior of mobile reactor core containment systems  

NASA Technical Reports Server (NTRS)

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.

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

1972-01-01

42

Development of an automated core model for nuclear reactors  

Microsoft Academic Search

This is the final report of a three-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The objective of this project was to develop an automated package of computer codes that can model the steady-state behavior of nuclear-reactor cores of various designs. As an added benefit, data produced for steady-state analysis also can be

Mosteller

1998-01-01

43

RMC - A Monte Carlo Code for Reactor Core Analysis  

NASA Astrophysics Data System (ADS)

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.

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

2014-06-01

44

Photoneutron effects on pulse reactor kinetics for the Annular Core Research Reactor (ACRR).  

SciTech Connect

The Annular Core Research Reactor (ACRR) is a swimming-pool type pulsed reactor that maintains an epithermal neutron flux and a nine-inch diameter central dry cavity. One of its uses is neutron and gamma-ray irradiation damage studies on electronic components under transient reactor power conditions. In analyzing the experimental results, careful attention must be paid to the kinetics associated with the reactor to ensure that the transient behavior of the electronic device is understood. Since the ACRR fuel maintains a substantial amount of beryllium, copious quantities of photoneutrons are produced that can significantly alter the expected behavior of the reactor power, especially following a reactor pulse. In order to understand these photoneutron effects on the reactor kinetics, the KIFLE transient reactor-analysis code was modified to include the photoneutron groups associated with the beryllium. The time-dependent behavior of the reactor power was analyzed for small and large pulses, assuming several initial conditions including following several pulses during the day, and following a long steady-state power run. The results indicate that, for these types of initial conditions, the photoneutron contribution to the reactor pulse energy can have a few to tens of percent effect.

Parma, Edward J., Jr.

2009-06-01

45

Transient bowing of core assemblies in advanced liquid metal fast reactors  

Microsoft Academic Search

Two alternative core restraint concepts are considered for a conceptual design of a 900 MWth liquid metal fast reactor core with a heterogeneous layout. The two concepts, known as limited free bowing and free flowering, are evaluated based on core bowing criteria that emphasize the enhancement of inherent reactor safety. The core reactivity change during a postulated loss of flow

S. A. Kamal; Y. Orechwa

1986-01-01

46

Iris microhemangiomas.  

PubMed

Iris microhemangiomas (IM) are benign proliferations of small, twisted blood vessels along the pupillary margin. They are usually bilateral and appear to be developmental in nature. IMs most commonly occur in patients with myotonic dystrophy and adult-onset diabetes mellitus, but have also been associated with respiratory disease, congential heart disease, and central retinal vein occlusion. Additionally, they may be found in individuals without obvious ocular or systemic abnormality. The etiology remains obscure. In this report, illustrative cases are followed by a brief discussion outlining typical clinical manifestations, etiological considerations, and possible implications. PMID:3183274

Roberts, D K; Haine, C L

1988-10-01

47

Coupled simulation of the reactor core using CUPID/MASTER  

SciTech Connect

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)

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

48

MCNP/MCNPX model of the annular core research reactor.  

SciTech Connect

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.

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

2006-10-01

49

A Solid Core Heatpipe Reactor with Cylindrical Thermoelectric Converter Modules  

NASA Astrophysics Data System (ADS)

A nuclear space power system that consists of a solid metal nuclear reactor core with heat pipes carrying energy to a cylindrical thermoelectric converter surrounding each of the heat pipes with a heat pipe radiator surrounding the thermoelectric converter is the most simple and reliable space power system. This means no single point of failure since each heat pipe and cylindrical converter is a separate power system and if one fails it will not affect the others. The heat pipe array in the solid core is designed so that if an isolated heat pipe or even two adjacent heat pipes fail, the remaining heat pipes will still transport the core heat without undue overheating of the uranium nitride fuel. The primary emphasis in this paper is on simplicity, reliability and fabricability of such a space nuclear power source. The core and heat pipes are made of Niobium 1% Zirconium alloy (Nb1Zr), with rhenium lined fuel tubes, bonded together by hot isostatic pressure, (HIPing) and with sodium as the heat pipe working fluid, can be operated up to 1250K. The cylindrical thermoelectric converter is made by depositing the constituents of the converter around a Nb1%Zr tube and encasing it in a Nb 1% Zr alloy tube and HIPing the structure to get final bonding and to produce residual compressive stresses in all brittle materials in the converter. A radiator heat pipe filled with potassium that operates at 850K is bonded to the outside of the cylindrical converter for cooling. The solid core heat pipe and cylindrical converter are mated by welding during the final assembly. A solid core reactor with 150 heat pipes with a 0.650-inch (1.65 cm) ID and a 30-inch (76.2 cm) length with an output of 8 Watts per square inch as demonstrated by the SP100 PD2 cell tests will produce about 80 KW of electrical power. An advanced solid core reactor made with molybdenum 47% rhenium alloy, with lithium heat pipes and the PD2 theoretical output of 11 watts per square inch or advanced higher temperature converter to operate at 1350K could produce a greater output of approximately 100KW.

Sayre, Edwin D.; Vaidyanathan, Sam

2006-01-01

50

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

SciTech Connect

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.

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

1989-01-01

51

Nuclear reactor spacer grid and ductless core component  

DOEpatents

The invention relates to a nuclear reactor spacer grid member for use in a liquid cooled nuclear reactor and to a ductless core component employing a plurality of these spacer grid members. The spacer grid member is of the egg-shell type and is constructed so that the walls of the cell members of the grid member are formed of a single thickness of metal to avoid tolerance problems. Within each cell member is a hydraulic spring which laterally constrains the nuclear material bearing rod which passes through each cell member against a hardstop in response to coolant flow through the cell member. This hydraulic spring is also suitable for use in a water cooled nuclear reactor. A core component constructed of, among other components, a plurality of these spacer grid members, avoids the use of a full length duct by providing spacer sleeves about the sodium tubes passing through the spacer grid members at locations between the grid members, thereby maintaining a predetermined space between adjacent grid members.

Christiansen, David W. (Kennewick, WA); Karnesky, Richard A. (Richland, WA)

1989-01-01

52

Replacement fuel scoping studies for the Annular Core Research Reactor  

SciTech Connect

Sandia National Laboratories Annular Core Research Reactor (ACRR) is undertaking a new mission for the Department of Energy: production of the radioisotope {sup 99}Mo used in nuclear medicine applications. Isotope production is significantly different from previous programs conducted at the ACRR that typically required high intensity, short duration pulses. The current UO{sub 2}-BeO fuel will power the initial startup phase of the production program, and can perform exceptionally well for this mission. However, this type of fuel is no longer available, commercially or otherwise. This paper presents the results of some preliminary studies of commercially available fuels.

Hays, K.; Martin, L.; Parma, E.

1995-07-01

53

In-Core Instrumentation and Fast Digital Data Acquisition for the Sefor Nuclear Test Reactor  

Microsoft Academic Search

The primary purpose of the SEFOR nuclear test reactor is to measure the dynamic behavior of a fast reactor core. Special in-core instrumentation has been developed for fuel and sodium temperature and sodium flow rate measurements. Neutron detectors have been built and tested which will measure reactor power peaks up to 106 MW. A fast computer-controlled data acquisition system allows

H. W. Glauner; G. R. Pflasterer; J. F. Momberger

1967-01-01

54

Lessons learned from Sandia National Laboratories' operational readiness review of the Annular Core Research Reactor  

Microsoft Academic Search

The Sandia National Laboratories (SNL) Annular Core Research Reactor (ACRR) (a Hazard Category 2 nuclear reactor facility) was defueled in early 1997 to convert the reactor core and control system to produce medical radioisotopes for the US Department of Energy (DOE) Medical Isotope Production Program. The DOE determined that an operational readiness review (ORR) per DOE 5480.31 or DOE 420.1

A. O. Bendure; J. W. Bryson

1999-01-01

55

Axial offset control of PWR nuclear reactor core using intelligent techniques  

Microsoft Academic Search

Improved load following capability is one of the main technical performances of advanced PWR (APWR). Controlling the nuclear reactor core during load following operation encounters some difficulties. These difficulties mainly arise from nuclear reactor core limitations in local power peaking, while the core is subject to large and sharp variation of local power density during transients. Axial offset (AO) is

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

2004-01-01

56

Reactor core configuration and important systems related to physics tests of Daya Bay NPP.  

National Technical Information Service (NTIS)

A brief introduction to reactor core configuration and important systems related to physics tests of Daya Bay NPP is given. These systems involve the reactor core system (COR), the full length rod control system (RGL), the in-core instrumentation system (...

Tao Shaoping

1995-01-01

57

Core conversion of the Portuguese research reactor to LEU fuel  

SciTech Connect

Core conversion of the Portuguese Research Reactor (RPI) to LEU fuel is being performed within IAEA's Technical Cooperation project POR/4/016, with financial support from the US and Portugal. CERCA was selected as manufacturer of the LEU assemblies by the IAEA after an international call for bids. CERCA provided a comprehensive package to the RPI which included the mechanical verification of the design of the assemblies, their manufacture and arrangements for a joint inspection of the finished assemblies. The LEU fuel assemblies were manufactured within 8 months upon final approval of the design. The safety analyses for the core conversion to LEU fuel were made with the assistance of the RERTR program and were submitted for review by the IAEA and by Portuguese authorities in January 2007. Revised documents were submitted in June 2007 addressing the issues raised during review. Regulatory approval was received in early August and core conversion was done in early September. All measured safety parameters are within the defined acceptance limits. Operation at full power is expected by the end of October. (author)

Marques, J.G.; Ramos, A.R. [Instituto Tecnologico e Nuclear, Estrada Nacional 10, P-2686-953 Sacavem (Portugal); Kocher, A. [AREVA CERCA, 10 rue Juliette Recamier, F-69456 Lyon Cedex 06 (France)

2008-07-15

58

Reactor core design and modeling of the MIT research reactor for conversion to LEU  

SciTech Connect

Feasibility design studies for conversion of the MIT Research Reactor (MITR) to LEU are described. Because the reactor fuel has a rhombic cross section, a special input processor was created in order to model the reactor in great detail with the REBUS-PC diffusion theory code, in 3D (triangular-z) geometry. Comparisons are made of fuel assembly power distributions and control blade worth vs. axial position, between REBUS-PC results and Monte Carlo predictions from the MCNP code. Results for the original HEU core at zero burnup are also compared with measurement. These two analysis methods showed remarkable agreement. Ongoing fuel cycle studies are summarized. A status report will be given as to results thus far that affect key design decisions. Future work plans and schedules to achieve completion of the conversion are presented. (author)

Newton, Thomas H. Jr. [Nuclear Reactor Laboratory, Massachusetts Institute of Technology, 138 Albany St., Cambridge, MA 02139 (United States); Olson, Arne P.; Stillman, John A. [RERTR Program, Argonne National Laboratory, Argonne, IL 60439 (United States)

2008-07-15

59

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.

60

Coupled edge-core model of fusion reactor  

NASA Astrophysics Data System (ADS)

A model has been developed which is capable to describe in a self consistent way the plasma dynamics in the centre and edge region of a fusion reactor. The core plasma is treated in the frame of the 0D model in which an empirical scaling law for the energy confinement time is included. The model accounts for energy losses due to Bremsstrahlung and line radiation as well as alpha particle heating. A 1D analytical model for plasma and impurity transport outside the last close magnetic surface (LCMS) is applied. The model accounts for the strong gradients of the plasma parameters along the magnetic field lines in the divertor. The sputtering phenomena at the plate and radiating cooling by injected impurities are treated self consistently in the model. The model has been used to investigate operating regimes of the ignition experiment. Analysis have been performed for different first wall materials (C, Ni, Mo, W) for ITER like tokamak.

Zagórski, R.; Kulinski, S.; Scholz, M.

1997-10-01

61

Plasma core reactor simulations using RF uranium seeded argon discharges  

NASA Technical Reports Server (NTRS)

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.

Roman, W. C.

1975-01-01

62

Post impact behavior of mobile reactor core containment systems.  

NASA Technical Reports Server (NTRS)

In the future, nuclear assemblies containing fission products will be transported at high speeds. An example is a reactor supplying power to a large subsonic airplane. In this case an accident can occur resulting in a ground impact at speeds up to 1000 ft/sec. This paper analyzes the containment vessel temperatures after impact and attempts to understand the design variables that affect the post impact survival of the system. The heat transfer analysis includes conduction, radiation, and convection in addition to the core material heats of fusion and vaporization under partial-burial conditions. Also, included is the fact that fission products vaporize and transport radially outward and condense on cooler surfaces, resulting in a moving heat source.

Puthoff, R. L.; Parker, W. G.; Van Bibber, L. E.

1972-01-01

63

Plasma core reactor simulations using RF uranium seeded argon discharges  

NASA Technical Reports Server (NTRS)

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.

Roman, W. C.

1976-01-01

64

Core reactivity estimation in space reactors using recurrent dynamic networks  

NASA Technical Reports Server (NTRS)

A recurrent multilayer perceptron network topology is used in the identification of nonlinear dynamic systems from only the input/output measurements. The identification is performed in the discrete time domain, with the learning algorithm being a modified form of the back propagation (BP) rule. The recurrent dynamic network (RDN) developed is applied for the total core reactivity prediction of a spacecraft reactor from only neutronic power level measurements. Results indicate that the RDN can reproduce the nonlinear response of the reactor while keeping the number of nodes roughly equal to the relative order of the system. As accuracy requirements are increased, the number of required nodes also increases, however, the order of the RDN necessary to obtain such results is still in the same order of magnitude as the order of the mathematical model of the system. It is believed that use of the recurrent MLP structure with a variety of different learning algorithms may prove useful in utilizing artificial neural networks for recognition, classification, and prediction of dynamic systems.

Parlos, Alexander G.; Tsai, Wei K.

1991-01-01

65

Core reactivity estimation in space reactors using recurrent dynamic networks  

SciTech Connect

A recurrent Multi Layer Perceptron (MLP) network topology is used in the identification of nonlinear dynamic systems from only the input/output measurements. This effort is part of a research program devoted in developing real-time diagnostics and predictive control techniques for large-scale complex nonlinear dynamic systems. The identification is performed in the discrete time domain, with the learning algorithm being a modified form of the Back Propagation (BP) rule. The Recurrent Dynamic Network (RDN) developed is applied for the total core reactivity prediction of a spacecraft reactor from only neutronic power level measurements. Results indicate that the RDN can reproduce the nonlinear response of the reactor while keeping the number of nodes roughly equal to the relative order of the system. As accuracy requirements are increased, the number of required nodes also increases, however, the order of the RDN necessary to obtain such results is still in the same order of magnitude as the order of the matematical model of the system. There are a number of issues identified regarding the behavior of the RDN, which at this point are unresolved and require further research. Nevertheless, it is believed that use of the recurrent MLP structure with a variety of different learning algorithms may prove useful in utilizing artifical neural networks (ANNs) for recognition, classification and prediction of dynamic systems.

Parlos, A.G. (Departments of Nuclear Engineering, Texas A M University, College Station, Texas (USA)); Tsai, W.K. (Department of Electrical and Computer Engineering, University of California at Irvine, Irvine, California (USA))

1991-01-10

66

Sensitivity of detecting in-core vibrations and boiling in pressurized water reactors using ex-core neutron detectors  

Microsoft Academic Search

Neutron transport and diffusion theory space- and energy-dependent reactor kinetics calculations were performed in the frequency domain to determine the sensitivity of an ex-core neutron detector to in-core vibrations and coolant boiling in a PWR. Results of these calculations indicate that the ex-core detectors are sensitive to neutron sources, to vibrations, and to boiling occurring over large regions of the

F. J. Sweeney; J. P. A. Renier

1984-01-01

67

Determination of optimal core fissile loadings in the TREAT Upgrade reactor  

Microsoft Academic Search

The TREAT Upgrade (TU) reactor design is presently nearing completion. The reactor will be used to test LMFBR fuel under simulated accident conditions. The physics of the TU core is complicated by a number of factors related to the planned application of the facility. In this paper the design approach used to produce the core fissile loading spatial distribution needed

S. K. Bhattacharyya; R. M. Lell; A. J. Ulrich; S. Yang

1983-01-01

68

The determination of neutron energy spectrum in reactor core C1 of reactor VR-1 Sparrow  

SciTech Connect

This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe. Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (author)

Vins, M. [Department of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University, V Holesovickach 2, 180 00 Prague 8 (Czech Republic)], E-mail: vinsmiro@seznam.cz

2008-07-15

69

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

NASA Technical Reports Server (NTRS)

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.

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

1974-01-01

70

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

DOEpatents

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.

Tokarz, R.D.

1981-10-27

71

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

NASA Technical Reports Server (NTRS)

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.

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

1991-01-01

72

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

SciTech Connect

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

Dobranich, D.

1987-05-01

73

Ordinal Measures for Iris Recognition  

Microsoft Academic Search

Images of a human iris contain rich texture information useful for identity authentication. A key and still open issue in iris recognition is how best to represent such textural information using a compact set of features (iris features). In this paper, we propose using ordinal measures for iris feature representation with the objective of characterizing qualitative relationships between iris regions

Zhenan Sun; Tieniu Tan

2009-01-01

74

Iris Recognition Using Wavelet Features  

Microsoft Academic Search

The traditional iris recognition systems require equal high quality human iris images. A cheap image acquisition system has difficulty in capturing equal high quality iris images. This paper describes a new feature representation method for iris recognition robust to noises. The disc-shaped iris image is first convolved with a low pass filter along the radial direction. Then, the radially smoothed

Jaemin Kim; Seongwon Cho; Jinsu Choi; Robert J. Marks II

2004-01-01

75

Interaction of the control system with core nuclear design for fast spectrum space power reactors  

NASA Astrophysics Data System (ADS)

The reactor control system and operating strategy are essential factors in assessing reactor reliability and safety. The control system and its mode of operation also exert major influences on mechanical design of core components and all aspects of nuclear design. This is especially true of reactors for space power applications because of the imposed requirements regarding compactness, minimum mass, and long term operational reliability without external intervention or maintenance. Generic features of the interaction between nuclear design and reactor control system design for fast spectrum space power reactors are outlined. Several basic control concepts were analyzed. These included ex-core control drums, in-core control rods, burnable poisons, dispersed poisons in the core, and movable fuel segments or regions. Cross sections for calculations were generated with MC sup 2 -2, and neutronics calculations were performed with the VIM Monte Carlo code, ONEDANT, and DIF3D.

Lell, R. M.; Hanan, N. A.

76

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

SciTech Connect

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.

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

77

McCARD for Neutronics Design and Analysis of Research Reactor Cores  

NASA Astrophysics Data System (ADS)

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.

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

2014-06-01

78

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

NASA Technical Reports Server (NTRS)

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.

Weinstein, H.; Lavan, Z.

1975-01-01

79

Burnup concept for a long-life fast reactor core using MCNPX.  

SciTech Connect

This report describes a reactor design with a burnup concept for a long-life fast reactor core that was evaluated using Monte Carlo N-Particle eXtended (MCNPX). The current trend in advanced reactor design is the concept of a small modular reactor (SMR). However, very few of the SMR designs attempt to substantially increase the lifetime of a reactor core, especially without zone loading, fuel reshuffling, or other artificial mechanisms in the core that %E2%80%9Cflatten%E2%80%9D the power profile, including non-uniform cooling, non-uniform moderation, or strategic poison placement. Historically, the limitations of computing capabilities have prevented acceptable margins in the temporal component of the spatial excess reactivity in a reactor design, due primarily to the error in burnup calculations. This research was performed as an initial scoping analysis into the concept of a long-life fast reactor. It can be shown that a long-life fast reactor concept can be modeled using MCNPX to predict burnup and neutronics behavior. The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional Light Water Reactors (LWRs) or other SMR designs. For the purpose of this study, a single core design was investigated: a relatively small reactor core, yielding a medium amount of power (~200 to 400 MWth). The results of this scoping analysis were successful in providing a preliminary reactor design involving metal U-235/U-238 fuel with HT-9 fuel cladding and sodium coolant at a 20% volume fraction.

Holschuh, Thomas Vernon,; Lewis, Tom Goslee,; Parma, Edward J.,

2013-02-01

80

Thermal hydraulics analysis of the MIT research reactor in support of a low enrichment uranium (LEU) core conversion  

E-print Network

The MIT research reactor (MITR) is converting from the existing high enrichment uranium (HEU) core to a low enrichment uranium (LEU) core using a high-density monolithic UMo fuel. The design of an optimum LEU core for the ...

Ko, Yu-Chih, Ph. D. Massachusetts Institute of Technology

2008-01-01

81

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

E-print Network

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 ...

Kersting, Alyssa (Alyssa Rae)

2011-01-01

82

MELCOR adaptation and validation for modeling of N Reactor core phenomena  

SciTech Connect

MELCOR has been adapted for use in modeling the N Reactor core as a part of the recently completed N Reactor probabilistic risk assessment. Significant adaptation of MELCOR was required because of the horizontal, water cooled, graphite-moderated nature of the N Reactor core. The generation and verification of the revised N Reactor core model are described in this paper. A hydrogen production and core damage benchmark calculation is presented in which all significant parameters calculated by MELCOR agreed with those in the reference calculation to within approximately 10%. The reference calculation required many CRAY CPU hours, while the MELCOR calculation was completed in less than 20 CPU minutes on a VAX 8700. 7 refs., 3 figs., 1 tab.

Wyss, G.D.; Summers, R.M.; Miller, L.A.

1990-01-01

83

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

E-print Network

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 ...

Aldrich, David C.

1979-01-01

84

Pigments of Iris pseudacorus  

Microsoft Academic Search

A NUMBER of blue and purple varieties of the Iris family have been included in their ``Survey of Anthocyanins'' by G. M. and R. Robinson, who have specified the particular anthocyanidin present in each case. They have found, for example, that Iris miranda1 contains a complex delphinidin diglycoside and that a malvidin 3: 5 dimonoside mixed with what is probably

W. F. O'Connor; P. J. Drumm

1941-01-01

85

How iris recognition works  

Microsoft Academic Search

Algorithms developed by the author for recogniz- ing persons by their iris patterns have now been tested in six eld and laboratory trials, producing no false matches in several million comparison tests. The recognition principle is the failure of a test of statis- tical independence on iris phase structure encoded by multi-scale quadrature wavelets. The combinatorial complexity of this phase

John Daugman

2004-01-01

86

How iris recognition works  

Microsoft Academic Search

Abstract: Algorithms developed by the author for recognizingpersons by their iris patterns have now been tested in manyfield and laboratory trials, producing no false matches in severalmillion comparison tests. The recognition principle is the failureof a test of statistical independence on iris phase structure encodedby multi-scale quadrature wavelets. The combinatorial complexityof this phase information across different persons spans about249 degrees

John Daugman

2002-01-01

87

Pebble Bed Reactor: core physics and fuel cycle analysis  

Microsoft Academic Search

The Pebble Bed Reactor is a gas-cooled, graphite-moderated high-temperature reactor that is continuously fueled with small spherical fuel elements. The projected performance was studied over a broad range of reactor applicability. Calculations were done for a burner on a throwaway cycle, a converter with recycle, a prebreeder and breeder. The thorium fuel cycle was considered using low, medium (denatured), and

D. R. Vondy; B. A. Worley

1979-01-01

88

Power distributions in fresh and depleted LEU and HEU cores of the MITR reactor.  

SciTech Connect

The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (UMo) is expected to allow the conversion of U.S. domestic high performance reactors like the MITR-II reactor. Toward this goal, core geometry and power distributions are presented. Distributions of power are calculated for LEU cores depleted with MCODE using an MCNP5 Monte Carlo model. The MCNP5 HEU and LEU MITR models were previously compared to experimental benchmark data for the MITR-II. This same model was used with a finer spatial depletion in order to generate power distributions for the LEU cores. The objective of this work is to generate and characterize a series of fresh and depleted core peak power distributions, and provide a thermal hydraulic evaluation of the geometry which should be considered for subsequent thermal hydraulic safety analyses.

Wilson, E.H.; Horelik, N.E.; Dunn, F.E.; Newton, T.H., Jr.; Hu, L.; Stevens, J.G. (Nuclear Engineering Division); (2MIT Nuclear Reactor Laboratory and Nuclear Science and Engineering Department)

2012-04-04

89

Lessons Learned from Sandia National Laboratories' Operational Readiness Review of the Annular Core Research Reactor (ACRR)  

Microsoft Academic Search

The Sandia ACRR (a Hazard Category 2 Nuclear Reactor Facility) was defueled in June 1997 to modify the reactor core and control system to produce medical radioisotopes for the Department of Energy (DOE) Isotope Production Program. The DOE determined that an Operational Readiness Review (ORR) was required to confirm readiness to begin operations within the revised safety basis. This paper

Albert O. Bendure; James W. Bryson

1999-01-01

90

100-KE REACTOR CORE REMOVAL PROJECT ALTERNATIVE ANALYSIS WORKSHOP REPORT  

SciTech Connect

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 Thermal Shield and Core Removal and develop the path forward to proceed with project delivery. Prior to this workshop, the S.A. Robotics (SAR) Obstruction Removal Alternatives Analysis (565-DLV-062) report was issued, for use prior to and throughout the session, to all the team members. The multidisciplinary team consisted ofrepresentatives from 100-KE Project Management, Engineering, Radcon, Nuclear Safety, Fire Protection, Crane/Rigging, SAR Project Engineering, the Department of Energy Richland Field Office, Environmental Protection Agency, Washington State Department of Ecology, Defense Nuclear Facility Safety Board, and Deactivation and Decommission subject matter experts from corporate CH2M HILL and Lucas. Appendix D contains the workshop agenda, guidelines and expectations, opening remarks, and attendance roster going into followed throughout the workshop. The team was successful in selecting the preferred alternative and developing an eight-point path forward action plan to proceed with conceptual design. Conventional Demolition was selected as the preferred alternative over two other alternatives: Diamond Wire with Options, and Harmonic Delamination with Conventional Demolition. The teams preferred alternative aligned with the SAR Obstruction Removal Alternative Analysis report conclusion. However, the team identified several Path Forward actions, in Appendix A, which upon completion will solidify and potentially enhance the Conventional Demolition alternative with multiple options and approaches to achieve project delivery. In brief, the Path Forward was developed to reconsider potential open air demolition areas; characterize to determine if any zircaloy exists, evaluate existing concrete data to determine additional characterization needs, size the new building to accommodate human machine interface and tooling, consider bucket thumb and use ofshape-charges in design, and finally to utilize complex-wide and industry explosive demolition lessons learned in the design approach. Appendix B documents these results from the team's use ofValue Engineering process tools entitled Weighted Analysis Alternative Matrix, Matrix Conclusions, Evaluation Criteria, and Alternative Advantages and Disadvantages. These results were further supported with the team's validation of parking-lot information sheets: memories (potential ideas to consider), issues/concerns, and assumptions, contained in Appendix C. Appendix C also includes the recorded workshop flipchart notes taken from the SAR Alternatives and Project Overview presentations. The SAR workshop presentations, including a 3-D graphic illustration demonstration video have been retained in the CHPRC project file, and were not included in this report due to size limitations. The workshop concluded with a round robin close-out where each member was engaged for any last minute items and meeting utility. In summary, the team felt the session was value added and looked forward to proceeding with the recommended actions and conceptual design.

HARRINGTON RA

2010-01-15

91

Sensitivity of pressurized water reactor source term inventory and decay power to core management parameters  

Microsoft Academic Search

ORIGEN2 was used to develop a data base of pressurized water reactor isotopic concentrations at various times after discharge with core burnup, specific power, enrichment, and neutron spectrum as variables. Results were analyzed to determine source term sensitivity to core management. Fuel rod power history was found to have an important effect on the source term. Activity and decay power

J. K. Wheeler; A. Sesonske

1986-01-01

92

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

Microsoft Academic Search

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

Edwin D. Sayre; Peter J. Ring; Neil Brown; Norbert B. Elsner; John C. Bass

2008-01-01

93

An efficient iris recognition system  

Microsoft Academic Search

Iris recognition, a relatively new biometric technology, has great advantages, such as variability, stability and security, and is most promising for high security environments. A new iris recognition algorithm is proposed in this paper, which adopts Independent Component Analysis (ICA) to extract iris texture feature and a competitive learning mechanism to recognize iris patterns. Experimental results show that the algorithm

Ya-Ping Huang; Si-Wei Luo; En-Yi Chen

2002-01-01

94

Effects of core excess reactivity and coolant average temperature on maximum operable time of NIRR-1 miniature neutron source reactor  

Microsoft Academic Search

We appraised in this study the effects of core excess reactivity and average coolant temperature on the operable time of the Nigeria Research Reactor-1 (NIRR-1), which is a miniature neutron source reactor (MNSR). The duration of the reactor operating time and fluence depletion under different operation mode as well as change in core excess reactivity with temperature coefficient was investigated

Y. A. Ahmed; I. B. Mansir; I. Yusuf; G. I. Balogun; S. A. Jonah

2011-01-01

95

A Practical Iris Recognition Algorithm  

Microsoft Academic Search

This paper proposes an iris recognition algorithm based on 2D zero-crossing detection and similarity classifier. Whole system is consisting of eye image capture, iris boundary localization, iris region normalization, feature extraction, pattern match, and yes\\/no decision. In iris feature extraction stage, iris normal region is filtered by using low frequency filter firstly, then the texture features are extracted by using

Qichuan Tian; Zhengguang Liu; Linsheng Li; Zhiyi Sun

2006-01-01

96

Optimization of hydride fueled pressurized water reactor cores  

E-print Network

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 ...

Shuffler, Carter Alexander

2004-01-01

97

Heat exchanger for reactor core and the like  

DOEpatents

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.

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

1986-01-01

98

Cavity temperature and flow characteristics in a gas-core test reactor  

NASA Technical Reports Server (NTRS)

A test reactor concept for conducting basic studies on a fissioning uranium plasma and for testing various gas-core reactor concepts is analyzed. The test reactor consists of a conventional fuel-element region surrounding a 61-cm-(2-ft-) diameter cavity region which contains the plasma experiment. The fuel elements provide the neutron flux for the cavity region. The design operating conditions include 60-MW reactor power, 2.7-MW cavity power, 200-atm cavity pressure, and an average uranium plasma temperature of 15,000 K. The analytical results are given for cavity radiant heat transfer, hydrogen transpiration cooling, and uranium wire or powder injection.

Putre, H. A.

1973-01-01

99

Annular core for Modular High-Temperature Gas-Cooled Reactor (MHTGR)  

SciTech Connect

The active core of the 350 MW(t) MHTGR is annular in configuration, shaped to provide a large external surface-to-volume ratio for the transport of heat radially to the reactor vessel in case of a loss of coolant flow. For a given fuel temperature limit, the annular core provides approximately 40% greater power output over a typical cylindrical configuration. The reactor core is made up of columns of hexagonal blocks, each 793-mm high and 360-mm wide. The active core is 3.5 m in o.d., 1.65 m in i.d., and 7.93-m tall. Fuel elements contain TRISO-coated microspheres of 19.8% enriched uranium oxycarbide and of fertile thorium oxide. The core is controlled by 30 control rods which enter the inner and outer side reflectors from above.

Turner, R.F.; Baxter, A.M.; Stansfield, O.M.; Vollman, R.E.

1987-08-01

100

IRIS Launch Animation  

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...

101

Internal Control Rod Drive Mechanisms, Design Options for IRIS  

SciTech Connect

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)

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

2004-07-01

102

Iris recognition technology  

Microsoft Academic Search

IriScan Inc. has been developing an identification\\/verification system capable of positively identifying and verifying the identity of individuals without physical contact or human intervention. A new technology, using the unique patterns of the human iris, shows promise of overcoming previous shortcomings and providing positive identification of an individual without contact or invasion, at extremely high confidence levels. The video-based system

G. O. Williams

1997-01-01

103

Transient bowing of core assemblies in advanced liquid metal fast reactors  

SciTech Connect

Two alternative core restraint concepts are considered for a conceptual design of a 900 MWth liquid metal fast reactor core with a heterogeneous layout. The two concepts, known as limited free bowing and free flowering, are evaluated based on core bowing criteria that emphasize the enhancement of inherent reactor safety. The core reactivity change during a postulated loss of flow transient is calculated in terms of the lateral displacements and displacement-reactivity-worths of the individual assemblies. The NUBOW-3D computer code is utilized to determine the assembly deformations and interassembly forces that arise when the assemblies are subjected to temperature gradients and irradiation induced creep and swelling during the reactor operation. The assembly ducts are made of the ferritic steel HT-9 and remain in the reactor core for four-years at full power condition. Whereas both restraint systems meet the bowing criteria, a properly designed limited free bowing system appears to be more advantageous than a free flowering system from the point of view of enhancing the reactor inherent safety.

Kamal, S.A.; Orechwa, Y.

1986-01-01

104

A study on reactor core failure thresholds to safety operation of LMFBR  

SciTech Connect

Japan Nuclear Safety Organization (JNES) has been developing the methodology and computer codes for applying level-1 PSA to LMFBR. Many of our efforts have been directed to the judging conditions of reactor core damage and the time allowed to initiate the accident management. Several candidates of the reactor core failure threshold were examined to a typical proto-type LMFBR with MOX fuel based on the plant thermal-hydraulic analyses to the actual progressions leading to the core damage. The results of the present study showed that the judging condition of coolant-boundary integrity failure, 750 degree-C of the boundary temperature, is enough as the threshold of core damage to PLOHS (protected loss-of-heat sink). High-temperature fuel cladding creep failure will not take place before the coolant-boundary reaches the judging temperature and sodium boiling will not occur due to the system pressure rise. In cases of ATWS (anticipated transient without scrum) the accident progression is so fast and the reactor core damage will be inevitable even a realistic negative reactivity insertion due to the temperature rise is considered. Only in the case of ULOHS (unprotected loss-of-heat sink) a relatively long time of 11 min will be allowed till the shut-down of the reactor before the core damage. (authors)

Kazuo, Haga; Hiroshi, Endo; Tomoko, Ishizu; Yoshihisa, Shindo [Japan Nuclear Energy Safety Organization, Safety Analysis and Evaluation Division, Kamiya-cho MT Bldg., 4-3-20, Toranomon, Minato-ku, Tokyo (Japan)

2006-07-01

105

Core and Refueling Design Studies for the Advanced High Temperature Reactor  

SciTech Connect

The Advanced High Temperature Reactor (AHTR) is a design concept for a central generating station type [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. This report presents the current status of ongoing design studies of the core, in-vessel structures, and refueling options for the AHTR. The AHTR design remains at the notional level of maturity as important material, structural, neutronic, and hydraulic issues remain to be addressed. The present design space exploration, however, indicates that reasonable options exist for the AHTR core, primary heat transport path, and fuel cycle provided that materials and systems technologies develop as anticipated. An illustration of the current AHTR core, reactor vessel, and nearby structures is shown in Fig. ES1. The AHTR core design concept is based upon 252 hexagonal, plate fuel assemblies configured to form a roughly cylindrical core. The core has a fueled height of 5.5 m with 25 cm of reflector above and below the core. The fuel assembly hexagons are {approx}45 cm across the flats. Each fuel assembly contains 18 plates that are 23.9 cm wide and 2.55 cm thick. The reactor vessel has an exterior diameter of 10.48 m and a height of 17.7 m. A row of replaceable graphite reflector prismatic blocks surrounds the core radially. A more complete reactor configuration description is provided in Section 2 of this report. The AHTR core design space exploration was performed under a set of constraints. Only low enrichment (<20%) uranium fuel was considered. The coated particle fuel and matrix materials were derived from those being developed and demonstrated under the Department of Energy Office of Nuclear Energy (DOE-NE) advanced gas reactor program. The coated particle volumetric packing fraction was restricted to at most 40%. The pressure drop across the core was restricted to no more than 1.5 atm during normal operation to minimize the upward force on the core. Also, the flow velocity in the core was restricted to 3 m/s to minimize erosion of the fuel plates. Section 3.1.1 of this report discusses the design restrictions in more detail.

Holcomb, David Eugene [ORNL] [ORNL; Ilas, Dan [ORNL] [ORNL; Varma, Venugopal Koikal [ORNL] [ORNL; Cisneros, Anselmo T [ORNL] [ORNL; Kelly, Ryan P [ORNL] [ORNL; Gehin, Jess C [ORNL] [ORNL

2011-09-01

106

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

NASA Astrophysics Data System (ADS)

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

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

2014-06-01

107

Sodium Based Heat Pipe Modules for Space Reactor Concepts: Stainless Steel SAFE-100 Core  

NASA Technical Reports Server (NTRS)

A heat pipe cooled reactor is one of several candidate reactor cores being considered for advanced space power and propulsion systems to support future space exploration applications. Long life heat pipe modules, with designs verified through a combination of theoretical analysis and experimental lifetime evaluations, would be necessary to establish the viability of any of these candidates, including the heat pipe reactor option. A hardware-based program was initiated to establish the infrastructure necessary to build heat pipe modules. This effort, initiated by Los Alamos National Laboratory and referred to as the Safe Affordable Fission Engine (SAFE) project, set out to fabricate and perform non-nuclear testing on a modular heat pipe reactor prototype that can provide 100 kilowatt from the core to an energy conversion system at 700 C. Prototypic heat pipe hardware was designed, fabricated, filled, closed-out and acceptance tested.

Martin, James J.; Reid, Robert S.

2004-01-01

108

R&D program for core instrumentation improvements devoted for French Sodium fast reactors  

Microsoft Academic Search

Under the framework of French R&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&D sets out to achieve better inspection, maintenance, availability and decommissioning. In particular the instrumentation requirements for core monitoring and detection in the case

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

2011-01-01

109

Conceptual core design of Advanced Recycling Reactor based on mature technologies  

Microsoft Academic Search

This paper presents about comprehensive investigations into Advanced Recycling Reactor (ARR) based on existing and\\/or mature technologies (called “Early ARR”), aiming transuranics (TRU) burning and considering harmonization of TRU burning capability, technology readiness, economy and safety. The ARR is a 500MWe (1180MWth) oxide fueled sodium cooled fast reactor, which the low core height of 70cm and the large structure volume

Kazumi Ikeda; Kim O. Stein; Wataru Nakazato; Makoto Mito

2011-01-01

110

Monte Carlo simulation of core physics parameters of the Nigeria Research Reactor1 (NIRR-1)  

Microsoft Academic Search

The Monte Carlo N-Particle (MCNP) code, version 4C (MCNP4C) and a set of neutron cross-section data were used to develop an accurate three-dimensional computational model of the Nigeria Research Reactor-1 (NIRR-1). The geometry of the reactor core was modeled as closely as possible including the details of all the fuel elements, reactivity regulators, the control rod, all irradiation channels, and

S. A. Jonah; J. R. Liaw; J. E. Matos

2007-01-01

111

Operational performance of the three bean salad control algorithm on the ACRR (Annular Core Research Reactor)  

SciTech Connect

Experimental tests on the Annular Core Research Reactor have confirmed that the Three-Bean-Salad'' control algorithm based on the Pontryagin maximum principle can change the power of a nuclear reactor many decades with a very fast startup rate and minimal overshoot. The paper describes the results of simulations and operations up to 25 MW and 87 decades per minute. 3 refs., 4 figs., 1 tab.

Ball, R.M.; Madaras, J.J. (B and W Nuclear Technologies, Lynchburg, VA (USA). Space and Defense Systems); Trowbridge, F.R. Jr.; Talley, D.G.; Parma, E.J. Jr. (Sandia National Labs., Albuquerque, NM (USA))

1991-01-01

112

Comments on the feasibility of developing gas core nuclear reactors. [for manned interplanetary spacecraft propulsion  

NASA Technical Reports Server (NTRS)

Recent developments in the fields of gas core hydrodynamics, heat transfer, and neutronics indicate that gas core nuclear rockets may be feasible from the point of view of basic principles. Based on performance predictions using these results, mission analyses indicate that gas core nuclear rockets may have the potential for reducing the initial weight in orbit of manned interplanetary vehicles by a factor of 5 when compared to the best chemical rocket systems. In addition, there is a potential for reducing total trip times from 450 to 500 days for chemical systems to 250 to 300 days for gas core systems. The possibility of demonstrating the feasibility of gas core nuclear rocket engines by means of a logical series of experiments of increasing difficulty that ends with ground tests of full scale gas core reactors is considered.

Rom, F. E.

1969-01-01

113

Emergency core cooling system for a fast reactor. [LMFBR  

Microsoft Academic Search

The main heat transport system for a liquid-metal-cooled nuclear reactor is constructed with elevated piping and guard vessels or pipes around all components of the system below the elevation of the elevated piping so the head developed by the pumps at emergency motor speed will be unsufficient to lift the liquid-metal-coolant over the top of the guard tanks or pipes

H. G. Johnson; R. N. Madsen

1976-01-01

114

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

PubMed Central

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

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

2013-01-01

115

Nodal Diffusion Burnable Poison Treatment for Prismatic Reactor Cores  

SciTech Connect

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.

A. M. Ougouag; R. M. Ferrer

2010-10-01

116

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

SciTech Connect

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).

B. Boer; A. M. Ougouag

2010-09-01

117

Robust and accurate iris segmentation in very noisy iris images  

Microsoft Academic Search

Iris segmentation plays an important role in an accurate iri s recognition system. In less constrained environments where iris images are captured at-a-distance and on-the-move, iris segmentation becomes much more diffi cult due to the effects of significant variation of eye position and size, eyebr ows, eyelashes, glasses and contact lenses, and hair, together with illumination changes and varying focus

Peihua Li; Xiaomin Liu; Lijuan Xiao; Qi Song

2010-01-01

118

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

Microsoft Academic Search

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

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

2004-01-01

119

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

Microsoft Academic Search

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 blanket- seed

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

120

ORPHEE research reactor: 3D core depletion calculation using Monte-Carlo code TRIPOLI-4®  

NASA Astrophysics Data System (ADS)

ORPHEE is a research reactor located at CEA Saclay. It aims at producing neutron beams for experiments. This is a pool-type reactor (heavy water), and the core is cooled by light water. Its thermal power is 14 MW. ORPHEE core is 90 cm height and has a cross section of 27x27 cm2. It is loaded with eight fuel assemblies characterized by a various number of fuel plates. The fuel plate is composed of aluminium and High Enriched Uranium (HEU). It is a once through core with a fuel cycle length of approximately 100 Equivalent Full Power Days (EFPD) and with a maximum burnup of 40%. Various analyses under progress at CEA concern the determination of the core neutronic parameters during irradiation. Taking into consideration the geometrical complexity of the core and the quasi absence of thermal feedback for nominal operation, the 3D core depletion calculations are performed using the Monte-Carlo code TRIPOLI-4® [1,2,3]. A preliminary validation of the depletion calculation was performed on a 2D core configuration by comparison with the deterministic transport code APOLLO2 [4]. The analysis showed the reliability of TRIPOLI-4® to calculate a complex core configuration using a large number of depleting regions with a high level of confidence.

Damian, F.; Brun, E.

2014-06-01

121

A Novel Approach to Iris Localization for Iris Biometric Processing  

Microsoft Academic Search

Iris-based biometric system is gaining its importance in several applications. However, processing of iris biometric is a challenging and time consuming task. Detection of iris part in an eye image poses a number of challenges such as, inferior image quality, occlusion of eyelids and eyelashes etc. Due to these problems it is not possible to achieve 100% accuracy rate in

Somnath Dey; Debasis Samanta

2007-01-01

122

Pupil and Iris Localization for Iris Recognition in Mobile Phones  

Microsoft Academic Search

Until now, iris recognition has been used in many fields. Recently, there have been attempts to adopt iris recognition technology for the security of mobile phones. For example, in case of bank transaction service by using a mobile phone, using a mobile phone can use high level of security based on iris recognition. In this paper, we propose a new

Dal-ho Cho; Kang Ryoung Park; Dae Woong Rhee; Yanggon Kim; Jonghoon Yang

2006-01-01

123

The behavior of reactor power and flux resulting from changes in core-coolant temperature for a miniature neutron source reactor  

Microsoft Academic Search

In this work, measurements were performed to verify the theoretical predictions of reactor power and flux parameters that result from changes in core inlet temperature (Tin) and the temperature difference between the coolant inlet and outlet (?T) in the Nigeria Research Reactor-1 (NIRR-1), which is a Miniature Neutron Source Reactor (MNSR). The measured data shows that there is a strong

Y. A. Ahmed; G. I. Balogun; S. A. Jonah; I. I. Funtua

2008-01-01

124

Experimental Evaluation of Iris Recognition  

Microsoft Academic Search

Iris is an important biometric method with high reported accuracy. However, current iris recognition systems require substantial user cooperation in the image acquisition. Relatively little is known about how iris recognition might perform with less stringent control of image quality. We have re-implemented a Daugman-like iris matchingmethod, and evaluated its performance on an image dataset of over 12,000 images from

Xiaomei Liu; Kevin W. Bowyer; Patrick J. Flynn

2005-01-01

125

Full Core Reactor Analysis: Running Denovo on Jaguar  

SciTech Connect

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.

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

2013-01-01

126

Full Core Reactor Analysis: Running Denovo on Jaguar  

SciTech Connect

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.

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

2012-01-01

127

Full core reactor analysis: Running Denovo on Jaguar  

SciTech Connect

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)

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

128

Cosmic Ray Radiography of the Damaged Cores of the Fukushima Reactors  

NASA Astrophysics Data System (ADS)

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.

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

2012-10-01

129

Cosmic Ray Radiography of the Damaged Cores of the Fukushima Reactors  

E-print Network

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.

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

2012-09-13

130

Cosmic Ray Radiography of the Damaged Cores of the Fukushima Reactors  

E-print Network

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.

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

2012-01-01

131

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

PubMed

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

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

2012-10-12

132

Iris mammillations: Significance and associations  

Microsoft Academic Search

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

Nicola K Ragge; J Acheson; A Linn Murphree

1996-01-01

133

DCT-Based Iris Recognition  

Microsoft Academic Search

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

Donald M. Monro; Soumyadip Rakshit; Dexin Zhang

2007-01-01

134

An accurate iris location method for low quality iris images  

NASA Astrophysics Data System (ADS)

Iris location plays an important role in iris recognition system. Traditional iris location methods based on canny operator and integro-differential operator are affected by reflections, illumination inconsistency and eyelash. In this paper, we introduce an accurate iris location method for low quality iris images. First, a reflection removal method is used to interpolate the specular reflection. Then, we utilize Probable boundary (Pb) edge detection operator to detect papillary boundary with a lower interference point. Moreover, we optimize the Hough transform to obtain high accuracy result. Experimental results demonstrate that the location results of the proposed method are more accurate than other methods.

Wang, Ning; Li, Qiong; Abd El-Latif, Ahmed A.; Zhang, Tiejun; Peng, Jialiang

2012-04-01

135

Performance Potential of the Colloid Core Reactor Concept in Near-Earth Applications  

Microsoft Academic Search

An Air Force research program has produced performance estimates for the ; colloid core nuclear reactor rocket engine concept. These values are ; parametrically varied to determine their individual influence on an advanced ; nuclear upper stage to the Space Transportation System. Pessimistic and ; optimistic performance of the concept is estimated. The concept is compared with ; other propulsion

Capt. Thomas C. Meier

1973-01-01

136

Ex-Core CFD Analysis Results for the Prometheus Gas Reactor  

SciTech Connect

This paper presents the initial nozzle-to-nozzle (N2N) reactor vessel model scoping studies using computational fluid dynamics (CFD) analysis methods. The N2N model has been solved under a variety of different boundary conditions. This paper presents some of the basic hydraulic results from the N2N CFD analysis effort. It also demonstrates how designers were going to apply the analysis results to modify a number of the design features. The initial goals for developing a preliminary CFD N2N model were to establish baseline expectations for pressure drops and flow fields around the reactor core. Analysis results indicated that the averaged reactor vessel pressure drop for all analyzed cases was 46.9 kPa ({approx}6.8 psid). In addition, mass flow distributions to the three core fuel channel regions exhibited a nearly inverted profile to those specified for the in-core thermal/hydraulic design. During subsequent design iterations, the goal would have been to modify or add design features that would have minimized reactor vessel pressure drop and improved flow distribution to the inlet of the core.

Lorentz, Donald G. [Space Engineering, Bechtel Bettis, Inc. West Mifflin, PA 15122 (United States)

2007-01-30

137

Development of core fuel management code system for WWER-type reactors  

Microsoft Academic Search

In this article, a core fuel management program for hexagonal pressurized water type WWER reactors (CFMHEX) has been developed, which is based on advanced three-dimensional nodal method and integrated with thermal hydraulic code to realize the coupling of neutronics and thermal-hydraulics. In CFMHEX, all these feedback effects such as burnup, power distribution, moderator density, and control rod insertion are considered.

Bang-Yang XIA; Tao WANG; Zhong-Sheng XIE

2006-01-01

138

The evaluation of RCS depressurization to prevent core melting in pressure tube reactors (CANDU-type)  

Microsoft Academic Search

Pressure tube reactors, especially of the CANDU-type, have a low-pressure vessel calandria – under an internal pressure near atmospheric. The calandria vessel is immersed into the water contained inside a concrete structure – the calandria vault. In the case of accidents with the loss of normal core heat sinks, the moderator inside the calandria (heavy water) could become the ultimate

Stefan Mehedinteanu

2009-01-01

139

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

NASA Technical Reports Server (NTRS)

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.

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

2009-01-01

140

A Modification of the Inner and Outer Core for Reactor Pressure Vessel Lifetime Extension  

Microsoft Academic Search

The feasibility of nuclear power plant lifetime extension was examined by reducing the fast neutron fluence at the reactor pressure vessel (RPV) and relieving irradiation embrittlement of materials, and thus ensuring enough structural integrity beyond the design lifetime. Two fluence reduction options, peripheral assembly replacement and additional shield installation in the outer core structures, were applied to the Kori Unit-1

Bo Kyun Seo; Jong Kyung Kim; Chang Ho Shin; Tae Je Kwon

2001-01-01

141

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

SciTech Connect

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.

Christensen, L.J.

1989-01-01

142

Toward accurate and fast iris segmentation for iris biometrics.  

PubMed

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

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

2009-09-01

143

78 FR 56174 - In-Core Thermocouples at Different Elevations and Radial Positions in Reactor Core  

Federal Register 2010, 2011, 2012, 2013

...the adequacy of the AP1000 design and hydrogen control, the NRC regards this portion...reflooding an overheated core could generate hydrogen, at rates as high as 5.0 kg per second...the measured temperatures, the time evolution of the CET signal readings in the...

2013-09-12

144

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

SciTech Connect

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)

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

145

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

SciTech Connect

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.

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

146

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

SciTech Connect

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.

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

1984-01-01

147

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

SciTech Connect

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.

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

2009-04-01

148

Safety analysis for operating the Annular Core Research Reactor with Cintichem-type targets installed in the central region of the core  

SciTech Connect

Production of the molybdenum-99 isotope at the Annular Core Research Reactor requires highly enriched, uranium oxide loaded targets to be irradiated for several days in the high neutron-flux region of the core. This report presents the safety analysis for the irradiation of up to seven Cintichem-type targets in the central region of the core and compares the results to the Annular Core Research Reactor Safety Analysis Report. A 19 target grid configuration is presented that allows one to seven targets to be irradiated, with the remainder of the grid locations filled with aluminum ''void'' targets. Analyses of reactor, neutronic, thermal hydraulics, and heat transfer calculations are presented. Steady-state operation and accident scenarios are analyzed with the conclusion that the reactor can be operated safely with seven targets in the grid, and no additional risk to the public.

PARMA JR.,EDWARD J.

2000-01-01

149

Using crypts as iris minutiae  

NASA Astrophysics Data System (ADS)

Iris recognition is one of the most reliable biometric technologies for identity recognition and verification, but it has not been used in a forensic context because the representation and matching of iris features are not straightforward for traditional iris recognition techniques. In this paper we concentrate on the iris crypt as a visible feature used to represent the characteristics of irises in a similar way to fingerprint minutiae. The matching of crypts is based on their appearances and locations. The number of matching crypt pairs found between two irises can be used for identity verification and the convenience of manual inspection makes iris crypts a potential candidate for forensic applications.

Shen, Feng; Flynn, Patrick J.

2013-05-01

150

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

SciTech Connect

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

Boyd D. Christensen

2009-05-01

151

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

NASA Astrophysics Data System (ADS)

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.

Fallot, M.

2014-06-01

152

Congenital iris cysts  

Microsoft Academic Search

Unilateral, spontaneous, non-pigmented iris cysts appeared before the age of 2 years in four patients. Histopathological specimens obtained in three cases showed stratified to cuboidal, non-pigmented, epithelial lined cysts. Goblet cells were recognised in two of the three specimens. The clinical features and histopathological findings indicate that these cysts are derived from surface ectoderm and may be congenital.

R D Grutzmacher; T D Lindquist; M E Chittum; A H Bunt-Milam; R E Kalina

1987-01-01

153

Iris recognition technology  

Microsoft Academic Search

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

G. O. Williams

1996-01-01

154

Fisher Iris Data  

NSDL National Science Digital Library

Fisher (1936) explored 150 floral measurements in three species of blue flag irises in the Gaspe region of Quebec that were made by Anderson (1935). Anderson, E. 1935. The Irises of the Gaspe Peninsula. Bulletin of the American Iris Society 59: 2-5. Fisher, R. A. 1936. The Use of Multiple Measurements in Taxonomic Problems. Annals of Eugenics 7: 179-188.

Ethel Stanley (Beloit College;Biology)

2009-01-10

155

IRIS Resolving Unresolved Structure  

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...

156

NonOrthogonal View Iris Recognition System  

Microsoft Academic Search

This paper proposes a non-orthogonal view iris recognition system comprising a new iris imaging module, an iris segmentation module, an iris feature extraction module and a classification module. A dual-charge-coupled device camera was developed to capture four-spectral (red, green, blue, and near-infrared) iris images which contain useful information for simplifying the iris segmentation task. An intelligent random sample consensus iris

Chia-Te Chou; Sheng-Wen Shih; Wen-Shiung Chen; Victor W. Cheng; Duan-Yu Chen

2010-01-01

157

A Study on Iris Image Restoration  

Microsoft Academic Search

\\u000a Because iris recognition uses the unique patterns of the human iris, it is essential to acquire the iris images at high quality\\u000a for accurate recognition. Defocusing reduces the quality of the iris image and the performance of iris recognition, consequently.\\u000a In order to acquire a focused iris image at high quality, an iris recognition camera must control the focal length

Byung Jun Kang; Kang Ryoung Park

2005-01-01

158

IRIS Product Recommendations  

NASA Technical Reports Server (NTRS)

This report presents the Applied Meteorology Unit's (AMU) evaluation of SIGMET Inc.'s Integrated Radar Information System (IRIS) Product Generator and recommendations for products emphasizing lightning and microburst tools. The IRIS Product Generator processes radar reflectivity data from the Weather Surveillance Radar, model 74C (WSR-74C), located on Patrick Air Force Base. The IRIS System was upgraded from version 6.12 to version 7.05 in late December 1999. A statistical analysis of atmospheric temperature variability over the Cape Canaveral Air Force Station (CCAFS) Weather Station provided guidance for the configuration of radar products that provide information on the mixed-phase (liquid and ice) region of clouds, between 0 C and -20 C. Mixed-phase processes at these temperatures are physically linked to electrification and the genesis of severe weather within convectively generated clouds. Day-to-day variations in the atmospheric temperature profile are of sufficient magnitude to warrant periodic reconfiguration of radar products intended for the interpretation of lightning and microburst potential of convectively generated clouds. The AMU also examined the radar volume-scan strategy to determine the scales of vertical gaps within the altitude range of the 0 C to -20 C isotherms over the Kennedy Space Center (KSC)/CCAFS area. This report present's two objective strategies for designing volume scans and proposes a modified scan strategy that reduces the average vertical gap by 37% as a means for improving radar observations of cloud characteristics in the critical 0 C to -20 C layer. The AMU recommends a total of 18 products, including 11 products that require use of the IRIS programming language and the IRIS User Product Insert feature. Included is a cell trends product and display, modeled after the WSR-88D cell trends display in use by the National Weather Service.

Short, David A.

2000-01-01

159

Core damage severity evaluation for pressurized water reactors by artificial intelligence methods  

NASA Astrophysics Data System (ADS)

During the course of nuclear power evolution, accidents have occurred. However, in the western world, none of them had a severe impact on the public because of the design features of nuclear plants. In nuclear reactors, barriers constitute physical obstacles to uncontrolled fission product releases. These barriers are an important factor in safety analysis. During an accident, reactor safety systems become actuated to prevent the barriers from been breached. In addition, operators are required to take specified actions, meticulously depicted in emergency response procedures. In an accident, on-the-spot knowledge regarding the condition of the core is necessary. In order to make the right decisions toward mitigating the accident severity and its consequences, we need to know the status of the core [1, 3]. However, power plant instrumentation that can provide a direct indication of the status of the core during the time when core damage is a potential outcome, does not exist. Moreover, the information from instruments may have large uncertainty of various types. Thus, a very strong potential for misinterpreting incoming information exists. This research endeavor addresses the problem of evaluating the core damage severity of a Pressurized Water Reactor during a transient or an accident. An expert system has been constructed, that incorporates knowledge and reasoning of human experts. The expert system's inference engine receives incoming plant data that originate in the plethora of core-related instruments. Its knowledge base relies on several massive, multivariate fuzzy logic rule-sets, coupled with several artificial neural networks. These mathematical models have encoded information that defines possible core states, based on correlations of parameter values. The inference process classifies the core as intact, or as experiencing clad damage and/or core melting. If the system detects a form of core damage, a quantification procedure will provide a numerical assessment of the extent of the damage. The inference procedure is the Generalized Modus Ponens, which has its origin in the field of Approximate Reasoning. In addition, the use of neural networks enhances the accuracy of the quantification procedure. The model was tested for accuracy of assessment under severe accident conditions that compromised the reliability of instrumentation. The accuracy of the results established that the engagement of fuzzy logic in core state diagnosis constitutes a very promising method. Valid assessments were achieved in the vast majority of the test cases, in spite of troubling data deficiencies, which included inaccurate, distorted, or missing data.

Mironidis, Anastasios Pantelis

160

Advanced Core Design And Fuel Management For Pebble-Bed Reactors  

SciTech Connect

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.

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

2004-10-01

161

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

SciTech Connect

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.

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

162

Experimental and Analytic Study on the Core Bypass Flow in a Very High Temperature Reactor  

SciTech Connect

Core bypass flow has been one of key issues in the very high temperature reactor (VHTR) design for securing core thermal margins and achieving target temperatures at the core exit. The bypass flow in a prismatic VHTR core occurs through the control element holes and the radial and axial gaps between the graphite blocks for manufacturing and refueling tolerances. These gaps vary with the core life cycles because of the irradiation swelling/shrinkage characteristic of the graphite blocks such as fuel and reflector blocks, which are main components of a core's structure. Thus, the core bypass flow occurs in a complicated multidimensional way. The accurate prediction of this bypass flow and counter-measures to minimize it are thus of major importance in assuring core thermal margins and securing higher core efficiency. Even with this importance, there has not been much effort in quantifying and accurately modeling the effect of the core bypass flow. The main objectives of this project were to generate experimental data for validating the software to be used to calculate the bypass flow in a prismatic VHTR core, validate thermofluid analysis tools and their model improvements, and identify and assess measures for reducing the bypass flow. To achieve these objectives, tasks were defined to (1) design and construct experiments to generate validation data for software analysis tools, (2) determine the experimental conditions and define the measurement requirements and techniques, (3) generate and analyze the experimental data, (4) validate and improve the thermofluid analysis tools, and (5) identify measures to control the bypass flow and assess its performance in the experiment.

Richard Schultz

2012-04-01

163

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

SciTech Connect

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)

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

164

Fuel performance models for high-temperature gas-cooled reactor core design  

SciTech Connect

Mechanistic fuel performance models are used in high-temperature gas-cooled reactor core design and licensing to predict failure and fission product release. Fuel particles manufactured with defective or missing SiC, IPyC, or fuel dispersion in the buffer fail at a level of less than 5 x 10/sup -4/ fraction. These failed particles primarily release metallic fission products because the OPyC remains intact on 90% of the particles and retains gaseous isotopes. The predicted failure of particles using performance models appears to be conservative relative to operating reactor experience.

Stansfield, O.M.; Simon, W.A.; Baxter, A.M.

1983-09-01

165

Lunar in-core thermionic nuclear reactor power system conceptual design  

NASA Technical Reports Server (NTRS)

This paper presents a conceptual design of a lunar in-core thermionic reactor power system. The concept consists of a thermionic reactor located in a lunar excavation with surface mounted waste heat radiators. The system was integrated with a proposed lunar base concept representative of recent NASA Space Exploration Initiative studies. The reference mission is a permanently-inhabited lunar base requiring a 550 kWe, 7 year life central power station. Performance parameters and assumptions were based on the Thermionic Fuel Element (TFE) Verification Program. Five design cases were analyzed ranging from conservative to advanced. The cases were selected to provide sensitivity effects on the achievement of TFE program goals.

Mason, Lee S.; Schmitz, Paul C.; Gallup, Donald R.

1991-01-01

166

Lessons Learned from Sandia National Laboratories' Operational Readiness Review of the Annular Core Research Reactor (ACRR)  

SciTech Connect

The Sandia ACRR (a Hazard Category 2 Nuclear Reactor Facility) was defueled in June 1997 to modify the reactor core and control system to produce medical radioisotopes for the Department of Energy (DOE) Isotope Production Program. The DOE determined that an Operational Readiness Review (ORR) was required to confirm readiness to begin operations within the revised safety basis. This paper addresses the ORR Process, lessons learned from the Sandia and DOE ORRS of the ACRR, and the use of the ORR to confirm authorization basis implementation.

Bendure, Albert O.; Bryson, James W.

1999-05-17

167

Reactor physics analyses of the advanced neutron source three-element core  

SciTech Connect

A reactor physics analysis was performed for the Advanced Neutron Source reactor with a three-element core configuration. The analysis was performed with a two-dimensional r-z 20-energy-group finite-difference diffusion theory model of the 17-d fuel cycle. The model included equivalent r-z geometry representations of the central control rods, the irradiation and production targets, and reflector components. Calculated quantities include fuel cycle parameters, fuel element power distributions, unperturbed neutron fluxes in the reflector and target regions, reactivity perturbations, and neutron kinetics parameters.

Gehin, J.C.

1995-08-01

168

Neutronic analysis of three-element core configurations for the Advanced Neutron Source Reactor  

SciTech Connect

Calculations of several important neutronic parameters have been performed for ten different three-element configurations considered for the Advanced Neutron Source (ANS) Reactor. Six of these configurations (labeled ST, SB, MT, MB, LT, and LB) are there result of the permutations of the same three elements. Two configurations (ST- MOD and SB-MOD) have the same element configuration as their base core design (ST and SB) but have slightly different element dimensions, and two configurations (ST-OL1 and ST-OL2) have two overlapping elements to increase the neutron fluxes in the reflector. For each configuration, in addition to the conceptual two-element design, fuel-cycle calculations were performed with calculations required to obtain unperturbed fluxes. The element power densities, peak thermal neutron flux as a function of position throughout the cycle, fast flux, fast-to-thermal flux ratios, irradiation and production region fluxes, and control rod worth curves were determined. The effective multiplication factor for each fuel element criticality. A comparison shows that the ST core configurations have the best overall performance, and the fully overlapping core configuration ST-OL2 has the best performance by a large margin. Therefore, on the basis of the neutronics results, the fully overlapping configuration is recommended for further consideration in using a three-element ANS reactor core. Other considerations such as thermal-hydraulics, safety, and engineering that are not directly related to the core neutronic performance must be weighed before a final design is chosen.

Gehin, J.C.

1995-08-01

169

Effects of space-dependent cross sections on core physics parameters for compact fast spectrum space power reactors  

NASA Astrophysics Data System (ADS)

Effects of multigroup neutron cross section generation procedures on core physics parameters for compact fast spectrum reactors were examined. Homogeneous and space dependent multigroup cross section set were generated in 11 and 27 groups for a representative fast reactor core. These cross sections were used to compute various reactor physics parameters for the reference core. Coarse group structure and neglect of space dependence in the generation procedure resulted in inaccurate computations of reactor flux and power distributions and in significant errors regarding estimates of core reactivity and control system worth. Delayed neutron fraction was insensitive to cross section treatment, and computed reactivity coefficients were only slightly sensitive. However, neutron lifetime was found to be very sensitive to cross section treatment. Deficiencies in multigroup cross sections are reflected in core nuclear design and, consequently, in system mechanical design.

Lell, R. M.; Hanan, N. A.

170

Effects of Iris Surface Curvature on Iris Recognition  

SciTech Connect

To focus on objects at various distances, the lens of the eye must change shape to adjust its refractive power. This change in lens shape causes a change in the shape of the iris surface which can be measured by examining the curvature of the iris. This work isolates the variable of iris curvature in the recognition process and shows that differences in iris curvature degrade matching ability. To our knowledge, no other work has examined the effects of varying iris curvature on matching ability. To examine this degradation, we conduct a matching experiment across pairs of images with varying degrees of iris curvature differences. The results show a statistically signi cant degradation in matching ability. Finally, the real world impact of these ndings is discussed

Thompson, Joseph T [ORNL] [ORNL; Flynn, Patrick J [ORNL] [ORNL; Bowyer, Kevin W [University of Notre Dame, IN] [University of Notre Dame, IN; Santos-Villalobos, Hector J [ORNL] [ORNL

2013-01-01

171

An efficient iris segmentation approach  

NASA Astrophysics Data System (ADS)

Iris recognition system became a reliable system for authentication and verification tasks. It consists of five stages: image acquisition, iris segmentation, iris normalization, feature encoding, and feature matching. Iris segmentation stage is one of the most important stages. It plays an essential role to locate the iris efficiently and accurately. In this paper, we present a new approach for iris segmentation using image processing technique. This approach is composed of four main parts. (1) Eliminating reflections of light on the eye image based on inverting the color of the grayscale image, filling holes in the intensity image, and inverting the color of the intensity image to get the original grayscale image without any reflections. (2) Pupil boundary detection based on dividing an eye image to nine sub-images and finding the minimum value of the mean intensity for each sub-image to get a suitable threshold value of pupil. (3) Enhancing the contrast of outer iris boundary using exponential operator to have sharp variation. (4) Outer iris boundary localization based on applying a gray threshold and morphological operations on the rectangular part of an eye image including the pupil and the outer boundaries of iris to find the small radius of outer iris boundary from the center of pupil. The proposed approach has been tested on CASIA v1.0 iris image database and other collected iris image database. The experimental results show that the approach is able to detect pupil and outer iris boundary with high accuracy results approximately 100% and reduce time consuming.

Gomai, Abdu; El-Zaart, A.; Mathkour, H.

2011-10-01

172

Tutorial Notes Iris Recognition Tutorial @ BTAS 2013  

E-print Network

Tutorial Notes Iris Recognition Tutorial @ BTAS 2013 The notes for this tutorial are available online: www.cse.nd.edu/~kwb/Iris_Tutorial_2013.pdf Publications related to iris recognition: www.cse.nd.edu/~kwb/publications.htm September 29, 2013 #12;Iris Recognition in the Media Iris Recognition Tutorial @ BTAS 2013 September 29

Bowyer, Kevin W.

173

An Analysis of IrisCode  

Microsoft Academic Search

IrisCode is an iris recognition algorithm developed in 1993 and continuously improved by Daugman. It has been extensively applied in commercial iris recognition systems. IrisCode representing an iris based on coarse phase has a number of properties including rapid matching, binomial impostor distribution and a predictable false acceptance rate. Because of its successful applications and these properties, many similar coding

Adams Wai-Kin Kong; David Zhang; Mohamed S. Kamel

2010-01-01

174

Iris Recognition Algorithm Optimized for Hardware Implementation  

Microsoft Academic Search

Iris recognition is accepted as one of the most efficient biometric method. Implementing this method to the practical system requires the special image preprocessing where the iris feature extraction plays a crucial role. Recognition is preceeded by iris localization which consists in finding the iris boundaries as well as eyelids. In this paper the short introduction into iris localization and

Kamil Grabowski; Wojciech Sankowski; Malgorzata Napieralska; Mariusz Zubert; Andrzej Napieralski

2006-01-01

175

A new approach to iris pattern recognition  

Microsoft Academic Search

An iris identification algorithm is proposed based on adaptive thresholding. The iris images are processed fully in the spatial domain using the distinct features (patterns) of the iris. A simple adaptive thresholding method is used to segment these patterns from the rest of an iris image. This method could possibly be utilized for partial iris recognition since it relaxes the

Yingzi Du; Robert Ives; Delores M. Etter; Thad Welch

2004-01-01

176

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

E-print Network

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 ...

Yarsky, Peter

2005-01-01

177

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

NASA Technical Reports Server (NTRS)

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).

Kazeminezhad, F.; Anghai, S.

2008-01-01

178

LMFBR Prompt Burst Excursion (PBE) experiments in the Annular Core Pulse Reactor (ACPR)  

Microsoft Academic Search

A series of in-pile experiments has been initiated to provide information on the energetics of LMFBR cores subjected to hypothetical superprompt-critical excursions. In these experiments, a single fuel pin in an instrumented piston-loaded autoclave is pulse heated into partial vaporization by the ACPR with initial reactor periods as low as 1.4 msec. Nine experiments have been performed with fresh unrestructured

1976-01-01

179

Scale-model study of the seismic response of a nuclear reactor core  

SciTech Connect

The use of scale models to study the dynamics of a system of graphite core blocks used in certain nuclear reactor designs is described. Scaling laws, material selection, model instrumentation to measure collision forces, and the response of several models to simulated seismic excitation are covered. The effects of (1) Coulomb friction between the blocks, and (2) the clearance gaps between the blocks on the system response to seismic excitation are emphasized. 6 refs., 10 figs., 6 tabs.

Dove, R.C.; Dunwoody, W.E.; Rhorer, R.L.

1981-05-01

180

Fusion-power-core design of a compact Reversed-Field Pinch Reactor (CRFPR)  

SciTech Connect

A conceptual design of a fusion power core (FPC, i.e., plasma chamber, first wall, blanket, shield, coils) based on a Reversed-Field Pinch (RFP) has been completed. After a brief statement of rationale and description of the reactor configuration, the FPC integration is described in terms of power balance, thermal-hydraulics, and mechanical design. The engineering versatility, promise, and problems of this high-power-density approach to fusion are addressed.

Copenhaver, C.; Battak, M.E.; Cappiello, C.; Chaffee, A.D.; Davidson, J.W.; Hayenson, R.L.; Krakowski, R.A.; Lujan, R.E.; Mynard, R.C.; Schnurr, N.M.

1985-07-01

181

Fusion-power-core design of a compact Reversed-Field Pinch Reactor (CRFPR)  

SciTech Connect

A conceptual design of a fusion power core (FPC, i.e., plasma chamber, first wall, blanket, shield, coils) based on a Reversed-Field Pinch (RFP) has been completed. After a brief statement of rationale and description of the reactor configuration, the FPC integration is described in terms of power balance, thermal-hydraulics, and mechanical design. The engineering versatility, promise, and problems of this high-power-density approach to fusion are addressed.

Copenhaver, C.; Schnurr, N.M.; Krakowski, R.A.; Hagenson, R.L.; Mynard, R.C.; Cappiello, C.; Lujan, R.E.; Davidson, J.W.; Chaffee, A.D.; Battat, M.E.

1985-01-01

182

Integrated risk information system (IRIS)  

SciTech Connect

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).

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

1990-12-31

183

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

SciTech Connect

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.

Sterbentz, James W

2007-05-01

184

A Modification of the Inner and Outer Core for Reactor Pressure Vessel Lifetime Extension  

SciTech Connect

The feasibility of nuclear power plant lifetime extension was examined by reducing the fast neutron fluence at the reactor pressure vessel (RPV) and relieving irradiation embrittlement of materials, and thus ensuring enough structural integrity beyond the design lifetime. Two fluence reduction options, peripheral assembly replacement and additional shield installation in the outer core structures, were applied to the Kori Unit-1 reactor, and the fluence reduction effect was carefully analyzed. For an accurate estimate of the neutron fluence at the RPV and a reasonable description of the modified peripheral assemblies, a full-scope explicit modeling of a Monte Carlo simulation was employed in all calculations throughout this study. The Kori Unit-1 cycle-16 core was modeled on a three-dimensional representation by using the MCNP4B code, and the fluence distribution was estimated at the inner wall beltline around the circumferential weld of the RPV. On the basis of fracture toughness requirements of the RPV, the two modified cases were predicted to have an additional life of 7 to 10 effective full-power years. Throughout the core nuclear characteristics analyses, it was confirmed that the critical peaking factors for safe reactor operation were satisfied with the design limits.

Seo, Bo Kyun [Hanyang University (Korea, Republic of); Kim, Jong Kyung [Hanyang University (Korea, Republic of); Shin, Chang Ho [Hanyang University (Korea, Republic of); Kwon, Tae Je [Nuclear Fuel Company (Korea, Republic of)

2001-03-15

185

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

SciTech Connect

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.

Restrepo, L F

1992-08-01

186

Iris Recognition at a Distance  

Microsoft Academic Search

\\u000a We describe experiments demonstrating the feasibility of human iris recognition at up to 10 m distance between subject and\\u000a camera. The iris images of 250 subjects were captured with a telescope and infrared camera, while varying distance, capture\\u000a angle, environmental lighting, and eyewear. Automatic iris localization and registration algorithms, in conjunction with a\\u000a local correlation based matcher, were used to

Craig L. Fancourt; Luca Bogoni; Keith J. Hanna; Yanlin Guo; Richard P. Wildes; Naomi Takahashi; Uday Jain

2005-01-01

187

New Methods in Iris Recognition  

Microsoft Academic Search

Abstract—This paper presents the following four advances in iris recognition: 1) more disciplined methods for detecting and faithfully modeling the iris inner and outer boundaries with active contours, leading to more flexible embedded coordinate systems; 2) Fourier-based methods for solving problems in iris trigonome- try and projective geometry, allowing off-axis gaze to be handled by detecting it and “rotating” the

John Daugman

2007-01-01

188

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

SciTech Connect

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.

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

2003-08-04

189

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

SciTech Connect

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.

Hassan, Yassin

2013-10-22

190

Neutronic calculations for the conversion to LEU of a research reactor core  

SciTech Connect

For a five-year transitional period the Greek Research Reactor (GRR-1) was operating with a mixed core, containing both Low Enrichment (LEU) and High Enrichment (HEU) Uranium MTR- type fuel assemblies. The neutronic study of the GRR-1 conversion to LEU has been performed using a code system comprising the core-analysis code CITATION-LDI2 and the cell-calculation modules XSDRNPM and NITAWL-II of the SCALE code. A conceptual LEU core configuration was defined and analyzed with respect to the three dimensional multi-group neutron fluxes, the power distribution, the control-rod worth and the compliance with pre-defined Operation Limiting Conditions. Perturbation calculations and reactivity feedback computations were also carried out to provide input to a subsequent thermal-hydraulic study. (author)

Varvayanni, M.; Catsaros, N.; Stakakis, E. [National Center for Scientific Research 'DEMOKRITOS', 153 10 Aghia Paraskevi (Greece); Grigoriadis, D. [National Center for Scientific Research 'DEMOKRITOS', 153 10 Aghia Paraskevi (Greece); Department of Mechanical and Manufacturing Engineering, University of Cyprus, P.O. Box 20537, Nicosia 1678 (Cyprus)

2008-07-15

191

Non-Proliferative, Thorium-Based, Core and Fuel Cycle for Pressurized Water Reactors  

SciTech Connect

Two of the major barriers to the expansion of worldwide adoption of nuclear power are related to proliferation potential of the nuclear fuel cycle and issues associated with the final disposal of spent fuel. The Radkowsky Thorium Fuel (RTF) concept proposed by Professor A. Radkowsky offers a partial solution to these problems. The main idea of the concept is the utilization of the seed-blanket unit (SBU) fuel assembly geometry which is a direct replacement for a 'conventional' assembly in either a Russian pressurized water reactor (VVER-1000) or a Western pressurized water reactor (PWR). The seed-blanket fuel assembly consists of a fissile (U) zone, known as seed, and a fertile (Th) zone known as blanket. The separation of fissile and fertile allows separate fuel management schemes for the thorium part of the fuel (a subcritical 'blanket') and the 'driving' part of the core (a supercritical 'seed'). The design objective for the blanket is an efficient generation and in-situ fissioning of the U233 isotope, while the design objective for the seed is to supply neutrons to the blanket in a most economic way, i.e. with minimal investment of natural uranium. The introduction of thorium as a fertile component in the nuclear fuel cycle significantly reduces the quantity of plutonium production and modifies its isotopic composition, reducing the overall proliferation potential of the fuel cycle. Thorium based spent fuel also contains fewer higher actinides, hence reducing the long-term radioactivity of the spent fuel. The analyses show that the RTF core can satisfy the requirements of fuel cycle length, and the safety margins of conventional pressurized water reactors. The coefficients of reactivity are comparable to currently operating VVER's/PWR's. The major feature of the RTF cycle is related to the total amount of spent fuel discharged for each cycle from the reactor core. The fuel management scheme adopted for RTF core designs allows a significant decrease in the amount of discharged spent fuel, for a given energy production, compared with standard VVER/PWR. The total Pu production rate of RTF cycles is only 30 % of standard reactor. In addition, the isotopic compositions of the RTF's and standard reactor grade Pu are markedly different due to the very high burnup accumulated by the RTF spent fuel.

Todosow M.; Todosow M.; Raitses, G. (BNL) Galperin, A. (Ben Gurion University)

2009-07-12

192

Iris imaging system with adaptive optical elements  

NASA Astrophysics Data System (ADS)

Iris recognition utilizes distinct patterns found in the human iris to perform identification. Image acquisition is a critical first step toward successful operation of iris recognition systems. However, the quality of iris images required by standard iris recognition algorithms puts stringent constraints on the imaging systems, which results in a constrained capture volume. We have incorporated adaptive optical elements to expand the capture volume of a 3-m stand-off iris recognition system.

Choi, Junoh; Dixon, Kevin R.; Wick, David V.; Bagwell, Brett E.; Soehnel, Grant H.; Clark, Brian

2012-01-01

193

Proliferation resistance potential and burnup characteristics of an equilibrium core of novel natural uranium fueled nuclear research reactor  

Microsoft Academic Search

Standard reactor simulation codes WIMS-D\\/4 and CITATION were employed to analyze the proliferation resistance potential and burnup characteristics of a novel natural uranium fueled nuclear research reactor [Annals of Nuclear Energy 31(12), 1331–1356]. It was found that the proposed core, which provides twice the flux per unit core power compared to similar natural uranium fueled, light water cooled, heavy water

Mohammad Javed Khan; Aslam; Nasir Ahmad

2005-01-01

194

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

SciTech Connect

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.

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

2009-01-01

195

Comparison of iris recognition algorithms  

Microsoft Academic Search

In this paper, we have studied various well known algorithms for iris recognition. Four algorithms due to Sanchez-Avila et al. (2001), Li Ma et al. (2002), Tisse et al. (2002) and Daugman (2001) are implemented and compared on the CASIA iris image database. The results show that the Daugman's algorithm gave the highest accuracy of 99.9%.

Mayank Vatsa; Richa Singh; P. Gupta

2004-01-01

196

Non-Orthogonal Iris Segmentation.  

National Technical Information Service (NTIS)

The goal of this Trident Scholar project was to isolate the iris, the colored part of the eye, in a non-orthogonal, digital image of the human eye. A non-orthogonal image is an image where the eye is not looking directly at the camera. Iris pattern differ...

B. L. Bonney

2005-01-01

197

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

SciTech Connect

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.

B. Boer; A. M. Ougouag

2010-05-01

198

A 100 MWe advanced sodium-cooled fast reactor core concept  

SciTech Connect

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)

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

2012-07-01

199

Multimegawatt nuclear electric propulsion with gaseous and vapor core reactors with MHD  

NASA Astrophysics Data System (ADS)

This study investigated the development of a system concept for space power generation and nuclear electric propulsion based on a fissioning plasma core reactor (FPCR) with magnetohydrodynamic (MHD) power conversion system, coupled to a magnetoplasmadynamic (MPD) thruster. The FPCR is a liquid-vapor core reactor concept operating with metallic uranium or uranium tetrafluoride (UF4) vapor as the fissioning fuel and alkali metals or their fluorides as working fluid in a closed Rankine cycle with MHD energy conversion. Candidate working fluids include K, Li, Na, KF, LiF, NaF, etc. The system features core outlet temperatures of 3000 to 4000 K at pressures of about 1 to 10 MPa, MHD temperatures of 2000 to 3000 K, and radiator temperatures of 1200 to 2000 K. This combination of parameters offers the potential for low total system specific mass in the range of 0.4 to 0.6 kg/kWe. The MHD output could be coupled with minimal power conditioning to the variable specific impulse magnetoplasma rocket (VASIMR), MPD thrusters or other types of thruster for producing thrust at very high specific impulse (Isp=1500 to 10,000 s). .

Knight, Travis; Anghaie, Samim; Smith, Blair; Houts, Michael

2001-02-01

200

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

SciTech Connect

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.

Vondy, D.R.

1984-07-01

201

A survey of alternative once-through fast reactor core designs  

SciTech Connect

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)

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

202

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

SciTech Connect

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)

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

2013-07-01

203

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

NASA Technical Reports Server (NTRS)

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.

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

1981-01-01

204

Evaluation of surface deposits on the channel wall of trepanned reactor core graphite samples  

NASA Astrophysics Data System (ADS)

Samples have been trepanned from the fuel and interstitial channel walls of PGA graphite reactor cores of two Magnox gas cooled power stations after a period of service. These samples have been considered explicitly for the presence of deposits on the channel facing surfaces. A combination of focused ion beam milling and imaging has been used to determine the presence of such deposits and where present to make measurements of the thickness. These thicknesses vary from a few nanometres to tens of micrometres. In addition, both the chemical composition and chemical state have been investigated using energy dispersive X-ray microanalysis in a scanning electron microscope and Raman spectroscopy respectively.

Heard, P. J.; Payne, L.; Wootton, M. R.; Flewitt, P. E. J.

2014-02-01

205

BWR In-Core Monitor Housing Replacement Under Dry Condition of Reactor Pressure Vessel  

SciTech Connect

A new method of In-Core Monitor Housing replacement has been successfully applied to Tokai Unit 2 (BWR with 1100 MWe) in April of 2001. It was designed to replace a housing under dry condition of reactor pressure vessel (RPV): this enabled the elimination of water filled-up and drained processes during the replacement procedure resulting in the reduction of implementation schedule. To realize the dry condition, the radiation shields were placed in the RPV and the hollow guide pipe (GP) was adopted to transfer the apparatuses from the top to the bottom work area. (authors)

Tatsuo Ishida; Shoji Yamamoto; Fujitoshi Eguchi [Japan Atomic Power Company (Japan); Motomasa Fuse; Kouichi Kurosawa; Sadato Shimizu; Minoru Masuda [Hitachi Ltd. (Japan); Shinya Fujii; Junji Tanaka [General Electric International Inc. (Japan); Jacobson, Bryce A. [General Electric Company (United States)

2002-07-01

206

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

SciTech Connect

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.

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

1981-06-01

207

IAEA coordinated research projects on core physics benchmarks for high temperature gas-cooled reactors  

SciTech Connect

High-temperature Gas-Cooled Reactor (HTGR) designs present special computational challenges related to their core physics characteristics, in particular neutron streaming, double heterogeneities, impurities and the random distribution of coated fuel particles in the graphite matrix. In recent years, two consecutive IAEA Coordinated Research Projects (CRP 1 and CRP 5) have focused on code-to-code and code-to-experiment comparisons of representative benchmarks run by several participating international institutes. While the PROTEUS critical HTR experiments provided the test data reference for CRP-1, the more recent CRP-5 data has been made available by the HTTR, HTR-10 and ASTRA test facilities. Other benchmark cases are being considered for the GT-MHR and PBMR core designs. This paper overviews the scope and some sample results of both coordinated research projects. (authors)

Methnani, M. [International Atomic Energy Agency IAEA, Wagramerstrasse 5, A-1400 Vienna (Austria)

2006-07-01

208

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

SciTech Connect

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.

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

1995-06-01

209

Development and Validation of ARKAS cellule: An Advanced Core-Bowing Analysis Code for Fast Reactors  

SciTech Connect

An advanced analysis code, ARKAS cellule, has been developed to determine the core distortion and the mechanical behavior of fast reactors. In this code, each hexagonal subassembly duct is represented by a folded thin plate structure divided into a user-specified number of shell elements so that the interduct contact forms and the cross-sectional distortion effect of each duct are properly taken into account. In this paper, the numerical model of the ARKAS cellule code is introduced, and the analytical results for two validation problems are presented. From a single duct compaction analysis, the first validation problem, it is clarified that the new analytical model is applicable to simulating the change of duct compaction stiffness that depends on the loading conditions such as the loading pad forms and the number of contact faces. The second validation analysis has been conducted by comparison with the experimental values obtained by the National Nuclear Corporation Limited in the United Kingdom using the core restraint uniplanar experimental rig (CRUPER), an ex-reactor rig in which a cluster of 91 short ducts is compressed by 30 movable peripheral rams toward the center of the cluster in seven stages. The analysis clarified that the predictions obtained using ARKAS cellule agree well with the measured ram loads and interwrapper gap widths during the compaction sequence. One may conclude that ARKAS cellule is valid for quantitative analysis of the core mechanical behavior and will be particularly useful for the evaluation of transient deformation of core assemblies during accidents in which the distortion of loading pads have important effects on obtaining favorable reactivity feedback.

Ohta, Hirokazu [Central Research Institute of Electric Power Industry (Japan); Yokoo, Takeshi [Central Research Institute of Electric Power Industry (Japan); Nakagawa, Masatoshi [AITEL Corporation (Japan); Matsuyama, Shinichiro [Toshiba Corporation (Japan)

2004-05-15

210

The scalability of OTR (out-of-core thermionic reactor) space nuclear power systems  

SciTech Connect

In this document, masses of the STAR-C power system and an optimized out-of-core thermionic reactor (OTR) power system versus power level are investigated. The impacts of key system parameters on system performance are also addressed. The STAR-C is mass competitive below about 15 kWe, but at higher power levels the scalability is relatively poor. An optimized OR is the least massive space nuclear power system below 25 kWe, and scales well to 50 kWe. The system parameters that have a significant impact on the scalability of the STAR-C are core thermal flux, thermionic converter efficiency, and core length to diameter ratio. The emissivity of the core surface is shown to be a relatively unimportant parameter. For an optimized OR power system, the most significant system parameter is the maximum allowable fuel temperature. It is also shown that if advanced radiation-hardened electronics are used in the satellite payload, a very large mass savings is realized. 10 refs., 23 figs., 7 tabs.

Gallup, D.R.

1990-03-01

211

Iris Recognition Based on Multichannel Gabor Filtering  

Microsoft Academic Search

A new approach for personal identification based on iris recognition is presented in this paper. The body of this paper details the steps of iris recognition, including image preprocessing, feature extraction and classifier design. The proposed algorithm uses a bank of Gabor filters to capture both local and global iris characteristics to form a fixed length feature vector. Iris matching

Li Ma; Yunhong Wang; Tieniu Tan

212

Eyelash Removal Method for Human Iris Recognition  

Microsoft Academic Search

A novel eyelash removal method for preprocessing of human iris images in a human iris recognition system is presented. The method filters each occluded pixel along an axis perpendicular to the eyelash direction, and accepts the filtered value if it changes by more than a certain threshold. This allows partially occluded regions of the iris to be included in iris

Dexin Zhang; Donald M. Monro; Soumyadip Rakshit

2006-01-01

213

Iris Recognition Based on Multichannel Gabor Filter  

Microsoft Academic Search

Abstract Anew approach for personal identification based on iris recognition is presented in this paper. The body of this paper details the steps of iris recognition, including image preprocessing, feature extraction and classifier design. The proposed algorithm uses a bank ,of Gabor filters to capture ,both ,local and ,global iris characteristics to form a fixed length feature vector. Iris matching

L. Ma; Y. Wang; T. Tan

2002-01-01

214

Iris recognition using independent component analysis  

Microsoft Academic Search

This paper develops a new method for iris recognition based on independent component analysis. The iris recognition consisted of three major components: image preprocessing, feature extraction and classification. A three-step multiscale approach was employed in image preprocessing to realize iris localization, normalization and enhancement. In iris feature extraction, an efficient approach called independent component analysis was used which was statistically

Yong Wang; Jiu-Qiang Han

2005-01-01

215

Personal Identification Based on Iris Texture Analysis  

Microsoft Academic Search

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

Li Ma; Tieniu Tan; Yunhong Wang; Dexin Zhang

2003-01-01

216

Improving iris recognition accuracy via cascaded classifiers  

Microsoft Academic Search

As a reliable approach to human identification, iris recog- nition has received increasing attention in recent years. The most distinguishing feature of an iris image comes from the fine spatial changes of the image structure. So iris pattern representation must characterize the local intensity variations in iris signals. However, the measurements from minutiae are easily affected by noise, such as

Zhenan Sun; Yunhong Wang; Tieniu Tan; Jiali Cui

2005-01-01

217

International Referecne Ionosphere (IRI) - 2006  

NASA Astrophysics Data System (ADS)

With this presentation the newest version of IRI IRI-2006 will be officially released The new version includes a number of critical improvements and long sought-over additions For the electron density in the topside two new options were added a correction term based on Alouette ISIS topside sounder data and the new NeQuick model In the D-region new models are included for high-latitudes based on rocket and incoherent scatter data The occurrence probability of spread-F is added as a new parameter to IRI although currently only in the form of a regional model for the South-American sector IRI-2006 also includes new models for the description of topside ion composition and equatorial disturbance ion drift In addition the newest version of the International Geomagnetic Reference Field IGRF Version 10 is now implemented in IRI and used for all internal magnetic coordinate computations This presentation will describe and discuss the newest IRI version in great detail and give examples on how they improvements and new additions will benefit specific applications of the IRI model

Bilitza, D.

218

Passive safety features of low sodium void worth metal fueled cores in a bottom supported reactor vessel  

SciTech Connect

A study has been performed on the passive safety features of low-sodium-void-worth metallic-fueled reactors with emphasis on using a bottom-supported reactor vessel design. The reactor core designs included self-sufficient types as well as actinide burners. The analyses covered the reactor response to the unprotected, i.e. unscrammed, transient overpower accident and the loss-of-flow accident. Results are given demonstrating the safety margins that were attained. 4 refs., 4 figs., 2 tabs.

Chang, Y.I.; Marchaterre, J.F.; Wade, D.C.; Wigeland, R.A. (Argonne National Lab., IL (United States)); Kumaoka, Yoshio; Suzuki, Masao; Endo, Hiroshi (Toshiba Corp., Tokyo (Japan)); Nakagawa, Hiroshi (Japan Atomic Power Co., Tokyo (Japan))

1991-01-01

219

Recriticality in a BWR (boiling water reactor) following a core damage event  

SciTech Connect

This report describes the results of a study conducted by Pacific Northwest Laboratory to assist the US Nuclear Regulatory Commission in evaluating the potential for recriticality in boiling water reactors (BWRs) during certain low probability severe accidents. Based on a conservative bounding analysis, this report concludes that there is a potential for recriticality in BWRs if core reflood occurs after control blade melting has begun but prior to significant fuel rod melting. However, a recriticality event will most likely not generate a pressure pulse significant enough to fail the vessel. Instead, a quasi-steady power level would result and the containment pressure and temperature would increase until the containment failure pressure is reached, unless actions are taken to terminate the event. Two strategies are identified that would aid in regaining control of the reactor and terminate the recriticality event before containment failure pressures are reached. The first strategy involves initiating boration injection at or before the time of core reflood if the potential for control blade melting exists. The second strategy involves initiating residual heat removal suppression pool cooling to remove the heat load generated by the recriticality event and thus extend the time available for boration. 31 figs., 17 tabs.

Scott, W.B.; Harrison, D.G.; Libby, R.A.; Tokarz, R.D. (Pacific Northwest Lab., Richland, WA (USA)); Wooton, R.D.; Denning, R.S.; Tayloe, R.W. Jr. (Battelle Memorial Inst., Columbus, OH (USA))

1990-12-01

220

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

SciTech Connect

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).

David W. Nigg

2013-09-01

221

Bartus Iris biometrics  

SciTech Connect

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.

Johnston, R.; Grace, W.

1996-07-01

222

Shape Adaptive, Robust Iris Feature Extraction from Noisy Iris Images  

PubMed Central

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

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

2013-01-01

223

PERFLUOROOCTANE SULFONATE (PFOS) - IRIS ASSESSMENT  

EPA Science Inventory

An assessment of perfluorooctane sulfonate (PFOS) is underway that will establish the RfD/RfC as appropriate which will be made available to the public through the Integrated Risk Information System (IRIS)....

224

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...

225

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

NASA Astrophysics Data System (ADS)

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 the NNR fluid mechanics and heat transfer module STAR-CD for light water reactor applications. The coupling has been accomplished via an interface program, which is responsible for mapping the STAR-CD and DeCART meshes, managing communication, and monitoring convergence. DeCART obtains the solution of the 3-D Boltzmann transport equation by performing a series of 2-D modular ray tracing-based method of characteristics problems that are coupled within the framework of 3-D coarse-mesh finite difference. The relatively complex geometry and increased axial heterogeneity found in BWRs are beyond the modeling capability of the original version of DeCART. In this work, DeCART is extended in three primary areas. First, the geometric capability is generalized by extending the modular ray tracing scheme and permitting an unstructured mesh in the global finite difference kernel. Second, numerical instabilities, which arose as a result of the severe axial heterogeneity found in BWR cores, have been resolved. Third, an advanced nodal method has been implemented to improve the accuracy of the axial flux distribution. In this semi-analytic nodal method, the analytic solution to the transverse-integrated neutron diffusion equation is obtained, where the nonhomogeneous neutron source was first approximated by a quartic polynomial. The successful completion of these three tasks has allowed the application of the coupled DeCART/STAR-CD code to practical BWR problems.

Thomas, Justin W.

226

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

NASA Technical Reports Server (NTRS)

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.

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

1973-01-01

227

Measurements of Reaction Rates in Zone-Type Cores of Fast Critical Assembly Simulating High Conversion Light Water Reactor  

Microsoft Academic Search

Measurements of reaction rates have been performed in three uranium-fueled zone-type cores of the FCA constructed for a series of experiments on a high conversion light water reactor (HCLWR). These cores possess central test zones of different fuel enrichments and moderator to fuel volume ratios. Radial and axial fission rates of 236U, 239Pu, 238U and 23,Np were measured in each

Makoto ?BU; Tatsuo NEMOTO; Susumu IIJIMA; Takeshi SAKURAI; Yoshihisa TAHARA

1989-01-01

228

Core loading pattern optimization of thorium fueled heavy water breeder reactor using genetic algorithm  

SciTech Connect

In this work genetic algorithm was proposed to solve fuel loading pattern optimization problem in thorium fueled heavy water reactor. The objective function of optimization was to maximize the conversion ratio and minimize power peaking factor. Those objectives were simultaneously optimized using non-dominated Pareto-based population ranking optimal method. Members of non-dominated population were assigned selection probabilities based on their rankings in a manner similar to Baker's single criterion ranking selection procedure. A selected non-dominated member was bred through simple mutation or one-point crossover process to produce a new member. The genetic algorithm program was developed in FORTRAN 90 while neutronic calculation and analysis was done by COREBN code, a module of core burn-up calculation for SRAC. (authors)

Soewono, C. N.; Takaki, N. [Dept. of Applied Science Engineering, Faculty Tokai Univ., Kanagawa-ken, Hiratsuka-shi Kitakaname 4-1-1 (Japan)

2012-07-01

229

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

NASA Technical Reports Server (NTRS)

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.

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

1973-01-01

230

Properties of in-core reactor-irradiated amorphous Fe 40Ni 40B 20  

NASA Astrophysics Data System (ADS)

Amorphous Fe 40Ni 40B 20 has been exposed to in-core reactor irradiation with fluences of thermal neutrons up to 6.5 × 10 19 n cm -2. The nuclear reaction 10B (n, ?) 7Li + 2.3 MeV, caused by the capture of thermal neutrons, leads to damage levels of up to 26 dpa. Subsequent investigations of the irradiated material by transmission electron microscopy showed that the material remains amorphous under neutron irradiation. Irradiation-induced defects could not be detected. Thermal analysis of these specimens revealed that after irradiation to 6.5 × 10 19 n cm -2 the amorphous structure appeared to be less stable with respect to the as-quenched material. Bending tests yielded a drastic loss in ductility already after 1 × 10 19 n cm -2.

Gerling, R.; Wagner, R.

1982-06-01

231

Thermal-hydraulic calculations for the conversion to LEU of a research reactor core  

SciTech Connect

The thermal-hydraulic analysis performed for the needs of the conversion of the open pool 5MW Greek Research Reactor (GRR-1) to a pure Low Enrichment (LEU) configuration is presented. The methodology was based on a complete set of neutronic calculations performed for the new core configuration, in compliance with pre-defined Operation Limiting Conditions. The hottest channel analysis approach was adopted, and peaking factors were used to account for fabrication or measuring uncertainties. Calculations were carried out using the numerical codes NATCON, PLTEMP and PARET provided by Argonne National Laboratory (ANL). Two main different classes of conditions were considered, namely i) steady state normal operating conditions and ii) transient cases related to accidental events including reactivity feedback effects. For steady state operating conditions the behaviour of the new configuration was examined both for forced and natural convection cooling modes. Transient calculations considered several initiating events including reactivity insertion accidents (slow or fast reactivity insertion) and total or partial loss-of-flow accidents, i.e. in accordance to guidelines provided by the IAEA for research Reactors. (author)

Grigoriadis, D. [National Center for Scientific Research 'DEMOKRITOS', 153 10 Aghia Paraskevi (Greece); Department of Mechanical and Manufacturing Engineering, University of Cyprus, P.O. Box 20537, Nicosia 1678 (Cyprus); Varvayanni, M.; Catsaros, N.; Stakakis, E. [National Center for Scientific Research 'DEMOKRITOS', 153 10 Aghia Paraskevi (Greece)

2008-07-15

232

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

SciTech Connect

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.

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

233

Compact Reversed-Field Pinch Reactors (CRFPR): fusion-power-core integration study  

SciTech Connect

Using detailed two-dimensional neutronics studies based on the results of a previous framework study (LA-10200-MS), the fusion-power-core (FPC) integration, maintenance, and radio-activity/afterheat control are examined for the Compact Reversed-Field Pinch Reactor (CRFPR). While maintaining as a base case the nominal 20-MW/m/sup 2/ neutron first-wall loading design, CRFPR(20), the cost and technology impact of lower-wall-loading designs are also examined. The additional detail developed as part of this follow-on study also allows the cost estimates to be refined. The cost impact of multiplexing lower-wall-loading FPCs into a approx. 1000-MWe(net) plant is also examined. The CRFPR(20) design remains based on a PbLi-cooled FPC with pressurized-water used as a coolant for first-wall, pumped-limiter, and structural-shield systems. Single-piece FPC maintenance of this steady-state power plant is envisaged and evaluated on the basis of a preliminary layout of the reactor building. This follow-on study also develops the groundwork for assessing the feasibility and impact of impurity/ash control by magnetic divertors as an alternative to previously considered pumped-limiter systems. Lastly, directions for future, more-detailed power-plant designs based on the Reversed-Field Pinch are suggested.

Copenhaver, C.; Krakowski, R.A.; Schnurr, N.M.; Miller, R.L.; Bathke, C.G.; Hagenson, R.L.; Mynard, C.R.; Chaffee, A.D.; Cappiello, C.; Davidson, J.W.

1985-08-01

234

Natural Fueling of the Core and Edge in a Tokamak Fusion Reactor  

NASA Astrophysics Data System (ADS)

A natural fueling mechanismootnotetextW. Wan, S. E. Parker, Y. Chen and F. W. Perkins, Phys. Plasmas 17, 040701 (2010). that helps to maintain the main core deuterium and tritium (DT) density profiles in a tokamak fusion reactor is presented. In H-mode plasmas dominated by ion-temperature gradient (ITG) driven turbulence, cold DT ions near the edge will naturally pinch radially inward towards the core. This mechanism is due to the quasi-neutral heat flux dominated nature of ITG turbulence and still applies when trapped and passing kinetic electron effects are included. Fueling using shallow pellet injection or supersonic gas jets is augmented by an inward pinch of could DT fuel. The natural fueling mechanism is investigated using the gyrokinetic turbulence code GEM and is analyzed using quasilinear theory. Profiles similar to those used for conservative ITER transport modeling that have a completely flat density profile are examined and it is found that natural fueling actually reduces the linear growth rates and energy transport. Additionally, it is shown that the Helium ash diffuses radially outward as the cold fuel moves radially inward. The natural fueling effect may also apply to the edge pedestal density buildup. Recent DEGAS 2 calculations indicate the neutrals in the pedestal are colder than the background ions.ootnotetextD. Stotler, International Transport Task Force Meeting, Annapolis, MD (2010). We intend to do further work to determine what cold fuel profiles are needed to fuel the pedestal and if they are consistent with edge neutral source models. Natural fueling (either in the core or edge) requires a two component (hot bulk and cold fuel) plasma and charge exchange collisions tend to equilibrate the ion and neutral source temperature reducing the effect. We will further investigate the relevant collisional time scales and further demonstrate the viability of the natural fueling mechanism for ITER parameters.

Wan, Weigang

2010-11-01

235

On Techniques for Angle Compensation in Nonideal Iris Recognition  

Microsoft Academic Search

The popularity of the iris biometric has grown considerably over the past two to three years. Most research has been focused on the development of new iris processing and recognition algorithms for frontal view iris images. However, a few challenging directions in iris research have been identified, including processing of a nonideal iris and iris at a distance. In this

Stephanie A. C. Schuckers; Natalia A. Schmid; Aditya Abhyankar; Vivekanand Dorairaj; Christopher K. Boyce; Lawrence A. Hornak

2007-01-01

236

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

Microsoft Academic Search

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

T. Hoshino; T. Shirakawa

1982-01-01

237

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

Microsoft Academic Search

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

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

2005-01-01

238

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

E-print Network

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 ...

Helvenston, Edward M. (Edward March)

2006-01-01

239

Benign Adenoma of the Iris Pigment Epithelium: Clinical and Iris Fluorescein Angiographic Features  

Microsoft Academic Search

Benign adenoma of iris pigment epithelium is a rare neoformation characterized by a multinodular, dark brown to dark black, relatively stationary lesion, especially localized in the peripheral iris. We report a case of iris pigment epithelium adenoma with no evidence of change during the 6-year follow-up. Iris fluorescein angiography showed a mild staining of the lesion without evidence of abnormal

V. Isola; M. Battaglia Parodi; S. Calderini

1994-01-01

240

High accuracy modeling for advanced nuclear reactor core designs using Monte Carlo based coupled calculations  

NASA Astrophysics Data System (ADS)

The main objective of this PhD research is to develop a high accuracy modeling tool using a Monte Carlo based coupled system. The presented research comprises the development of models to include the thermal-hydraulic feedback to the Monte Carlo method and speed-up mechanisms to accelerate the Monte Carlo criticality calculation. Presently, deterministic codes based on the diffusion approximation of the Boltzmann transport equation, coupled with channel-based (or sub-channel based) thermal-hydraulic codes, carry out the three-dimensional (3-D) reactor core calculations of the Light Water Reactors (LWRs). These deterministic codes utilize nuclear homogenized data (normally over large spatial zones, consisting of fuel assembly or parts of fuel assembly, and in the best case, over small spatial zones, consisting of pin cell), which is functionalized in terms of thermal-hydraulic feedback parameters (in the form of off-line pre-generated cross-section libraries). High accuracy modeling is required for advanced nuclear reactor core designs that present increased geometry complexity and material heterogeneity. Such high-fidelity methods take advantage of the recent progress in computation technology and coupled neutron transport solutions with thermal-hydraulic feedback models on pin or even on sub-pin level (in terms of spatial scale). The continuous energy Monte Carlo method is well suited for solving such core environments with the detailed representation of the complicated 3-D problem. The major advantages of the Monte Carlo method over the deterministic methods are the continuous energy treatment and the exact 3-D geometry modeling. However, the Monte Carlo method involves vast computational time. The interest in Monte Carlo methods has increased thanks to the improvements of the capabilities of high performance computers. Coupled Monte-Carlo calculations can serve as reference solutions for verifying high-fidelity coupled deterministic neutron transport methods with detailed and accurate thermal-hydraulic models. The development of such reference high-fidelity coupled multi-physics scheme is described in this dissertation on the basis of MCNP5, NEM, NJOY and COBRA-TF (CTF) computer codes. This work presents results from studies performed and implemented at the Pennsylvania State University (PSU) on both accelerating Monte Carlo criticality calculations by using hybrid nodal diffusion Monte Carlo schemes and thermal-hydraulic feedback modeling in Monte Carlo core calculations. The hybrid MCNP5/CTF/NEM/NJOY coupled code system is proposed and developed in this dissertation work. The hybrid coupled code system contains a special interface developed to update the required MCNP5 input changes to account for dimension and density changes provided by the thermal-hydraulics feedback module. The interface has also been developed to extract the flux and reaction rates calculated by MCNP5 to later transform the data into the power feedback needed by CTF (axial and radial peaking factors). The interface is contained in a master program that controls the flow of the calculations. Both feedback modules (thermal-hydraulic and power subroutines) use a common internal interface to further accelerate the data exchange. One of the most important steps to correctly include the thermal hydraulic feedback into MCNP5 calculations begins with temperature dependent cross section libraries. If the cross sections used for the calculations are not at the correct temperature, the temperature feedback cannot be included into MCNP5 (referred to the effect of temperature on cross sections: Doppler boarding of resolve and unresolved resonances, thermal scattering and elastic scattering). The only method of considering the temperature effects on cross sections is through the generation (or as introduced in this dissertation through a novel interpolation mechanism) of continuous energy temperature-dependent cross section libraries. An automated methodology for generation of continuous energy temperature-dependent cross section libraries has been developed

Espel, Federico Puente

241

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

NASA Astrophysics Data System (ADS)

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.

Prayudhatama, D.; Waris, A.; Kurniasih, N.; Kurniadi, R.

2010-06-01

242

An In-Core Power Deposition and Fuel Thermal Environmental Monitor for Long-Lived Reactor Cores  

SciTech Connect

The primary objective of this program is to develop the Constant Temperature Power Sensor (CTPS) as in-core instrumentation that will provide a detailed map of local nuclear power deposition and coolant thermal-hydraulic conditions during the entire life of the core.

Don W. Miller

2004-09-28

243

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

NASA Technical Reports Server (NTRS)

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.

Anghaie, S.; Chen, G.

1996-01-01

244

Developmental validation of the IrisPlex system: determination of blue and brown iris colour for forensic intelligence.  

PubMed

The IrisPlex system consists of a highly sensitive multiplex genotyping assay together with a statistical prediction model, providing users with the ability to predict blue and brown human eye colour from DNA samples with over 90% precision. This 'DNA intelligence' system is expected to aid police investigations by providing phenotypic information on unknown individuals when conventional DNA profiling is not informative. Falling within the new area of forensic DNA phenotyping, this paper describes the developmental validation of the IrisPlex assay following the Scientific Working Group on DNA Analysis Methods (SWGDAM) guidelines for the application of DNA-based eye colour prediction to forensic casework. The IrisPlex assay produces complete SNP genotypes with only 31pg of DNA, approximately six human diploid cell equivalents, and is therefore more sensitive than commercial STR kits currently used in forensics. Species testing revealed human and primate specificity for a complete SNP profile. The assay is capable of producing accurate results from simulated casework samples such as blood, semen, saliva, hair, and trace DNA samples, including extremely low quantity samples. Due to its design, it can also produce full profiles with highly degraded samples often found in forensic casework. Concordance testing between three independent laboratories displayed reproducible results of consistent levels on varying types of simulated casework samples. With such high levels of sensitivity, specificity, consistency and reliability, this genotyping assay, as a core part of the IrisPlex system, operates in accordance with SWGDAM guidelines. Furthermore, as we demonstrated previously, the IrisPlex eye colour prediction system provides reliable results without the need for knowledge on the bio-geographic ancestry of the sample donor. Hence, the IrisPlex system, with its model-based prediction probability estimation of blue and brown human eye colour, represents a useful tool for immediate application in accredited forensic laboratories, to be used for forensic intelligence in tracing unknown individuals from crime scene samples. PMID:20947461

Walsh, Susan; Lindenbergh, Alexander; Zuniga, Sofia B; Sijen, Titia; de Knijff, Peter; Kayser, Manfred; Ballantyne, Kaye N

2011-11-01

245

Effects of mascara on iris recognition  

NASA Astrophysics Data System (ADS)

Iris biometrics systems rely on analysis of a visual presentation of the human iris, which must be extracted from the periocular region. Topical cosmetics can greatly alter the appearance of the periocular region, and can occlude portions of the iris texture. In this paper, the presence of topical cosmetics is shown to negatively impact the authentic distribution of iris match scores, causing an increase in the false non-match rate at a fixed false match rate.

Doyle, James S.; Flynn, Patrick J.; Bowyer, Kevin W.

2013-05-01

246

A Phase-Based Iris Recognition Algorithm  

Microsoft Academic Search

This paper presents an efficient algorithm for iris recognition using phase-based image matching. The use of phase components in two- dimensional discrete Fourier transforms of iris images makes possible to achieve highly robust iris recognition with a simple matching algorithm. Experimental evaluation using the CASIA iris image database (ver. 1.0 and ver. 2.0) clearly demonstrates an efficient performance of the

Kazuyuki Miyazawa; Koichi Ito; Takafumi Aoki; Koji Kobayashi; Hiroshi Nakajima

2006-01-01

247

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

SciTech Connect

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.

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

248

Estimation of Specific Mass for Multimegawatt NEP Systems Based on Vapor Core Reactors with MHD Power Conversion  

NASA Astrophysics Data System (ADS)

Very low specific-mass power generation in space is possible using Vapor Core Reactors with Magnetohydrodynamic (VCR/MHD) generator. These advanced reactors at the conceptual design level have potential for the generation of tens to hundreds of megawatts of power in space with specific mass of about 1 kg/kWe. Power for nuclear electric propulsion (NEP) is possible with almost direct power conditioning and coupling of the VCR/MHD power output to the VASIMR engine, MPD, and a whole host of electric thrusters. The VCR/MHD based NEP system is designed to power space transportation systems that dramatically reduce the mission time for human exploration of the entire solar system or for aggressive long-term robotic missions. There are more than 40 years of experience in the evaluation of the scientific and technical feasibility of gas and vapor core reactor concepts. The proposed VCR is based on the concept of a cavity reactor made critical through the use of a reflector such as beryllium or beryllium oxide. Vapor fueled cavity reactors that are considered for NEP applications operate at maximum core center and wall temperatures of 4000 K and 1500K, respectively. A recent investigation has resulted in the conceptual design of a uranium tetrafluoride fueled vapor core reactor coupled to a MHD generator. Detailed neutronic design and cycle analyses have been performed to establish the operating design parameters for 10 to 200 MWe NEP systems. An integral system engineering-simulation code is developed to perform parametric analysis and design optimization studies for the VCR/MHD power system. Total system weight and size calculated based on existing technology has proven the feasibility of achieving exceptionally low specific mass (? ~1 kg/kWe) with a VCR/MHD powered system.

Knight, Travis; Anghaie, Samim

2004-02-01

249

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

SciTech Connect

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.

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

2004-10-06

250

Iris Recognition Using Circular Symmetric Filters  

Microsoft Academic Search

This paper proposes a new method for personal identification based on iris recognition. The method consists of three major components: image preprocessing, feature extraction and classifier design. A bank of circular symmetric filters is used to capture local iris characteristics to form a fixed length feature vector. In iris matching, an efficient approach called nearest feature line (NFL) is used.

Li Ma; Yunhong Wang; Tieniu Tan

2002-01-01

251

Minimal template size for iris-recognition  

Microsoft Academic Search

A method to achieve improvement in template size for an iris-recognition system is reported. To achieve this result, the biological characteristics of the human iris have been studied. Processing has been performed by image processing techniques, isolating the iris and enhancing the area of study, after which multi resolution analysis is made. Reduction of the pattern obtained has been obtained

R. Sanchez-Reillo; C. Sanchez-Avila; J. A. Martin-Pereda

1999-01-01

252

Towards non-cooperative iris recognition systems  

Microsoft Academic Search

Iris Technology has been successfully applied to person verification and identification. However, all commercial products require user cooperation for iris image capture. This paper examines the new challenges of iris recognition when extended to less cooperative situations. With the current stress on security and surveillance, this has been an important consideration. First, a summary of research findings of the past

Eric Sung; Xilin Chen; Jie Zhu; Jie Yang

2002-01-01

253

Person identification technique using human iris recognition  

Microsoft Academic Search

The biometric person authentication technique based on the pattern of the human iris is well suited to be applied to any access control system requiring a high level of security. This paper examines a new iris recognition system that implements (i) gradient decomposed Hough transform \\/ integro -differential operators combination for iris localization and (ii) the \\

Lionel MARTIN; Lionel TORRES; Michel ROBERT; ZI Rousset

2002-01-01

254

FEASIBILITY STUDY FOR THE CONCEPTUAL DESIGN OF ADUAL-CORE BOILING SUPERHEAT REACTOR.  

E-print Network

??For research concerning economical applications of high temperature reactortechnology, a novel approach for creating a Boiling Superheat Reactor (BSR) byaugmenting an Advanced Boiling Water Reactor… (more)

Ross, Jacob

2009-01-01

255

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

Microsoft Academic Search

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

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

2010-01-01

256

MHTGR [modular high-temperature gas-cooled reactor] core physics validation plan  

SciTech Connect

This document contains the verification and validation (V&V) plan for analytical methods utilized in the nuclear design for normal and off-normal conditions within the Modular High-Temperature Gas-Cooled Reactor (MHTGR). Regulations, regulatory guides, and industry standards have been reviewed and the approach for V&V has been developed. MHTGR core physics methods are described and the status of previous V&V is summarized within this document. Additional work required to verify and validate these methods is identified. The additional validation work includes comparison of calculations with available experimental data, benchmark comparison of calculations with available experimental data, benchmark comparisons with other validated codes, results from a cooperative program now underway at the Arbeitsgemeinschaft Versuchs-Reaktor GmbH (AVR) facility in Germany, results from a planned series of experiments on the Compact Nuclear Power Source (CNPS) facility at Los Alamos, and detailed documentation of all V&V studies. In addition, information will be obtained from planned international cooperative agreements to provide supplemental data for V&V. The regulatory technology development plan will be revised to include these additional experiments. A work schedule and cost estimate for completing this plan is also provided. This work schedule indicates the timeframe in which major milestones must be performed in order to complete V&V tasks prior to the issuance of preliminary design approval from the NRC. The cost to complete V&V tasks for core physics computational methods is estimated to be $2.2M. 41 refs., 13 figs., 8 tabs.

Baxter, A.; Hackney, R.

1988-01-01

257

Study on Jet Breakup Behavior at Core Disruptive Accident for Fast Breeder Reactor  

SciTech Connect

It is important to estimate the cooling possibility of the molten jet in coolant during a core disruptive accident (CDA) of a fast breeder reactor (FBR). In the present study, the molten jet of U-alloy 78 simulating the core material is injected into the water simulating the coolant. The visual data of the molten jet breakup behavior is observed by using the high-speed video camera. The front velocity of the molten jet is estimated by using the image processing technique from the visual data. It shows that the front velocity of the molten jet can be divided into three time regions. In the first region, the front velocity of the molten jet increases. In the second region, the front velocity of the molten jet suddenly decreases. In the third region, the front velocity of the molten jet keeps at low and steady. In first region, the column diameter of the molten jet decreases with the passage of time. At the location between first region and second region, the column of the molten jet breaks up and disappears. In the present study, the jet breakup length is defined as the distance from the water surface to the location where the jet column disappears. The results show that the jet breakup length depends on the injection nozzle diameter, but does not depend on the jet penetration velocity. This tendency agrees with the prediction by Epstein's equation. After the experiment, the solidified fragments are collected and the mass median diameter is measured. The mass median diameter is compared with the existing theories. Furthermore, a model to estimate the cooling possibility during a CDA of a FBR is constructed, reflecting the above-mentioned results. (authors)

Eiji Matsuo; Yutaka Abe; Hideki Nariai [University of Tsukuba, Tsukuba, 305-8577 Ibaraki (Japan); Keiko Chitose [Mitsubishi Heavy Industries, Ltd. (Japan); Kazuya Koyama [Advanced Reactor Technology Company, Ltd., 15-1, Tomihisa-cho, Shinjuku-ku, Tokyo 162 (Japan); Kazuhiro Itoh [University of Hyogo, Higashikawasaki-cho, Chuo-ku, Kobe-shi, Hyogo (Japan)

2006-07-01

258

Benchmark analysis of high temperature engineering test reactor core using McCARD code  

SciTech Connect

A benchmark calculation has been performed for a startup core physics test of Japan's High Temperature Engineering Test Reactor (HTTR). The calculation is carried out by the McCARD code, which adopts the Monte Carlo method. The cross section library is ENDF-B/VII.0. The fuel cell is modeled by the reactivity-equivalent physical transform (RPT) method. Effective multiplication factors with different numbers of fuel columns have been analyzed. The calculation shows that the HTTR becomes critical with 19 fuel columns with an excess reactivity of 0.84% ?k/k. The discrepancies between the measurements and Monte Carlo calculations are 2.2 and 1.4 % ?k/k for 24 and 30 columns, respectively. The reasons for the discrepancy are thought to be the current version of cross section library and the impurity in the graphite which is represented by the boron concentration. In the future, the depletion results will be proposed for further benchmark calculations. (authors)

Jeong, Chang Joon; Jo, Chang Keun; Lee, Hyun Chul; Noh, Jae Man [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong, Daejeon, 305-353 (Korea, Republic of)

2013-07-01

259

CFD Analysis of Core Bypass Flow and Crossflow in the Prismatic Very High Temperature Gas-cooled Nuclear Reactor  

E-print Network

spectrum nuclear reactor moderated by graphite and cooled by helium. It employs the Brayton cycle to increase the efficiency of reactor. VHTR will use tristructural-isotropic (TRISO) fuel for its high integrity and high pressure boundary of keeping... sustainable chain reactions. Fig. 2 Core Arrangement of GT- MHR (Tak, N.I., Kim, M.-H., et al., 2008) 4 The fuel used for the VHTR is called TRISO fuel which has fuel kernel in the center of the sphere and several layers...

Wang, Huhu 1985-

2012-12-13

260

End-of-life irradiation performance of core structural components in the Shippingport Light Water Breeder Reactor  

SciTech Connect

Nondestructive and destructive end-of-life examinations of Light Water Breeder Reactor (LWBR) core structural components were performed following operation in the Shippingport Atomic Power Station for 29,047 effective full power hours. The Shippingport LWBR demonstrated that breeding can be achieved in a light water reactor with thorium and uranium-233 oxide fuel pellets contained in Zircaloy-4 tubes. The purpose of this presentation is to report results of LWBR core structural component examinations that were carried out to assess the effects of irradiation on support structure and to provide a data base for the evaluation of design procedures. The postirradiation nondestructive examinations included visual inspection and, in some cases, dye penetrant testing to assess structural integrity and surface conditions of the components. Destructive metallography was performed to assess cracking, corrosion buildup, and microstructural condition.

Clayton, J.C.; Smith, B.C.

1991-12-31

261

SAS3D analysis of unprotected loss-of-flow transients for 1200MW(electric) liquid-metal fast breeder reactor homogeneous and heterogeneous core designs  

Microsoft Academic Search

A safety comparison was made for two 1200-MW(electric) liquid-metal fast breeder reactor cores with homogeneous and heterogeneous fuel arrangements, respectively. The two cores were conceptually designed to be identical except for those parameters affected by different fuel arrangements. The comparison was limited to the issue of initiating phase energetics in the hypothetical core disruptive accident. Both cores were assumed to

T. A. Shih; M. I. Temme

1978-01-01

262

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

Microsoft Academic Search

The Molten Salt Reactor (MSR) systems are liquid-fueled reactors that can be used for actinide burning, production of electricity, production of hydrogen, and production of fissile fuels (breeding). Thorium (Th) and uranium-233 (²³³U) are fertile and fissile of the MSR systems, and dissolved in a high-temperature molten fluoride salt (fuel salt) with a very high boiling temperature (up to 1650

Takahisa Yamamoto; Koshi Mitachi; Masatoshi Nishio

2006-01-01

263

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

E-print Network

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...

Hill, Natasha J.; Parks, Geoffrey T.

2014-08-24

264

Space-dependent core\\/reflector boundary conditions generated by the boundary element method for pressurized water reactors  

Microsoft Academic Search

This paper reports on the boundary element method used to generate energy-dependent matrix-type boundary conditions along core\\/reflector interfaces and along baffle-plate surfaces of pressurized water reactors. This method enables one to deal with all types of boundary geometries including convex and concave corners. The method is applicable to neutron diffusion problems with more than two energy groups and also can

M. Itagaki; C. A. Brebbia

1991-01-01

265

A variational transport theory method for two-dimensional reactor core calculations  

NASA Astrophysics Data System (ADS)

It seems very likely that the next generation of reactor analysis methods will be based largely on neutron transport theory, at both the assembly and core levels. Significant progress has been made in recent years toward the goal of developing a transport method that is applicable to large, heterogeneous coarse-meshes. Unfortunately, the major obstacle hindering a more widespread application of transport theory to large-scale calculations is still the computational cost. In this dissertation, a variational heterogeneous coarse-mesh transport method has been extended from one to two-dimensional Cartesian geometry in a practical fashion. A generalization of the angular flux expansion within a coarse-mesh was developed. This allows a far more efficient class of response functions (or basis functions) to be employed within the framework of the original variational principle. New finite element equations were derived that can be used to compute the expansion coefficients for an individual coarse-mesh given the incident fluxes on the boundary. In addition, the non-variational method previously used to converge the expansion coefficients was developed in a new and more thorough manner by considering the implications of the fission source treatment imposed by the response expansion. The new coarse-mesh method was implemented for both one and two-dimensional (2-D) problems in the finite-difference, multigroup, discrete ordinates approximation. An efficient set of response functions was generated using orthogonal boundary conditions constructed from the discrete Legendre polynomials. Several one and two-dimensional heterogeneous light water reactor benchmark problems were studied. Relatively low-order response expansions were used to generate highly accurate results using both the variational and non-variational methods. The expansion order was found to have a far more significant impact on the accuracy of the results than the type of method. The variational techniques provide better accuracy, but at substantially higher computational costs. The non-variational method is extremely robust and was shown to achieve accurate results in the 2-D problems, as long as the expansion order was not very low.

Mosher, Scott W.

266

In vitro propagation of Japanese garden iris, Iris ensata Thunb  

Microsoft Academic Search

Explants of young scapes of Iris ensata were cultured on MS medium with 1 mg\\/l NAA, 1 mg\\/l 6-BA, 30 g\\/l sucrose and 10 g\\/l agar, and this species was characterized by high variety specificity for callus, shoot and root induction. Among 23 varieties and one wild form tested, ‘Okichidori’, ‘Miyukisudare’ and ‘Meiji-l’ exhibited a considerable rate of shoot induction,

T. Yabuya; Y. Ikeda; T. Adachi

1991-01-01

267

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

SciTech Connect

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.

Shadday, M.A.

2001-07-17

268

XRETRIEVE (request, i.a.) (iris.washington.edu)  

E-print Network

XRETRIEVE (request, i.a.) BDSN MEDNET IRIS DMC (iris.washington.edu) GEOSCOPE BREQ_FAST (request) email to BREQ_FAST@sob.iris.washington.edu interactive non-interactive customized pre-assembled customized pre-assembled www.iris.washington.edu (seismiquery/data sources) www.iris.washington.edu (FARM

Laske, Gabi

269

Analysis of partial iris recognition using a 1D approach  

Microsoft Academic Search

Iris recognition has been shown to be very accurate for human identification. We investigate the performance of the use of a partial iris for recognition. A partial iris identification system based on a one-dimensional approach to iris identification is developed. Experiment results show that a more distinguishable and individually unique signal is found in the inner rings of the iris.

Yingzi Du; Bradford Bonney; Robert Ives; Delores Etter; Robert Schultz

2005-01-01

270

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

SciTech Connect

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.

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

271

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

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.

Michael A. Pope

2011-10-01

272

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

NASA Astrophysics Data System (ADS)

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. .

Wright, Steven A.; Lipinski, Ronald J.

2001-02-01

273

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

SciTech Connect

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.

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

1991-10-01

274

Iris microhaemangioma: a management strategy  

PubMed Central

AIM To analyse previous literature and to formulate a management strategy for iris microhaemangiomas (IMH). METHODS A review of the literature in English language articles on IMH. RESULTS Thirty five English language articles fulfilled the criteria for inclusion to the study and based on the contents on these articles a management strategy was formulated. Age at presentation ranged from 42 to 80 years with no sex or racial predisposition. Most patients with IMH have no systemic disease but a higher incidence had been reported in patients with diabetes mellitus, myotonic dystrophy, chronic obstructive pulmonary disease (COPD) and several other systemic and ophthalmic co-morbidities. Most patients remained asymptomatic until they experienced a sudden blurring of vision due to a hyphaema. Some patients only develop a self-limiting single episode of hyphaema and therefore the laser or surgical photocoagulation of iris should be reserved for the cases complicated with recurrent hyphaema. In some patients, several laser photocoagulation sessions may be needed and the recurrent iris vascular tufts may require more aggressive treatment. Iris fluorescein angiography (IFA) is useful in identifying the true extent of the disease and helps to improve the precision of the laser treatment. Surgical excision (iridectomy) should only be considered in patients who fail to respond to repeated laser treatment. In some cases IMHs has been initially misdiagnosed as amaurosis fugax, iritis and Posner-Schlossman syndrome. CONCLUSION Owing to its scarcity, there is no good quality scientific evidence to support the management of IMH. The authors discuss the various treatment options and present a management strategy based on the previous literature for the management for this rare condition and its complications. PMID:23638431

Dharmasena, Aruna; Wallis, Simon

2013-01-01

275

Applications of the IRI in Southern Africa  

NASA Astrophysics Data System (ADS)

The IRI forms the basis of the Single Site Location Direction Finding networks of the South African Defence Force as well as theNational Intelligence Agency. It is also used in "Path Analysis" applications where the possible transmitter coverage is calculated. Another application of the IRI is in HF frequency predictions, especially for the South African Defence Force involved in peace keeping duties in Africa. The IRI is either used independently or in conjunction with vertical ionosondes. In the latter case the scaled F2 peak parameters (foF2, hmF2) are used as inputs to the IRI. The IRI thus gets "calibrated" to extend the area covered by the ionosonde(s). The IRI has proved to be a very important tool in South Africa and Africa in the fight against crime, drug trafficking, political instability and maintaining the peace in potentially unstable countries.

Coetzee, P. J.

2004-01-01

276

Fiber-Pigtailed Electrothermal MEMS Iris VOA  

Microsoft Academic Search

A high-performance variable optical attenuator, which is based on an electrothermally actuated iris with a square pupil, is demonstrated. The device is fabricated from two separate dies that are formed by deep reactive ion etching of bonded silicon-on-insulator material. The iris die is inserted into an elastic clamp on a base-plate die carrying spring-mounted fiber alignment features, allowing the iris

Hadi Veladi; Richard R. A. Syms; Helin Zou

2007-01-01

277

Local intensity variation analysis for iris recognition  

Microsoft Academic Search

As an emerging biometric for human identification, iris recognition has received increasing attention in recent years. This paper makes an attempt to reflect shape information of the iris by analyzing local intensity variations of an iris image. In our framework, a set of one-dimensional (1-D) intensity signals is constructed to contain the most important local variations of the original two-dimensional

Li Ma; Tieniu Tan; Yunhong Wang; Dexin Zhang

2004-01-01

278

Robust Iris Recognition Using Advanced Correlation Techniques  

Microsoft Academic Search

\\u000a The iris is considered one of the most reliable and stable biometrics as it is believed to not change significantly during\\u000a a person’s lifetime. Standard techniques for iris recognition, popularized by Daugman, apply Gabor wavelet analysis for feature\\u000a extraction. In this paper, we consider an alternative method for iris recognition, the use of advanced distortion-tolerant\\u000a correlation filters for robust pattern

Jason Thornton; Marios Savvides; B. V. K. Vijaya Kumar

2005-01-01

279

Iris Recognition with Support Vector Machines  

Microsoft Academic Search

\\u000a We propose an iris recognition system for the identification of persons using support vector machines. Canny’s edge detection\\u000a and the Hough transform are used to find the iris\\/pupil boundary and a simple thresholding method is employed for eyelash\\u000a detection. The Gabor wavelet technique is deployed in order to extract the deterministic features in the transformed iris\\u000a of a person in

Kaushik Roy; Prabir Bhattacharya

2006-01-01

280

Iris segmentation using variational level set method  

NASA Astrophysics Data System (ADS)

Continuous efforts have been made to process degraded iris images for enhancement of the iris recognition performance in unconstrained situations. Recently, many researchers have focused on developing the iris segmentation techniques, which can deal with iris images in a non-cooperative environment where the probability of acquiring unideal iris images is very high due to gaze deviation, noise, blurring, and occlusion by eyelashes, eyelids, glasses, and hair. Although there have been many iris segmentation methods, most focus primarily on the accurate detection of iris images captured in a closely controlled environment. The novelty of this research effort is that we propose to apply a variational level set-based curve evolution scheme that uses a significantly larger time step to numerically solve the evolution partial differential equation (PDE) for segmentation of an unideal iris image accurately, and thereby, speeding up the curve evolution process drastically. The iris boundary represented by the variational level set may break and merge naturally during evolution, and thus, the topological changes are handled automatically. The proposed variational model is also robust against poor localization and weak iris/sclera boundaries. In order to solve the size irregularities occurring due to arbitrary shapes of the extracted iris/pupil regions, a simple method is applied based on connection of adjacent contour points. Furthermore, to reduce the noise effect, we apply a pixel-wise adaptive 2D Wiener filter. The verification and identification performance of the proposed scheme is validated on three challenging iris image datasets, namely, the ICE 2005, the WVU Unideal, and the UBIRIS Version 1.

Roy, Kaushik; Bhattacharya, Prabir; Suen, Ching Y.

2011-04-01

281

Iris Recognition in Less Constrained Environments  

Microsoft Academic Search

Iris recognition is one of the most accurate forms of biometric identifi- cation. However, current commercial off-the-shelf\\u000a (COTS) systems generally impose significant constraints on the subject. This chapter discusses techniques for iris image capture\\u000a that reduce those constraints, in particular enabling iris image capture from moving subjects and at greater distances than\\u000a have been available in the COTS systems. The

James R. Matey; David Ackerman; James Bergen; Michael Tinker

282

Securing iris recognition systems against masquerade attacks  

NASA Astrophysics Data System (ADS)

A novel two-stage protection scheme for automatic iris recognition systems against masquerade attacks carried out with synthetically reconstructed iris images is presented. The method uses different characteristics of real iris images to differentiate them from the synthetic ones, thereby addressing important security flaws detected in state-of-the-art commercial systems. Experiments are carried out on the publicly available Biosecure Database and demonstrate the efficacy of the proposed security enhancing approach.

Galbally, Javier; Gomez-Barrero, Marta; Ross, Arun; Fierrez, Julian; Ortega-Garcia, Javier

2013-05-01

283

IRIS thermal balance test within ESTEC LSS  

NASA Technical Reports Server (NTRS)

The Italian Research Interim Stage (IRIS) thermal balance test was successfully performed in the ESTEC Large Space Simulator (LSS) to qualify the thermal design and to validate the thermal mathematical model. Characteristics of the test were the complexity of the set-up required to simulate the Shuttle cargo bay and allowing IRIS mechanism actioning and operation for the first time in the new LSS facility. Details of the test are presented, and test results for IRIS and the LSS facility are described.

Messidoro, Piero; Ballesio, Marino; Vessaz, J. P.

1988-01-01

284

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

SciTech Connect

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)

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

285

Biometric iris recognition system using a fast and robust iris localization and alignment procedure  

Microsoft Academic Search

Iris recognition as a biometric technique for personal identification and verification is examined. The motivation for this stems from the observation that the human iris provides a unique structure suitable for non-invasive biometric assessment. In particular the irises are as distinct as fingerprints or patterns of retinal blood vessels and the appearance of the iris is amenable to remote examination.

Balaji Ganeshan; Dhananjay Theckedath; Rupert Young; Chris Chatwin

2006-01-01

286

Direct Attacks Using Fake Images in Iris Verification  

Microsoft Academic Search

In this contribution, the vulnerabilities of iris-based recognition systems to direct attacks are studied. A database of fake\\u000a iris images has been created from real iris of the BioSec baseline database. Iris images are printed using a commercial printer\\u000a and then, presented at the iris sensor. We use for our experiments a publicly available iris recognition system, which some\\u000a modifications

Virginia Ruiz-albacete; Pedro Tome-gonzalez; Fernando Alonso-fernandez; Javier Galbally; Julian Fiérrez-aguilar; Javier Ortega-garcia

2008-01-01

287

Use of PRA Techniques to Optimize the Design of the IRIS Nuclear Power Plant  

Microsoft Academic Search

True design optimization of a plant=s inherent safety and performance characteristics results when a probabilistic risk assessment (PRA) is integrated with the plant- level design process. This is the approach being used throughout the design of the International Reactor Innovative and Secure (IRIS) nuclear power plant to maximize safety. A risk-based design optimization tool employing a \\

M. D. Muhlheim; J. W. Cletcher

288

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

NASA Astrophysics Data System (ADS)

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, SiO2 particles have been coated with TiO2 using tetrakis-dimethylamino titanium (TDMAT) and H2O 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 SiO2 particles were coated with a 1.6 nm homogenous shell of TiO2.

Didden, Arjen P.; Middelkoop, Joost; Besling, Wim F. A.; Nanu, Diana E.; van de Krol, Roel

2014-01-01

289

Leveraging community support for Education and Outreach: The IRIS E&O Program  

NASA Astrophysics Data System (ADS)

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.

Taber, J.; Hubenthal, M.; Wysession, M. E.

2009-12-01

290

Iris recognition using Dynamic Programming Matching Pursuit  

Microsoft Academic Search

In this paper, we propose a new method named dynamic programming based matching pursuit algorithm for iris-based personal identification. Based on the matching pursuit algorithm, it selects the most representative path to do iris recognition. Our system consists of identification and verification. Finally, we use the experimental results to demonstrate the efficacy of the proposed method and show that it

Yueh-Shiun Lee; Chung-Lin Huang; Meng-Fen Ho; Wen-Liang Huang

2008-01-01

291

Iris indocyanine green videoangiography in diabetic iridopathy  

Microsoft Academic Search

AIMS\\/BACKGROUND: Iris fluorescein angiography (IFA) is not commonly used in clinical practice, although its value has been demonstrated especially in cases of diabetic disease. IFA is able to show neovascular tufts in order to guide the laser treatment, and it is highly recommended in diabetic patients who need cataract surgery or vitrectomy. Nevertheless, IFA fails to demonstrate the iris vascular

M B Parodi; E Bondel; D Russo; G Ravalico

1996-01-01

292

Optimizations in Iris Recognition A Dissertation  

E-print Network

error rate. We utilized an active contour model to refine the noise detection results and optimizedOptimizations in Iris Recognition A Dissertation Submitted to the Graduate School of the University Rights Reserved #12;Optimizations in Iris Recognition Abstract by Xiaomei Liu Biometric verification

Flynn, Patrick J.

293

A system for automated iris recognition  

Microsoft Academic Search

This paper describes a prototype system for personnel verification based on automated iris recognition. The motivation for this endeavour stems from the observation that the human iris provides a particularly interesting structure on which to base a technology for noninvasive biometric measurement. In particular, it is known in the biomedical community that irises are as distinct as fingerprints or patterns

R. P. Wildes; J. C. Asmuth; G. L. Green; S. C. Hsu; R. J. Kolczynski; J. R. Matey; S. E. McBride

1994-01-01

294

Computational imaging systems for iris recognition  

Microsoft Academic Search

Computational imaging systems are modern systems that consist of generalized aspheric optics and image processing capability. These systems can be optimized to greatly increase the performance above systems consisting solely of traditional optics. Computational imaging technology can be used to advantage in iris recognition applications. A major difficulty in current iris recognition systems is a very shallow depth-of-field that limits

Robert J. Plemmons; Michael Horvath; Emily Leonhardt; Paul Pauca; Sudhakar Prasad; Stephen B. Robinson; Harsha Setty; Todd C. Torgersen; Joseph van der Gracht; Edward Dowski; Ramkumar Narayanswamy; Paulo E. X. Silveira

2004-01-01

295

IRIS: Recent Earthquake Teachable Moments  

NSDL National Science Digital Library

This page offers a compilation of multimedia resources for creating "teachable moments" on the earthquakes. Of particular note is the material for the earthquake/tsunami that struck northern Japan in 2011. Tectonic maps, computer animations, seismograms, aerial and ground photographs, Power Point slides for teachers, news footage, preliminary rupture models, and a comparison map that shows tectonic similarities between the coast of northeastern Japan and the west coast of the United States are available. This web site is maintained by IRIS, the Incorporated Research Institutions for Seismology, a consortium of laboratories and data collection centers who act in concert to ensure flow of data to the international seismological research community.

2011-03-16

296

Analyses of Greek Research Reactor with mixed HEU-LEU Be reflected core  

SciTech Connect

The fuel-cycle analyses presented in this paper provide specific steps to be taken in the transition from a 36-element water-reflected HEU core to a 33-element LEU equilibrium core with a Be reflector on two faces. The first step will be to install the Be reflector and remove the highest burnup HEU fuel. The smaller Be-reflected core will be refueled with LEU fuel. All analyses were performed using a planar 5-group REBUS3 model benchmarked to VIM Monte Carlo. In addition to fuel cycle results, the control rod worth, reactivity response to increased fuel and water temperature and decreased water density were compared for the transition core and the reference HEU core.

Deen, J.R.; Snelgrove, J.L. [Argonne National Lab., IL (United States); Papastergiou, K. [National Center for Scientific Research, Athens (Greece)

1993-12-31

297

Enhanced iris recognition method based on multi-unit iris images  

NASA Astrophysics Data System (ADS)

For the purpose of biometric person identification, iris recognition uses the unique characteristics of the patterns of the iris; that is, the eye region between the pupil and the sclera. When obtaining an iris image, the iris's image is frequently rotated because of the user's head roll toward the left or right shoulder. As the rotation of the iris image leads to circular shifting of the iris features, the accuracy of iris recognition is degraded. To solve this problem, conventional iris recognition methods use shifting of the iris feature codes to perform the matching. However, this increases the computational complexity and level of false acceptance error. To solve these problems, we propose a novel iris recognition method based on multi-unit iris images. Our method is novel in the following five ways compared with previous methods. First, to detect both eyes, we use Adaboost and a rapid eye detector (RED) based on the iris shape feature and integral imaging. Both eyes are detected using RED in the approximate candidate region that consists of the binocular region, which is determined by the Adaboost detector. Second, we classify the detected eyes into the left and right eyes, because the iris patterns in the left and right eyes in the same person are different, and they are therefore considered as different classes. We can improve the accuracy of iris recognition using this pre-classification of the left and right eyes. Third, by measuring the angle of head roll using the two center positions of the left and right pupils, detected by two circular edge detectors, we obtain the information of the iris rotation angle. Fourth, in order to reduce the error and processing time of iris recognition, adaptive bit-shifting based on the measured iris rotation angle is used in feature matching. Fifth, the recognition accuracy is enhanced by the score fusion of the left and right irises. Experimental results on the iris open database of low-resolution images showed that the averaged equal error rate of iris recognition using the proposed method was 4.3006%, which is lower than that of other methods.

Shin, Kwang Yong; Kim, Yeong Gon; Park, Kang Ryoung

2013-04-01

298

Scaling analysis of the coupled heat transfer process in the high-temperature gas-cooled reactor core  

SciTech Connect

The differential equations representing the coupled heat transfer from the solid nuclear core components to the helium in the coolant channels are scaled in terms of representative quantities. This scaling process identifies the relative importance of the various terms of the coupled differential equations. The relative importance of these terms is then used to simplify the numerical solution of the coupled heat transfer for two bounding cases of full-power operation and depressurization from full-system operating pressure for the Fort St. Vrain High-Temperature Gas-Cooled Reactor. This analysis rigorously justifies the simplified system of equations used in the nuclear safety analysis effort at Oak Ridge National Laboratory.

Conklin, J.C.

1986-08-01

299

KINE--A one-dimensional dynamics program for pressurized water reactors with partial boiling in the core  

SciTech Connect

A short description of the reactor dynamics program KINE is given. The KINE code is a onedimensional code solving the time-dependent neutron diffusion equations in a two-group representation, taking into account six groups of delayed neutron precursors, one heating channel composed of fuel, canning, and a coolant region in which boiling may occur. Special attention is given to radial averaging of coolant density and fuel temperature in a core in which partial boiling can exist. The method of solution is a modified backward extrapolation procedure. Stationary starting conditions are achieved by adjustment of control rod position or liquid poison concentration.

Fiebig, R.; Kruger, A.

1982-06-01

300

KUGEL: a thermal, hydraulic, fuel performance, and gaseous fission product release code for pebble bed reactor core analysis  

SciTech Connect

The KUGEL computer code is designed to perform thermal/hydraulic analysis and coated-fuel particle performance calculations for axisymmetric pebble bed reactor (PBR) cores. This computer code was developed as part of a Department of Energy (DOE)-funded study designed to verify the published core performance data on PBRs. The KUGEL code is designed to interface directly with the 2DB code, a two-dimensional neutron diffusion code, to obtain distributions of thermal power, fission rate, fuel burnup, and fast neutron fluence, which are needed for thermal/hydraulic and fuel performance calculations. The code is variably dimensioned so that problem size can be easily varied. An interpolation routine allows variable mesh size to be used between the 2DB output and the two-dimensional thermal/hydraulic calculations.

Shamasundar, B.I.; Fehrenbach, M.E.

1981-05-01

301

Relationship between core size, coolant choice, fuel type, and neutron flux in a fast irradiation test reactor  

SciTech Connect

Currently, the United States has no domestic capability for large volume irradiation testing in a fast-spectrum system to support the development of advanced fuels or materials. The recently-proposed Global Nuclear Energy Partnership includes provisions for a sodium-cooled Advanced Burner Test Reactor which could provide this testing capability. In addition to sodium, lead-bismuth eutectic and helium coolants are being considered for future energy systems. In this paper, sodium, lead-bismuth eutectic, and helium-cooled systems are evaluated to determine the impact on fast flux and fuel enrichment resulting from varying core diameter, fuel volume fraction, fuel type, and coolant. While fast flux is most strongly influenced by core diameter at fixed power, fuel enrichment is a more complicated function of all four parameters. In the end, the combination of high fast flux and low enrichment can best be achieved by a sodium-cooled system. (authors)

Fanning, T. H. [Nuclear Engineering Div., Argonne National Laboratory, Argonne, IL 60439 (United States)

2006-07-01

302

Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2006  

SciTech Connect

Neutronics and thermal-hydraulics studies show that, for equivalent operating power [85 MW(t)], a low-enriched uranium (LEU) fuel cycle based on uranium-10 wt % molybdenum (U-10Mo) metal foil with radially, “continuously graded” fuel meat thickness results in a 15% reduction in peak thermal flux in the beryllium reflector of the High Flux Isotope Reactor (HFIR) as compared to the current highly enriched uranium (HEU) cycle. The uranium-235 content of the LEU core is almost twice the amount of the HEU core when the length of the fuel cycle is kept the same for both fuels. Because the uranium-238 content of an LEU core is a factor of 4 greater than the uranium-235 content, the LEU HFIR core would weigh 30% more than the HEU core. A minimum U-10Mo foil thickness of 84 ?m is required to compensate for power peaking in the LEU core although this value could be increased significantly without much penalty. The maximum U-10Mo foil thickness is 457?m. Annual plutonium production from fueling the HFIR with LEU is predicted to be 2 kg. For dispersion fuels, the operating power for HFIR would be reduced considerably below 85 MW due to thermal considerations and due to the requirement of a 26-d fuel cycle. If an acceptable fuel can be developed, it is estimated that $140 M would be required to implement the conversion of the HFIR site at Oak Ridge National Laboratory from an HEU fuel cycle to an LEU fuel cycle. To complete the conversion by fiscal year 2014 would require that all fuel development and qualification be completed by the end of fiscal year 2009. Technological development areas that could increase the operating power of HFIR are identified as areas for study in the future.

Primm, R. T. [ORNL] [ORNL; Ellis, R. J. [ORNL] [ORNL; Gehin, J. C. [ORNL] [ORNL; Clarno, K. T. [ORNL] [ORNL; Williams, K. A. [ORNL] [ORNL; Moses, D. L. [ORNL] [ORNL

2006-11-01

303

Development of a low enrichment uranium core for the MIT reactor  

E-print Network

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 ...

Newton, Thomas Henderson

2006-01-01

304

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

E-print Network

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 ...

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

2007-01-01

305

A core reload pattern and composition optimization methodology for pressurized water reactors  

E-print Network

The primary objective of this research was the development of a comprehensive, rapid and conceptually simple methodology for PWR core reload pattern and fuel composition optimization, capable of systematic incorporation ...

Sauer, Ildo Luis

1985-01-01

306

Fuzzy difference-of-Gaussian-based iris recognition method for noisy iris images  

NASA Astrophysics Data System (ADS)

Iris recognition is used for information security with a high confidence level because it shows outstanding recognition accuracy by using human iris patterns with high degrees of freedom. However, iris recognition accuracy can be reduced by noisy iris images with optical and motion blurring. We propose a new iris recognition method based on the fuzzy difference-of-Gaussian (DOG) for noisy iris images. This study is novel in three ways compared to previous works: (1) The proposed method extracts iris feature values using the DOG method, which is robust to local variations of illumination and shows fine texture information, including various frequency components. (2) When determining iris binary codes, image noises that cause the quantization error of the feature values are reduced with the fuzzy membership function. (3) The optimal parameters of the DOG filter and the fuzzy membership function are determined in terms of iris recognition accuracy. Experimental results showed that the performance of the proposed method was better than that of previous methods for noisy iris images.

Kang, Byung Jun; Park, Kang Ryoung; Yoo, Jang-Hee; Moon, Kiyoung

2010-06-01

307

An Iris Segmentation Algorithm based on Edge Orientation for Off-angle Iris Recognition  

SciTech Connect

Iris recognition is known as one of the most accurate and reliable biometrics. However, the accuracy of iris recognition systems depends on the quality of data capture and is negatively affected by several factors such as angle, occlusion, and dilation. In this paper, we present a segmentation algorithm for off-angle iris images that uses edge detection, edge elimination, edge classification, and ellipse fitting techniques. In our approach, we first detect all candidate edges in the iris image by using the canny edge detector; this collection contains edges from the iris and pupil boundaries as well as eyelash, eyelids, iris texture etc. Edge orientation is used to eliminate the edges that cannot be part of the iris or pupil. Then, we classify the remaining edge points into two sets as pupil edges and iris edges. Finally, we randomly generate subsets of iris and pupil edge points, fit ellipses for each subset, select ellipses with similar parameters, and average to form the resultant ellipses. Based on the results from real experiments, the proposed method shows effectiveness in segmentation for off-angle iris images.

Karakaya, Mahmut [ORNL; Barstow, Del R [ORNL; Santos-Villalobos, Hector J [ORNL; Boehnen, Chris Bensing [ORNL

2013-01-01

308

IRI, an International Standard for the Ionosphere  

NASA Astrophysics Data System (ADS)

The International Reference Ionosphere (IRI) is a data-based model of the ionosphere that has been steadily improved and updated by a joint working group of the Committee on Space Research and the International Union of Radio Science. We will report about the most recent IRI workshops and the improvements and additions planned for the next version of the model. In particular new models will be included for the D-region electron density (Friedrich et al., 2002), and for the ion densities (Triskova et al., 2003) the latter based on Atmosphere Explorer C, D, E and Intercosmos 24 data. A correction term will be introduced in the topside electron density model to alleviate problems at high solar activities and high altitudes (Bilitza, 2002). A special IRI task groups is working on an occurrence probability model for spread-F (Abdu et al., 2003) for inclusion in IRI. A quantitative description of ionospheric variability (standard deviation from monthly mean) is the goal of a special IRI task force activity at the International Center for Theoretical Physics (Radicella 2002). We will also report about activities to update IRI with actual measurements and thus obtain a more accurate description of the actual ionosphere. A proposal to make the IRI model the ISO standard for the ionosphere is now pending before the International Standardization Organization (ISO). The IRI homepage is at http://nssdc.gsfc.nasa.gov/space/model/ionos/iri.html and a web-interface for computing and plotting IRI parameters can be found at http://nssdc.gsfc.nasa.gov/space/model/models/iri.html . Abdu, M. A., J. R de Souza, I. S. Batista, and J. H. A. Sobral, Equatorial Spread F statistics and their empirical modeling for the IRI: A regional model for the Brazilian longitude sector, Adv. Space Res., in press, 2003. Triskova, L., V. Truhlik and J. Smilauer, An empirical model of ion composition in the outer ionosphere, Adv. Space Res., in press, 2003 Bilitza, D., A Correction for the IRI Topside Model Based on Alouette/ISIS Data, World Space Congress, Houston, Texas, 2002. Friedrich, M., M. Harrich, R. Steiner, K. M. Torkar, and F.-J. Luebken, The quiet auroral ionosphere and its neutral background, World space congress, Houston, Texas, 2002.

Bilitza, D.; Reinisch, B.; Triskova, L.; Friedrich, M.

2003-04-01

309

Contribution to modeling of the reflooding of a severely damaged reactor core using PRELUDE experimental results  

SciTech Connect

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. The reflooding (injection of water into core) may be applied if the availability of safety injection is recovered during accident. If the injection becomes available only in the late phase of accident, water will enter a core configuration that will differ significantly from original rod-bundle geometry. Any attempt to inject water after significant core degradation can lead to further fragmentation of core material. The fragmentation of fuel rods 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), i.e., a high permeability porous medium. The French 'Institut de Radioprotection et de Surete Nucleaire' is developing experimental programs (PEARL and PRELUDE) and simulation tools (ICARE-CATHARE and ASTEC) to study and optimize the severe accident management strategy and to assess the probabilities to stop the progress of in-vessel core degradation. It is shown that the quench front exhibits either a ID behaviour or a 2D one, depending on injection rate or bed characteristics. The PRELUDE experiment covers a rather large range of variation of parameters, for which the developed model appears to be quite predictive. (authors)

Bachrata, A.; Fichot, F.; Repetto, G. [Institut de Radioprotection et de Surete Nucleaire IRSN, Cadarache (France); Quintard, M. [Universite de Toulouse, INPT, UPS, IMFT Institut de Mecanique des Fluides de Toulouse, Allee Camille Soula, F-31400 Toulouse (France); CNRS, IMFT, F-31400 Toulouse (France); Fleurot, J. [Institut de Radioprotection et de Surete Nucleaire IRSN, Cadarache (France)

2012-07-01

310

Atmospheric reentry of the in-core thermionic SP-100 reactor system  

NASA Technical Reports Server (NTRS)

Presumed end-of-life atmospheric reentry of the GA SP-100 system was studied to assess dispersal feasibility and associated hazards. Reentry was studied by sequential use of an orbital trajectory and a heat analysis computer program. Two heating models were used. The first model assumed a thermal equilibrium condition between the stagnation point aerodynamic heating and the radiative cooling of the skin material surface. The second model allowed for infinite conductivity of the skin material. Four reentering configurations were studied representing stages of increased SP-100 breakup: (1) radiator, shield and reactor, (2) shield and reactor, (3) reactor with control drums, and (4) reactor without control drums. Each reentering configuration was started from a circular orbit at 116 km having an inertial velocity near Mach 25. The assumed failing criterion was the attainment of melting temperature of a critical system component. The reentry analysis revealed breakup of the vessel in the neighborhood of 61 km altitude and scattering of the fuel elements. Subsequent breakup of the fuel elements was not predicted. Oxidation of the niobium skin material was calculated to cause an increase in surface temperature of less than ten percent. The concept of thermite analogs for enhancing reactor reentry dispersal was assessed and found to be feasible in principle. A conservative worst-case hazards analysis was performed for radioactive and nonradioactive toxic SP-100 materials assumed to be dispersed during end-of-life reentry. The hazards associated with this phase of the SP-100 mission were calculated to be insignificant.

Stamatelatos, M. G.; Barsell, A. W.; Harris, P. A.; Francisco, J.

1987-01-01

311

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 ...

312

Boosting ordinal features for accurate and fast iris recognition  

Microsoft Academic Search

In this paper, we present a novel iris recognition method based on learned ordinal features.Firstly, taking full advan- tages of the properties of iris textures, a new iris representa- tion method based on regional ordinal measure encoding is presented, which provides an over-complete iris feature set for learning. Secondly, a novel Similarity Oriented Boost- ing (SOBoost) algorithm is proposed to

Zhaofeng He; Zhenan Sun; Tieniu Tan; Xianchao Qiu; Cheng Zhong; Wenbo Dong

2008-01-01

313

Real-Time Image Restoration for Iris Recognition Systems  

Microsoft Academic Search

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

Byung Jun Kang; Kang Ryoung Park

2007-01-01

314

Iris recognition using self-organizing neural network  

Microsoft Academic Search

Among biometric systems for user verification, iris recognition systems represent a relatively new technology. Our system consists of two main parts: a localizing iris and iris pattern recognition. The raw image is captured using a digital camera. The iris is then extracted from the background after enhancement and noise elimination. Due to noise and the high degree of freedom in

Lye Wil Liam; A. Chekima; Liau Chung Fan; J. A. Dargham

2002-01-01

315

Review of iris recognition: cameras, systems, and their applications  

Microsoft Academic Search

Purpose – To overview the iris cameras, iris recognition systems, and their applications. Design\\/methodology\\/approach – Introduced and examined commercially available or lab prototype iris cameras and systems to compare their functionalities and applications. Findings – Each kind of camera has its advantage and disadvantage. From the application view, each iris recognition system has its unique values. Originality\\/value – This paper

Yingzi Eliza Du

2006-01-01

316

The relative distance of key point based iris recognition  

Microsoft Academic Search

Iris recognition has received increasing attention in recent years as a reliable approach to human identification. This paper makes an attempt to analyze the local feature structure of iris texture information based on the relative distance of key points. When preprocessed, the annular iris is normalized into a rectangular block. Multi-channel 2-D Gabor filters are used to capture the iris

Li Yu; David Zhang; Kuanquan Wang

2007-01-01

317

Toward Noncooperative Iris Recognition: A Classification Approach Using Multiple Signatures  

Microsoft Academic Search

This paper focuses on noncooperative iris recognition, i.e., the capture of iris images at large distances, under less controlled lighting conditions, and without active participation of the subjects. This increases the probability of capturing very heterogeneous images (regarding focus, contrast, or brightness) and with several noise factors (iris obstructions and reflections). Current iris recognition systems are unable to deal with

Hugo Proença; Luís A. Alexandre

2007-01-01

318

Analytical support for the ORR (Oak Ridge Research Reactor) whole-core LEU U/sub 3/Si/sub 2/-Al fuel demonstration  

SciTech Connect

Analytical methods used to analyze neutronic data from the whole-core LEU fuel demonstration in the Oak Ridge Research Reactor are briefly discussed. Calculated eigenvalues corresponding to measured critical control rod positions are presented for each core used in the gradual transition from an all HEU to an all LEU configuration. Some calculated and measured results, including ..beta../sub eff//l/sub p/, are compared for HEU and LEU fresh fuel criticals. Finally, the perturbing influences of the six voided beam tubes on certain core parameters are examined. For reasons yet to be determined, differential shim rod worths are not well-calculated in partially burned cores.

Bretscher, M.M.

1986-01-01

319

Generation of iris codes using 1D Log-Gabor filter  

Microsoft Academic Search

In general, a typical iris recognition system includes iris imaging, iris liveness detection, iris image quality assessment, and iris recognition. This paper presents an algorithm focusing on the last two steps. The novelty of this algorithm includes improving the speed and accuracy of the iris segmentation process, assessing the iris image quality such that only the clear images are accepted

A. T. Kahlil; F. E. M. Abou-Chadi

2010-01-01

320

Dynamic Comparison of Three- and Four-Equation Reactor Core Models in a Full-Scope Power Plant Training Simulator  

SciTech Connect

A comparative analysis of the dynamic behavior of a boiling water reactor in a full-scope power plant simulator for operator training is presented. Three- and four-equation reactor core models were used to examine three transients following tests described in acceptance test procedures: scram, loss of feedwater flow, and closure of main isolation valves. The three-equation model consists of water and steam mixture momentum, including mass and energy balances. The four-equation model is based on liquid and gas phase mass balances, together with a drift-flux approach for the analysis of phase separation. Analysis of the models showed that the scram transient was slightly different for three- and four-equation models. The drift-flux effects can explain such differences. Regarding the loss-of-feedwater transient, the predicted steam flow after scram is larger for the three-equation model. Finally, for the transient related to the closure of main steam isolation valves, the three-equation model provides slightly different results for the pressure change, which affects reactor level behavior.

Espinosa-Paredes, Gilberto [Universidad Autonoma Metropolitana-Iztapalapa (Mexico); Alvarez-Ramirez, Jose [Universidad Autonoma Metropolitana-Iztapalapa (Mexico); Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias (Mexico); Garcia-Gutierrez, Alfonso [Instituto de Investigaciones Electricas (Mexico); Martinez-Mendez, Elizabeth Jeannette [Instituto de Investigaciones Electricas (Mexico)

2004-02-15

321

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)

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.

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

322

SUMER-IRIS Observations of AR11875  

NASA Astrophysics Data System (ADS)

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.

Schmit, Donald; Innes, Davina

2014-05-01

323

Chromospheric Diagnostics from IRIS and DST  

NASA Astrophysics Data System (ADS)

Using data obtained during a coordinated observing campaign in September 2013, we compare the spectral and imaging diagnostics from IRIS and the instruments at the Dunn Solar Telescope (DST). We focus on a small active region observed for approximately one hour with IRIS (NUV, FUV, and SJI) in conjunction with IBIS, FIRS, and ROSA from the DST.In particular, we examine the line widths and intensities in the different chromospheric lines (H-alpha, Ca II 8542, Mg II) and the temporal evolution of these different diagnostics. This allows us to better relate the views from new window provided by IRIS to previous studies of the chromosphere.

Cauzzi, Gianna; Reardon, Kevin P.; Jaeggli, Sarah A.; Reid, Aaron

2014-06-01

324

Myopia and iris colour: a possible connection?  

PubMed

Myopia is a common ocular disease in the world. Its prevalence has increased rapidly worldwide, especially in some East-Asian countries. Genetic factors and environmental factors both affect myopia's onset and its progress. Iris colour is an important characteristic of a person. It is a possible risk factor for myopia by affecting the amount and the colour of light entering eyes. The study of iris colour may contribute to the understanding of myopia mechanism and provide good suggestive evidence for studies on other eye diseases. In this article, the possible connection between myopia and iris colour is proposed. Approaches to dissect any link are suggested. PMID:22465466

Meng, Weihua; Butterworth, Jacqueline; Calvas, Patrick; Malecaze, Francois

2012-06-01

325

Asymmetric Introgressive Hybridization Among Louisiana Iris Species  

PubMed Central

In this review, we discuss findings from studies carried out over the past 20+ years that document the occurrence of asymmetric introgressive hybridization in a plant clade. In particular, analyses of natural and experimental hybridization have demonstrated the consistent introgression of genes from Iris fulva into both Iris brevicaulis and Iris hexagona. Furthermore, our analyses have detected certain prezygotic and postzygotic barriers to reproduction that appear to contribute to the asymmetric introgression. Finally, our studies have determined that a portion of the genes transferred apparently affects adaptive traits. PMID:24710008

Arnold, Michael L.; Tang, Shunxue; Knapp, Steven J.; Martin, Noland H.

2010-01-01

326

Edge detection techniques for iris recognition system  

NASA Astrophysics Data System (ADS)

Nowadays security and authentication are the major parts of our daily life. Iris is one of the most reliable organ or part of human body which can be used for identification and authentication purpose. To develop an iris authentication algorithm for personal identification, this paper examines two edge detection techniques for iris recognition system. Between the Sobel and the Canny edge detection techniques, the experimental result shows that the Canny's technique has better ability to detect points in a digital image where image gray level changes even at slow rate.

Tania, U. T.; Motakabber, S. M. A.; Ibrahimy, M. I.

2013-12-01

327

[Managing complications in intraoperative floppy iris syndrome].  

PubMed

The intraoperative floppy iris syndrome (IFIS) describes an ophthalmologically relevant phenomenon which is observed after systemic intake of alpha blockers for treatment of benign prostate hyperplasia. This leads to an increase in intraoperative complications in cataract surgery characterized by a flaccid iris which billows in response to currents with a tendency to prolapse towards the area of surgery. This results in damage to the iris by the instruments used or posterior capsule rupture with loss of lens material. We describe the preoperative and intraoperative measures and techniques to deal with this challenging situation in order to minimize development of IFIS and reduce the complication rate. PMID:23338529

Handzel, D M; Rausch, S; Kälble, T; Briesen, S

2013-04-01

328

Sensitivity of power peaking analysis to large reactor core modeling. [LMFBR  

Microsoft Academic Search

Various models of large LMFBRs, based on cylindrical and hexagonal geometries, are examined in regard to application to nuclear power peaking analysis. It is shown that the general behavior of power distributions during burnup in these large reactors implies that power shaping by control rod movement is desirable for minimizing the peaking factor. Due to current limitations of available three-dimensional

1976-01-01

329

Core melt progression and consequence analysis methodology development in support of the Savannah River Reactor PSA  

SciTech Connect

A three-level Probabilistic Safety Assessment (PSA) of production reactor operation has been underway since 1985 at the US Department of Energy's Savannah River Site (SRS). The goals of this analysis are to: Analyze existing margins of safety provided by the heavy-water reactor (HWR) design challenged by postulated severe accidents; Compare measures of risk to the general public and onsite workers to guideline values, as well as to those posed by commercial reactor operation; and Develop the methodology and database necessary to prioritize improvements to engineering safety systems and components, operator training, and engineering projects that contribute significantly to improving plant safety. PSA technical staff from the Westinghouse Savannah River Company (WSRC) and Science Applications International Corporation (SAIC) have performed the assessment despite two obstacles: A variable baseline plant configuration and power level; and a lack of technically applicable code methodology to model the SRS reactor conditions. This paper discusses the detailed effort necessary to modify the requisite codes before accident analysis insights for the risk assessment were obtained.

O'Kula, K.R.; Sharp, D.A. (Westinghouse Savannah River Co., Aiken, SC (United States)); Amos, C.N.; Wagner, K.C.; Bradley, D.R. (Science Applications International Corp., Albuquerque, NM (United States))

1992-01-01

330

A Non-linear Normalization Model for Iris Recognition  

Microsoft Academic Search

\\u000a Iris-based biometric recognition outperforms other biometric methods in terms of accuracy. In this paper an iris normalization\\u000a model for iris recognition is proposed, which combines linear and non-linear methods to unwrap the iris region. First, non-linearly\\u000a transform all iris patterns to a reference annular zone with a predefined ?, which is the ratio of the radii of inner and outer

Xiaoyan Yuan; Pengfei Shi

2005-01-01

331

Region-based SIFT approach to iris recognition  

Microsoft Academic Search

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

Craig Belcher; Yingzi Du

2009-01-01

332

Use of one-dimensional iris signatures to rank iris pattern similarities  

Microsoft Academic Search

A one-dimensional approach to iris recognition is presented. It is translation-, rotation-, illumination-, and scale-invariant. Traditional iris recognition systems typically use a two-dimensional iris signature that requires circular rotation for pattern matching. The new approach uses the Du measure as a matching mechanism, and generates a set of the most probable matches (ranks) instead of only the best match. Since

Yingzi Du; Robert W. Ives; Delores M. Etter; Thad B. Welch

2006-01-01

333

Methods of fabricating uranium monocarbide, mononitride, and carbonitride cores for fast reactor fuel elements  

Microsoft Academic Search

Laboratory methods of obtaining uranium carbide, nitride, and ; carbonitride, and fabricating fuel-element cores from them were developed. The ; uranium compounds were obtained from reactions between metallic uranium and ; nitrogen, preceded by hydrogenation of the metal. A continuous process was ; achieved using specially developed equipment that yielded 1.5 to 2 kg of product ; per hour. Uranium

F. G. Reshetnikov; R. B. Kotelnikov; B. D. Rogozkin; S. N. Bashlykov; I. A. Samokhvalov; G. V. Titov; M. G. Shishkov; V. S. Belevantsev; Yu. E. Fedorov; V. P. Simonov

1973-01-01

334

Pediatric HIV Immune Reconstitution Inflammatory Syndrome (IRIS)  

PubMed Central

Summary Purpose of Review Little is known regarding HIV immune reconstitution inflammatory syndrome (IRIS) in children. As the ART roll out has gathered pace since 2004 in resource limited settings, pediatric IRIS has emerged as a clinical challenge. Recent Findings The incidence of IRIS appears to be between 10-20%. The commonest causes are mostly mycobacterial, including tuberculosis, atypical mycobacteria and BCG-related However, in many pediatric cohorts, a marked early mortality within the first 90 days of antiretroviral therapy (ART) occurs. This mortality is poorly understood, and the contribution of IRIS to this mortality is unknown. Summary Children after starting ART may have paradoxical worsening of previously treated opportunistic infections. However due to the differences in children’s immunology with vertical HIV transmission, children are likely are greater risk of unmasking occult, subclinical infections during immune reconstitution. PMID:19373006

Boulware, David R.; Callens, Steven; Pahwa, Savita

2009-01-01

335

Iris Recognition with Low Template Size  

Microsoft Academic Search

Among all the biometric techniques known nowadays, Iris Recognition is taken as the most promising of all, due to its low\\u000a error rates without being invasive and with low relation to police records. Based on Daugman’s work, the authors have developed\\u000a their own Iris Recognition system, obtaining results that show the performance of the prototype and proves the excellences\\u000a of

Raul Sánchez-reillo; Carmen Sanchez-avila

2001-01-01

336

Review of the SIMMER-II analyses of liquid-metal-cooled fast breeder reactor core-disruptive accident fuel escape  

SciTech Connect

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.

DeVault, G.P.; Bell, C.R.

1985-01-01

337

A model for calculating composition and density of the core melt in the water-moderated water-cooled reactor in case of severe accident  

Microsoft Academic Search

Thermochemical behavior of the core melt in the VVER-type reactor at severe accident is discussed. Experimental information\\u000a gained made it possible to construct a thermodynamic model of the O-U-Zr-Fe system. The model describes the immiscibility\\u000a of the oxide and metal phases of the core melt and makes it possible to estimate their densities. Parameters of the model\\u000a were obtained by

V. D. Ozrin; O. V. Tarasov; V. F. Strizhov; A. S. Filippov

2010-01-01

338

A model for calculating composition and density of the core melt in the water-moderated water-cooled reactor in case of severe accident  

Microsoft Academic Search

Thermochemical behavior of the core melt in the VVER-type reactor at severe accident is discussed. Experimental information gained made it possible to construct a thermodynamic model of the O-U-Zr-Fe system. The model describes the immiscibility of the oxide and metal phases of the core melt and makes it possible to estimate their densities. Parameters of the model were obtained by

V. D. Ozrin; O. V. Tarasov; V. F. Strizhov; A. S. Filippov

2010-01-01

339

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

E-print Network

10 TRISO Fuel Particle and Homogenized Fuel Compact???????...... 33 11 Arrangement of Materials in Triangular-Pitch Unit Cell???????? 34 12 Fuel Compact Arrangement in DANCOFF-MC???????????. 35 13 Map of Core and Fuel Zones...????????????????????.. 46 18 Triangular-Pitch Unit Cell (homogenized model)...?????????... 47 19 TRISO Fuel Particle and Homogenized Fuel Pebble?????????.. 49 20 Triangular Pitch Unit Cell (homogenized fuel pebble model)?????? 50 21 BCC Unit Cell...

Ames, David E, II

2006-10-30

340

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

SciTech Connect

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

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

2012-07-01

341

Assessment of core damage frequency owing to possible fires at NPP with RBMK type reactors  

SciTech Connect

According to Scientific and Technical Cooperation between the USA and Russia in the field of nuclear engineering the Idaho National Laboratory has transferred to the possession of the National Research Center ' Kurchatov Inst. ' the SAPHIRE software without any fee. With the help of the software Kurchatov Inst. developed a Pilot Living PSA- Model of Leningrad NPP Unit 1. Computations of core damage frequencies were carried out for additional Initiating Events. In the submitted paper such additional Initiating Events are fires in various compartments of the NPP. During the computations of each fire, structure of the PSA - Model was not changed, but Fault Trees for the appropriate systems, which are removed from service during the fire, were changed. It follows from the computations, that for ten fires Core Damaged Frequencies (CDF) are not changed. Other six fires will cause additional core damage. On the basis of the calculated results it is possible to determine a degree of importance of these fires and to establish sequence of performance of fire-prevention measures in various places of the NPP. (authors)

Vinnikov, B. [National Research Centre Kurchatov Inst., 1, Kurchatov Square, Moscow, 123 182 (Russian Federation); NRC Kurchatov Inst. (Russian Federation)

2012-07-01

342

Fibrous congenital iris membranes with pupillary distortion.  

PubMed Central

BACKGROUND: In 1986 Cibis and associates described 2 children with a new type of congenital pupillary-iris-lens membrane with goniodysgenesis that was unilateral, sporadic, and progressive. These membranes were different from the common congenital pupillary strands that extend from 1 portion of the iris collarette to another or from the iris collarette to a focal opacity on the anterior lens surface. They also differed from the stationary congenital hypertrophic pupillary membranes that partially occlude the pupil, originating from multiple sites on the iris collarette, but not attaching directly to the lens. CASE MATERIAL: The present report is an account of 7 additional infants with congenital iris membranes, similar to those reported by Cibis and associates, which caused pupillary distortion and were variably associated with adhesions to the lens, goniodysgenesis, and progressive occlusion or seclusion of the pupil. Six of the 7 patients required surgery to open their pupils for visual purposes or to abort angle closure glaucoma. A remarkable finding was that the lenses in the area of the newly created pupils were clear, allowing an unobstructed view of normal fundi. CONCLUSION: This type of fibrous congenital iris membrane is important to recognize because of its impact on vision and its tendency to progress toward pupillary occlusion. Timely surgical intervention can abort this progressive course and allow vision to be preserved. PMID:11797319

Robb, R M

2001-01-01

343

Improved iris localization by using wide and narrow field of view cameras for iris recognition  

NASA Astrophysics Data System (ADS)

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.

Kim, Yeong Gon; Shin, Kwang Yong; Park, Kang Ryoung

2013-10-01

344

Fuel Summary for Peach Bottom Unit 1 High-Temperature Gas-Cooled Reactor Cores 1 and 2  

SciTech Connect

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.

Karel I. Kingrey

2003-04-01

345

Particle image velocimetry measurements in a representative gas-cooled prismatic reactor core model for the estimation of bypass flow  

NASA Astrophysics Data System (ADS)

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%.

Conder, Thomas E.

346

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

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

C. L. Cowan; R. Protsik; J. W. Lewellen

1984-01-01

347

Thin gauge nickel-iron cores--Specification for use in high-power pulse transformers and saturable reactors with reset  

Microsoft Academic Search

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.

P. C. HOOKE; D. Driver; R. Major

1974-01-01

348

Modeling & analysis of core debris recriticality during hypothetical severe accidents in the Advanced Neutron Source Reactor  

SciTech Connect

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 KEN05A-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 KEN05-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 criticality event in the ANS is described. Strong dependencies on geometry, material constituents, and thermal-hydraulic conditions are described. The introduction of designed mitigative features are described.

Kim, S.H.; Georgevich, V.; Simpson, D.B.; Slater, C.O.; Taleyarkhan, R.P.

1992-10-01

349

Review of cystic and solid tumors of the iris  

PubMed Central

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

Shields, Carol L.; Shields, Patrick W.; Manalac, Janet; Jumroendararasame, Chaisiri; Shields, Jerry A.

2013-01-01

350

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.  

Code of Federal Regulations, 2013 CFR

...cooling systems for light-water nuclear power reactors. 50.46 Section...cooling systems for light-water nuclear power reactors. (a)(1...boiling or pressurized light-water nuclear power reactor fueled...

2013-01-01

351

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.  

...cooling systems for light-water nuclear power reactors. 50.46 Section...cooling systems for light-water nuclear power reactors. (a)(1...boiling or pressurized light-water nuclear power reactor fueled...

2014-01-01

352

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.  

Code of Federal Regulations, 2012 CFR

...cooling systems for light-water nuclear power reactors. 50.46 Section...cooling systems for light-water nuclear power reactors. (a)(1...boiling or pressurized light-water nuclear power reactor fueled...

2012-01-01

353

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.  

Code of Federal Regulations, 2011 CFR

...cooling systems for light-water nuclear power reactors. 50.46 Section...cooling systems for light-water nuclear power reactors. (a)(1...boiling or pressurized light-water nuclear power reactor fueled...

2011-01-01

354

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.  

Code of Federal Regulations, 2010 CFR

...cooling systems for light-water nuclear power reactors. 50.46 Section...cooling systems for light-water nuclear power reactors. (a)(1...boiling or pressurized light-water nuclear power reactor fueled...

2010-01-01

355

Improving IRI at low solar activity  

NASA Astrophysics Data System (ADS)

A number of recent studies have found that thermospheric and ionospheric densities reached extremely low values during the last solar minimum, which was an unusually deep and extended minimum. This was also documented in comparisons with the International Reference Ionosphere (IRI) model which as an empirical model represents conditions as recorded during prior minima. IRI significantly overestimates the electron density in the topside ionosphere while it shows good agreement with ionosonde measurements of the F-peak density. The ionosphere was much more contracted during this minimum than IRI predicted based on the data from prior solar minima. In addition the extremely low neutral densities also resulted in a lowering of the F-layer. The combined effect explains the large (60-80%) data-model discrepancies. We have continued the study of ionospheric behavior during the last solar minimum with the large volume of ionospheric Total Electron Content (TEC) data recorded by the TOPEX/Jason satellites from 1992 to 2011. Specifically we investigate differences to the prior solar minimum in terms of measured TEC as well as IRI predictions. We will discuss IRI improvements based on these comparisons.

Bilitza, Dieter; Brown, Steven; Beckley, Brian

2012-07-01

356

ORNL Biometric Eye Model for Iris Recognition  

SciTech Connect

Iris recognition has been proven to be an accurate and reliable biometric. However, the recognition of non-ideal iris images such as off angle images is still an unsolved problem. We propose a new biometric targeted eye model and a method to reconstruct the off-axis eye to its frontal view allowing for recognition using existing methods and algorithms. This allows for existing enterprise level algorithms and approaches to be largely unmodified by using our work as a pre-processor to improve performance. In addition, we describe the `Limbus effect' and its importance for an accurate segmentation of off-axis irides. Our method uses an anatomically accurate human eye model and ray-tracing techniques to compute a transformation function, which reconstructs the iris to its frontal, non-refracted state. Then, the same eye model is used to render a frontal view of the reconstructed iris. The proposed method is fully described and results from synthetic data are shown to establish an upper limit on performance improvement and establish the importance of the proposed approach over traditional linear elliptical unwrapping methods. Our results with synthetic data demonstrate the ability to perform an accurate iris recognition with an image taken as much as 70 degrees off-axis.

Santos-Villalobos, Hector J [ORNL; Barstow, Del R [ORNL; Karakaya, Mahmut [ORNL; Chaum, Edward [University of Tennessee, Knoxville (UTK); Boehnen, Chris Bensing [ORNL

2012-01-01

357

IRI and GPS Variations Over Ilorin Nigeria  

NASA Astrophysics Data System (ADS)

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.

Okonkwo, Perpetua; Okoh, Daniel

358

Interests & Recruitment In Science http://iris.fp-7.org/  

E-print Network

in STEM-related decision-making. Interests and Recruitment In Science (IRIS): Aims Project IRIS PUBLISHABLE SUMMARY Interests & Recruitment In Science Factors influencing recruitment, retention and gender individuals and provides opportunities for self development, career and democratic participation. Improved

Johansen, Tom Henning

359

Information Security: Securing Smart Cards With IRIS Recognition.  

National Technical Information Service (NTIS)

This thesis examines the application of iris recognition technology to the problem of keeping smart cards secure. In order to understand the technology, a comprehensive literature review was conducted. The biological components of the iris were examined t...

O. E. Phelps

2001-01-01

360

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

SciTech Connect

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.

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

2006-07-31

361

On a Quest to Improve the Solar Forcing in IRI  

NASA Astrophysics Data System (ADS)

The International Reference Ionosphere (IRI) is an empirical model of the ionosphere based on a large volume of ground and space measurements that was developed under the auspices of the Committee on Space Research (COSPAR) and the International Union of Radio Science (URSI) and that earlier this year became an international standard of the International Standard Organization (ISO). IRI currently uses several solar and ionospheric indices to describe the variations of ionospheric parameters with solar variability. These indices are used at an averaging level of 81 days or even a whole year. We have investigated the performance of these different indices at different averaging lengths using over 30 years of ionosonde foF2 data from the three stations Boulder, Jicamarca, and Grahamstown employing daily and monthly averages of foF2. In addition to the indices currently used in IRI our study also included indices composed of measured EUV fluxes (Lyman alpha -121.5nm, MgII-core-wing-flux-ratio, Integral flux 0-105nm). However, coverage gaps during the last two solar cycle maxima introduce uncertainties for these indices. We get the best results with Lyman alpha fluxes at an averaging length of about 81 days (3 solar rotations). The ionospheric-effective solar index IG, which is based on ionosonde data from five selected stations, performs almost equally well as the Lyman-alpha flux index. Surprisingly, we find that the monthly IG index performs as well if not better than the 12-month running mean of monthly IG that is currently used in IRI. This opens interesting possibilities for using a GIRO-based IG index (IGiro) that could be determined by averaging across a global selection of ionosonde stations available on the Global Ionospheric Radio Observatory (GIRO) at a much higher time resolution (down to 15 minutes) in near real-time. Most importantly such a new index could be designed such that it would not be limited by the constraints of the current IG index, which is determined with only noon data and with using the CCIR maps and thus should not be applied for nighttime and/or URSI maps.

Bilitza, D.; Brown, S.; Chamberlin, P. C.

2013-12-01

362

Structural response of reactor-core hexcan subassemblies subjected to dynamic overpressurization under accident conditions  

SciTech Connect

This paper presents a two-dimensional structural analysis for the evaluation of a single core subassembly due to internal overpressure associated with possible failure of fuel pins having high fission gas plenum pressure. Structural models are developed for the subassemblies and their surroundings with emphasis on the critical physical aspects of the problem. With these models the strains, deformations and the extent of permanent damage (plastic strain) to the subassemblies can be assessed. The nonlinear structural analyses was performed with a finite element program called STRAW (Structural Transient Response of Assembly Wrappers). This finite element program is applicable to nonlinear large displacement problems. The results of this study indicate that the permanent deformation (damage) is strongly influenced by the rise time (time to reach peak pressure) of the pressure pulse and the pressure in the fuel pin. The rise time is influenced by the opening time of the flow path for release of gas from the fuel pin plenum. Several examples are illustrated with various rise times and pressure magnitudes and the resulting permanent deformation of the hexcan wall.

Pfeiffer, P.A.; Kulak, R.F.

1993-01-01

363

Structural response of reactor-core hexcan subassemblies subjected to dynamic overpressurization under accident conditions  

SciTech Connect

This paper presents a two-dimensional structural analysis for the evaluation of a single core subassembly due to internal overpressure associated with possible failure of fuel pins having high fission gas plenum pressure. Structural models are developed for the subassemblies and their surroundings with emphasis on the critical physical aspects of the problem. With these models the strains, deformations and the extent of permanent damage (plastic strain) to the subassemblies can be assessed. The nonlinear structural analyses was performed with a finite element program called STRAW (Structural Transient Response of Assembly Wrappers). This finite element program is applicable to nonlinear large displacement problems. The results of this study indicate that the permanent deformation (damage) is strongly influenced by the rise time (time to reach peak pressure) of the pressure pulse and the pressure in the fuel pin. The rise time is influenced by the opening time of the flow path for release of gas from the fuel pin plenum. Several examples are illustrated with various rise times and pressure magnitudes and the resulting permanent deformation of the hexcan wall.

Pfeiffer, P.A.; Kulak, R.F.

1993-06-01

364

Dynamic structural response of reactor-core subassemblies (hexcans) due to accident overpressurization  

SciTech Connect

This paper presents a two-dimensional structural analysis for the evaluation of a single core subassembly due to internal overpressure associated with possible failure of fuel pins having high fission gas plenum pressure. Structural models are developed for the subassemblies and their surroundings with emphasis on the critical physical aspects of the problem. With these models the strains, deformations and the extent of permanent damage (plastic strain) to the subassemblies can be assessed. The nonlinear structural analyses was performed with a finite element program called STRAW (Structural Transient Response of Assembly Wrappers). This finite element program is applicable to nonlinear large displacement problems. The results of this study indicate that the permanent deformation (damage) is strongly influenced by the rise time (time to reach peak pressure) of the pressure pulse and the pressure in the fuel pin. The rise time is influenced by the opening time of the flow path for release of gas from the fuel pin plenum. Several examples are illustrated with various rise times and pressure magnitudes and the resulting permanent deformation of the hexcan wall.

Pfeiffer, P.A.; Kulak, R.F.

1993-08-01

365

A Novel Method to Extract Features for Iris Recognition System  

Microsoft Academic Search

In general, the iris recognition systems have used the wavelet transform as feature extraction techniques. Since the wavelet\\u000a transform does not have the shift-invariant property, the iris features are inconsistently extracted due to the eye image\\u000a rotation and the inexact iris localization. In this paper, a novel method to extract features is proposed for iris recognition\\u000a system. Two types of

Seung-in Noh; Kwanghyuk Bae; Jaihie Kim

2003-01-01

366

Graph Matching Iris Image Blocks with Local Binary Pattern  

Microsoft Academic Search

Abstract. Iris-based personal identification has attracted much ,attention in re- cent years. Almost all the state-of-the-art iris recognition algorithms are based onstatistical classifier and local image features, which are noise sensitive and hardly to deliver perfect recognition performance. In this paper, we propose a novel iris recognition method, using the histogram of local binary pattern for global iris texture representation

Zhenan Sun; Tieniu Tan; Xianchao Qiu

2006-01-01

367

Enhancement of REBUS-3/DIF3D for whole-core neutronic analysis of prismatic very high temperature reactor (VHTR).  

SciTech Connect

Enhancements have been made to the REBUS-3/DIF3D code suite to facilitate its use for the design and analysis of prismatic Very High Temperature Reactors (VHTRs). A new cross section structure, using table-lookup, has been incorporated to account for cross section changes with burnup and fuel and moderator temperatures. For representing these cross section dependencies, three new modules have been developed using FORTRAN 90/95 object-oriented data structures and implemented within the REBUS-3 code system. These modules provide a cross section storage procedure, construct microscopic cross section data for all isotopes, and contain a single block of banded scattering data for efficient data management. Fission products other than I, Xe, Pm, and Sm, can be merged into a single lumped fission product to save storage space, memory, and computing time without sacrificing the REBUS-3 solution accuracy. A simple thermal-hydraulic (thermal-fluid) feedback model has been developed for prismatic VHTR cores and implemented in REBUS-3 for temperature feedback calculations. Axial conduction was neglected in the formulation because of its small magnitude compared to radial (planar) conduction. With the simple model, the average fuel and graphite temperatures are accurately estimated compared to reference STAR-CD results. The feedback module is currently operational for the non-equilibrium fuel cycle analysis option of REBUS-3. Future work should include the extension of this capability to the equilibrium cycle option of the code and additional verification of the feedback module. For the simulation of control rods in VHTR cores, macroscopic cross section deviations (deltas) have been defined to account for the effect of control rod insertion. The REBUS-3 code has been modified to use the appropriately revised cross sections when control rods are inserted in a calculation node. In order to represent asymmetric core blocks (e.g., fuel blocks or reflector blocks containing asymmetric absorber rods), surface-dependent discontinuity factors based on nodal equivalence theory have been introduced into the nodal diffusion theory option of the DIF3D code (DIF3D-nodal) to improve modeling accuracy. Additionally, the discontinuity factors based on the Simplified Equivalence Theory (SET) have been incorporated as an alternative and may be employed for both the DIF3D-nodal and DIF3D-VARIANT (nodal transport) solution options. Two- and three-dimensional core calculations have been performed using the routines developed and modified in this work, along with cross sections generated from single fuel block and one-dimensional or two-dimensional fuel-reflector model. Generally, REBUS-3/DIF3D results for the core multiplication factor and power distribution are found to be in good agreement with reference results (generated with MCNP continuous energy calculations) particularly when discontinuity factors are applied. The DIF3D-VARIANT option was found to provide a more accurate solution in its diffusion approximation than the DIF3D-nodal option. Control rod worths can be estimated with acceptably small errors compared to MCNP results. However, estimation of the core power tilt needs to be improved by introducing the surface-dependent discontinuity factor capability in DIF3D-VARIANT.

Lee, C. H.; Zhong, Z.; Taiwo, T.A.; Yang, W.S.; Khalil, H.S.; Smith, M.A.; Nuclear Engineering Division

2006-10-13

368

Jets and Bombs: Characterizing IRIS Spectra  

NASA Astrophysics Data System (ADS)

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.

Schmit, Donald; Innes, Davina

2014-06-01

369

Tunable liquid iris actuated using electrowetting effect  

NASA Astrophysics Data System (ADS)

A configuration for a tunable liquid iris, which consists simply of two immiscible liquids and two flat indium tin oxide (ITO) glass substrates, is proposed. The two immiscible liquids are transparent salt solution and opaque oil, respectively. The top ITO electrode was precoated with a 2-?m-thick polydimethylsiloxane film as the dielectric layer, while the surface of the bottom electrode was specially treated using ultraviolet irradiation to define specific hydrophilic regions. The iris aperture's diameter could easily be regulated by varying the direct current bias voltages between the two electrodes. Results show that the aperture diameter can be continuously varied from 1.5 mm at the voltage-off state to 3.5 mm at a bias of 350 V. This liquid iris takes the advantages of low fabrication cost, fast response time, low-power consumption, and easy reversibility without the need of any mechanical movable parts.

Yu, Cheng-Chian; Ho, Jeng-Rong; Cheng, J.-W. John

2014-05-01

370

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

371

Increased Iris-lens Contact Following Spontaneous Blinking: Mathematical Modeling  

PubMed Central

The purpose of this work was to study in silico how iris root rotation due to spontaneous blinking alters the iris contour. An axisymmetric finite-element model of the anterior segment was developed that included changes in the iris contour and the aqueous humor flow. The model geometry was based on average values of ocular dimensions. Blinking was modeled by rotating the iris root posteriorly and returning it back to the anterior. Simulations with maximum rotations of 2°, 4°, 6°, and 8° were performed. The iris-lens contact distance and the pressure difference between the posterior and anterior chambers were calculated. When the peak iris root rotation was 2°, the maximum iris-lens contact increased gradually from 0.28 to 0.34 mm within eight blinks. When the iris root was rotated 6° and 8°, the pressure difference between the posterior and anterior chambers dropped from a positive value (1.23 Pa) to negative values (?0.86 and ?1.93 Pa) indicating the presence of reverse pupillary block. Apparent iris-lens contact increased with steady blinking, and the increase became more pronounced as posterior rotation increased. We conclude that repeated iris root rotation caused by blinking could maintain the iris in a posterior position under normal circumstances, which would then lead to the clinically-observed anterior drift of the iris when blinking is prevented. PMID:22819357

Amini, Rouzbeh; Jouzdani, Sara

2012-01-01

372

An Effective Iris Recognition System Based on Biomimetic Pattern Recognition  

Microsoft Academic Search

This paper presents a new method for effective iris recognition using Biomimetic Pattern Recognition (BPR), which is a new theory proposed by academician ShoujueWang. A model for iris recognition that is based on BPR was introduced and thoroughly discussed here. Experimental results on the Chinese Academy of Sciences, Institute of Automation (CASIA) iris image database clearly demonstrates that the use

Junying Zeng; Yikui Zhai; Junying Gan; Ying Xu

2009-01-01

373

Biomicroscopy and fluorescein angiography of pigmented iris tumors  

Microsoft Academic Search

The classification of pigmented iris tumors is a difficult clinical problem. Based on the retrospective observation of colour photographs and iris angiograms of 44 pigmented iris tumors observed over 19 years, the authors present an original grading scheme with scores depending on both the biomicroscopical and the fluoroiridographic patterns of the tumors. The biomicroscopical parameters considered were: thickening of the

Francesco Bandello; Rosario Brancato; Rosangela Lattanzio; Alfonso Carnevalini; Antonio Rossi; Gabriel Coscas

1994-01-01

374

Iris colour and relationship of tyrosinase activity to adrenergic innervation  

Microsoft Academic Search

IRIS colour to some degree depends on the sympathetic innervation of the eye. Chance clinical observations of humans and experimental studies in animals have demonstrated consistently a gradual lessening of colour intensity in the iris after interruption of the sympathetic pathways to the eye1,2. Similarly, full iris pigmentation fails to develop when the sympathetic innervation to the eye is absent

Alan M. Laties

1975-01-01

375

Cataract surgery combined with implantation of an artificial iris  

Microsoft Academic Search

We describe 6 patients who presented with cataract or aphakia and absent or nonfunctional irides. The etiologies included congenital aniridia, traumatic iris loss, and chronic mydriasis secondary to recurrent herpetic uveitis. In 5 eyes, a prosthetic iris was successfully implanted in combination with small incision cataract surgery. In 2 eyes, a single-piece iris diaphragm and optical lens was implanted. Artificial

Robert H Osher; Scott E Burk

1999-01-01

376

Non-Ideal Iris Segmentation Using Graph Cuts  

Microsoft Academic Search

A non-ideal iris segmentation approach using graph cuts is presented. Unlike many existing algorithms for iris local- ization which extensively utilize eye geometry, the propos ed approach is predominantly based on image intensities. In a step-wise procedure, first eyelashes are segmented from the input images using image texture, then the iris is segmented using grayscale information, followed by a postprocessing

Shrinivas J. PundlikDamon; L. Woodard; Stanley T. Birchfield

2008-01-01

377

Estimating and Fusing Quality Factors for Iris Biometric Images  

Microsoft Academic Search

Iris recognition, the ability to recognize and distinguish individuals by their iris pattern, is one of the most reliable biometrics in terms of recognition and identification performance. However, the performance of these systems is affected by poor-quality imaging. In this paper, we extend iris quality assessment research by analyzing the effect of various quality factors such as defocus blur, off-angle,

Nathan D. Kalka; Jinyu Zuo; Natalia A. Schmid; Bojan Cukic

2010-01-01

378

Robust Direction Estimation of Gradient Vector Field for Iris Recognition  

Microsoft Academic Search

As a reliable personal identification method, iris recognition has been receiving increasing attention. Based on the theory of robust statistics, a novel geometry-driven method for iris recognition is presented in this paper. An iris image is considered as a 3D surface of piecewise smooth patches. The direction of the 2D vector, which is the planar projection of the normal vector

Zhenan Sun; Yunhong Wang; Tieniu Tan; Jiali Cui

2004-01-01

379

Cascading statistical and structural classifiers for iris recognition  

Microsoft Academic Search

Reliable human identification using iris pattern has recently gained growing interests from pattern recognition researchers. In literature of iris recognition, almost all algorithms are based on statistical information. In this paper, a structural iris image analysis method is proposed, which provides complementary information to statistical classifier. In order to save computational cost, the structural matcher is not consulted unless the

Zhenan Sun; Yunhong Wang; Tieniu Tan; Jiali Ciri

2004-01-01

380

Effect of Severe Image Compression on Iris Recognition Performance  

Microsoft Academic Search

Abstract We investigate three schemes for severe compression of iris images, in order to assess what their impact would be on recognition performance of the algorithms deployed today for identifying persons by this biometric feature. Currently, standard iris images are 600 times larger than the IrisCode templates computed from them for database storage and search; but it is administratively desired

John Daugman; Cathryn Downing

2008-01-01

381

A robust eyelash detection based on iris focus assessment  

Microsoft Academic Search

For accurate iris recognition, it is essential to detect eyelash regions and remove them for iris code generation, since eyelashes act as noise factors in the iris recognition. In addition, eyelash positions can be changed for enrollment and recognition and this may cause FR (false rejection). To overcome these problems, we propose a new method for detecting eyelashes in this

Byung Jun Kang; Kang Ryoung Park

2007-01-01

382

Iris recognition based on score level fusion by using SVM  

Microsoft Academic Search

In conventional iris recognition methods, due to the difficulty of selecting one optimal wavelet filter for iris feature extraction, multiple wavelet filters (with different frequencies and kernel sizes) are adopted. However, this causes the processing time and the extracted feature size to increase. To overcome this problem, feature level fusion of the extracted iris features has been proposed, but this

Hyun-ae Park; Kang Ryoung Park

2007-01-01

383

IRIS TOXICOLOGICAL REVIEW AND SUMMARY DOCUMENTS FOR BERYLLIUM AND COMPOUNDS  

EPA Science Inventory

EPA's assessment of the noncancer health effects and carcinogenic potential of Beryllium was added to the IRIS database in 1998. The IRIS program is updating the IRIS assessment for Beryllium. This update will incorporate health effects information published since the last assess...

384

[Iris-fixated intraocular lenses: reinforced monitoring].  

PubMed

In 1986, the concept of the claw lens was applied to correct myopia in phakic patients. Since then, progress has made iris-fixated phakic intraocular lenses (IOL) relatively safe, predictable, and effective for the correction of myopia, hyperopia, and astigmatism. All these models have undergone a series of design improvements to prevent complications. Despite having excellent refractive results, the principal risk is a potential progressive endothelial cell loss. Many authors have presented encouraging results. Phakic iris-fixated IOL surgery is a potentially reversible procedure, but the surgeon cannot rule out the possibility of complications. Therefore, long-term follow-up is mandatory. PMID:19520458

Fournié, P; Malecaze, F

2009-11-01

385

Advanced MOX Core Design Study of Sodium-Cooled Reactors in Current Feasibility Study on Commercialized Fast Reactor Cycle Systems in Japan  

SciTech Connect

Sodium-cooled mixed-oxide core design studies are performed with a target burnup of 150 GWd/t and possible measures against the recriticality issues in core disruptive accidents. Four types of core are compared from the viewpoints of core performance and reliability. Results show that all the types of core satisfy the target and that a homogeneous core with an axial blanket partial elimination subassembly is the superior concept, although experimental demonstration is required of molten fuel motion for mitigation of recriticality following fuel melting and loss of fuel pin integrity.

Mizuno, Tomoyasu; Niwa, Hajime [Japan Nuclear Cycle Development Institute (Japan)

2004-05-15

386

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)

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.

Gavrilov, A. V.; Kritskii, V. G.; Rodionov, Yu. A.; Berezina, I. G.

2013-07-01

387

Estimate of radiation-induced steel embrittlement in the BWR core shroud and vessel wall from reactor-grade MOX/UOX fuel for the nuclear power plant at Laguna Verde, Veracruz, Mexico  

NASA Astrophysics Data System (ADS)

The government of Mexico has expressed interest to utilize the Laguna Verde boiling water reactor (BWR) nuclear power plant for the disposition of reprocessed spent uranium oxide (UOX) fuel in the form of reactor-grade mixed-oxide (MOX) fuel. MOX fuel would replace spent UOX fuel as a fraction in the core from 18--30% depending on the fuel loading cycle. MOX fuel is expected to increase the neutron fluence, flux, fuel centerline temperature, reactor core pressure, and yield higher energy neutrons. There is concern that a core with a fraction of MOX fuel (i.e., increased 239Pu wt%) would increase the radiation-induced steel embrittlement within the core shroud and vessel wall as compared to only conventional, enriched UOX fuel in the core. The evaluation of radiation-induced steel embrittlement within the core shroud and vessel wall is a concern because of the potentially adverse affect to plant and public safety, environment, and operating life of the reactor. This dissertation provides computational results of the neutron fluence, flux, energy spectrum, and radiation damage displacements per atom per second (dpa-s-1) in steel within the core shroud and vessel wall of the Laguna Verde Unit 1 BWR. The results were computed using the nuclear data processing code NJOY99 and the continuous energy Monte Carlo Neutral Particle transport code MCNP4B. The MCNP4B model of the reactor core was for maximum core loading fractions of ? MOX and ? UOX reactor-grade fuel in an equilibrium core. The primary conclusion of this dissertation was that the addition of the maximum fraction of ? MOX fuel to the LV1 BWR core did significantly accelerate the radiation-induced steel embrittlement such that without mitigation of steel embrittlement by periodic thermal annealing or reduction in operating parameters such as, neutron fluence, core temperature and pressure, it posed a potentially adverse affect to the plant and public safety, environment, and operating life of the reactor.

Vickers, Lisa Rene

388

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

SciTech Connect

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

Farzad Rahnema

2009-11-12

389

IRIS Update Batch 1, Group 1  

EPA Science Inventory

Update the following IRIS chemical dose-response assessments: Barium (cancer, RfC), o-Cresol (RfD, cancer), carbon disulfied (RfD, RfC), 1,1-Dichloroethane (cancer), 2,4-Dimethylphenol (RfD), 1,4-Dibromobenzene (RfD), 1-chloro-1,1-difluroelfane (RfC, Acetyl chloride (cancer),2,4...

390

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...

391

IRIS Neutron Guide: Options for the Future.  

National Technical Information Service (NTIS)

The aim of the present work is to study the feasibility of various ideas for further improving the performance of the IRIS neutron guide on the ISIS pulsed neutron source. In particular, the following questions are addressed: (1) How to increase the inten...

P. Benassi, C. J. Carlile

1991-01-01

392

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.

2010-11-03

393

Pigment Melanin: Pattern for Iris Recognition  

Microsoft Academic Search

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

S. Mahdi Hosseini; Babak Nadjar Araabi; Hamid Soltanian-Zadeh

2010-01-01

394

Malignant transformation of an iris melanocytoma  

Microsoft Academic Search

A 34-year-old Caucasian woman was diagnosed as having a pigmented iris tumor showing recent growth and satellite lesions. The tumor was associated with pigmentation of the anterior chamber angle and secondary unilateral glaucoma. After local excision, histopathologic studies revealed the plump polyhedral cells typical of melanocytoma. However, the examination of additional sections showed evidence of malignancy. The diagnosis of a

Arnaldo P. Cialdini; Jose A. Sahel; Alex E. Jalkh; John J. Weiter; Kamal Zakka; Daniel M. Albert

1989-01-01

395

Iris Murdoch, Liberal Education and Human Flourishing  

ERIC Educational Resources Information Center

Articulating the good of liberal education--what we should teach and why we should teach it--is necessary to resist the subversion of liberal education to economic or political ends and the mania for measurable skills. I argue that Iris Murdoch's philosophical writings enrich the work of contemporary Aristotelians, such as Joseph Dunne and…

Evans, William

2009-01-01

396

A Report on IRI Scoring and Interpretation.  

ERIC Educational Resources Information Center

Noting that most classroom teachers use informal reading inventories (IRI) as diagnostic instruments, a study examined what oral reading accuracy level is most appropriate for the instructional level and whether repetitions should be counted as oral reading errors. Randomly selected students from the second through fifth grades at two elementary…

Anderson, Betty

397

Thermal hydraulic limits analysis for the MIT Research Reactor low enrichment uranium core conversion using statistical propagation of parametric uncertainties  

E-print Network

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 ...

Chiang, Keng-Yen

2012-01-01

398

Design of a low enrichment, enhanced fast flux core for the Massachusetts Institute of Technology Research Reactor  

E-print Network

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 ...

Ellis, Tyler Shawn

2009-01-01

399

Self-actuated nuclear reactor shutdown system using induction pump to facilitate sensing of core coolant temperature  

DOEpatents

A self-actuated shutdown system incorporated into a reactivity control assembly in a nuclear reactor includes pumping means for creating an auxiliary downward flow of a portion of the heated coolant exiting from the fuel assemblies disposed adjacent to the control assembly. The shutdown system includes a hollow tubular member which extends through the outlet of the control assembly top nozzle so as to define an outer annular flow channel through the top nozzle outlet separate from an inner flow channel for primary coolant flow through the control assembly. Also, a latching mechanism is disposed in an inner duct of the control assembly and is operable for holding absorber bundles in a raised position in the control assembly and for releasing them to drop them into the core of the reactor for shutdown purposes. The latching mechanism has an inner flow passage extending between and in flow communication with the absorber bundles and the inner flow channel of the top nozzle for accommodating primary coolant flow upwardly through the control assembly. Also, an outer flow passage separate from the inner flow passage extends through the latching mechanism between and in flow communication with the inner duct and the outer flow channel of the top nozzle for accommodating inflow of a portion of the heated coolant from the adjacent fuel assemblies. The latching mechanism contains a magnetic material sensitive to temperature and operable to cause mating or latching together of the components of the latching mechanism when the temperature sensed is below a known temperature and unmating or unlatching thereof when the temperature sensed is above a given temperature. The temperature sensitive magnetic material is positioned in communication with the heated coolant flow through the outer flow passage for directly sensing the temperature thereof. Finally, the pumping means includes a jet induction pump nozzle and diffuser disposed adjacent the bottom nozzle of the control assembly and in flow communication with the inlet thereof. The pump nozzle is operable to create an upward driving flow of primary coolant through the pump diffuser and then to the absorber bundles. The upward driving flow of primary coolant, in turn, creates a suction head within the outer flow channel of the top nozzle and thereby an auxiliary downward flow of the heated coolant portion exiting from the upper end of the adjacent fuel assemblies through the outer flow channel to the pump nozzle via the outer flow passage of the latching mechanism and an annular space between the outer and inner spaced ducts of the control assembly housing. The temperature of the heated coolant exiting from the adjacent fuel assemblies can thereby be sensed directly by the temperature sensitive magnetic material in the latching mechanism.

Sievers, Robert K. (N. Huntingdon, PA); Cooper, Martin H. (Churchill, PA); Tupper, Robert B. (Greensburg, PA)

1987-01-01

400

Vapor core propulsion reactors  

NASA Technical Reports Server (NTRS)

Many research issues were addressed. For example, it became obvious that uranium tetrafluoride (UF4) is a most preferred fuel over uranium hexafluoride (UF6). UF4 has a very attractive vaporization point (1 atm at 1800 K). Materials compatible with UF4 were looked at, like tungsten, molybdenum, rhenium, carbon. It was found that in the molten state, UF4 and uranium attacked most everything, but in the vapor state they are not that bad. Compatible materials were identified for both the liquid and vapor states. A series of analyses were established to determine how the cavity should be designed. A series of experiments were performed to determine the properties of the fluid, including enhancement of the electrical conductivity of the system. CFD's and experimental programs are available that deal with most of the major issues.

Diaz, Nils J.

1991-01-01

401

Coaxial optical structure for iris recognition from a distance  

NASA Astrophysics Data System (ADS)

Supporting an unconstrained user interface is an important issue in iris recognition. Various methods try to remove the constraint of the iris being placed close to the camera, including portal-based and pan-tilt-zoom (PTZ)-based solutions. Generally speaking, a PTZ-based system has two cameras: one scene camera and one iris camera. The scene camera detects the eye's location and passes this information to the iris camera. The iris camera captures a high-resolution image of the person's iris. Existing PTZ-based systems are divided into separate types and parallel types, according to how the scene camera and iris camera combine. This paper proposes a novel PTZ-based iris recognition system, in which the iris camera and the scene camera are combined in a coaxial optical structure. The two cameras are placed together orthogonally and a cold mirror is inserted between them, such that the optical axes of the two cameras become coincident. Due to the coaxial optical structure, the proposed system does not need the optical axis displacement-related compensation required in parallel-type systems. Experimental results show that the coaxial type can acquire an iris image more quickly and accurately than a parallel type when the stand-off distance is between 1.0 and 1.5 m.

Jung, Ho Gi; Jo, Hyun Su; Park, Kang Ryoung; Kim, Jaihie

2011-05-01

402

Determination of total serum insulin (IRI) in insulin-treated diabetic patients  

Microsoft Academic Search

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

Lise G. Heding

1972-01-01

403

An Iris Recognition System to Enhance E-security Environment Based on Wavelet Theory  

Microsoft Academic Search

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

Jafar M. H. Ali; Aboul Ella

404

Robust Encoding of Local Ordinal Measures: A General Framework of Iris Recognition  

Microsoft Academic Search

The randomness of iris pattern makes it one of the most reliable bio- metric traits. On the other hand, the complex iris image structure and various sources of intra-class variations result in the difficulty of iris representation. Al- though diverse iris recognition methods have been proposed, the fundamentals of iris recognition have not a unified answer. As a breakthrough of

Zhenan Sun; Tieniu Tan; Yunhong Wang

2004-01-01

405

An effective and fast iris recognition system based on a combined multiscale feature extraction technique  

Microsoft Academic Search

The randomness of iris pattern makes it one of the most reliable biometric traits. On the other hand, the complex iris image structure and the various sources of intra-class variations result in the difficulty of iris representation. Although, a number of iris recognition methods have been proposed, it has been found that several accurate iris recognition algorithms use multiscale techniques,

Makram Nabti; Ahmed Bouridane

2008-01-01

406

[Cultivation of Iris ensata Thunb. callus tissue].  

PubMed

A continuous callus culture was obtained from zygotic embryos of Japanese iris (Iris ensata Thunb.) on the Murashige-Skoog medium supplemented with 2 mg/l alpha-naphthylacetic acid and 0.5 mg/l 6-benzylaminopurine (BAP). It was found that a successful callusogenesis required isolated embryos at the wax stage of endosperm development. The optimal combination of phytohormones for the growth of callus tissue was 1 mg/l 2,4-dichlorophenoxyacetic acid and 0.5 mg/l BAP. The pigment composition of I. ensata callus tissue was studied. It was demonstrated that subcultivated callus tissue contained red pigments of flavonoid nature. Under stress cultivation conditions, yellow pigments were formed and the content of red pigments increased. PMID:15125204

Boltenkov, E V; Rybin, V G; Zarembo, E V

2004-01-01

407

IRIS: Videos on Plate Tectonics and Earthquakes  

NSDL National Science Digital Library

This is a collection of short informational videos on dynamic Earth processes, developed to teach how earthquakes happen and why they are studied. The videos explore tectonic plate motion, elastic rebound, fault models, types of boundaries, locating earthquake epicenters, seismic wave paths, and more. This resource is part of IRIS, the Incorporated Research Institutions for Seismology, a consortium of international laboratories and data collection centers.

2011-03-18

408

Pigment Melanin: Pattern for Iris Recognition  

Microsoft Academic Search

Recognition of iris based on Visible Light (VL) imaging is a difficult\\u000aproblem because of the light reflection from the cornea. Nonetheless, pigment\\u000amelanin provides a rich feature source in VL, unavailable in Near-Infrared\\u000a(NIR) imaging. This is due to biological spectroscopy of eumelanin, a chemical\\u000anot stimulated in NIR. In this case, a plausible solution to observe such\\u000apatterns

Mahdi S. Hosseini; Babak Nadjar Araabi; Hamid Soltanian-Zadeh

2009-01-01

409

Pigmented free-floating iris cysts  

Microsoft Academic Search

Free-floating iris cysts are rare. These cysts may be located in the vitreous or the anterior chamber. Anterior chamber cysts can be idiopathic or induced by trauma or surgery. Vitreous cysts may be associated with the remnants of the hyaloid system and therefore be congenital, or can result from trauma or ocular disease.Case 1: An 8-year-old girl presented for routine

Gurdeep Singh; Kalpana Narendran; Veerappan R Saravanan; V Narendran

2007-01-01

410

Optimal wavelength band clustering for multispectral iris recognition.  

PubMed

This work explores the possibility of clustering spectral wavelengths based on the maximum dissimilarity of iris textures. The eventual goal is to determine how many bands of spectral wavelengths will be enough for iris multispectral fusion and to find these bands that will provide higher performance of iris multispectral recognition. A multispectral acquisition system was first designed for imaging the iris at narrow spectral bands in the range of 420 to 940 nm. Next, a set of 60 human iris images that correspond to the right and left eyes of 30 different subjects were acquired for an analysis. Finally, we determined that 3 clusters were enough to represent the 10 feature bands of spectral wavelengths using the agglomerative clustering based on two-dimensional principal component analysis. The experimental results suggest (1) the number, center, and composition of clusters of spectral wavelengths and (2) the higher performance of iris multispectral recognition based on a three wavelengths-bands fusion. PMID:22772098

Gong, Yazhuo; Zhang, David; Shi, Pengfei; Yan, Jingqi

2012-07-01

411

The Interface Region Imaging Spectrograph (IRIS)  

E-print Network

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...

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

412

Progressive growth of an iris melanocytoma in a child  

Microsoft Academic Search

PURPOSE: To document progressive growth of a benign iris melanocytoma.METHODS: Interventional case report. A 9-year-old male underwent removal of a pigmented iris tumor that had been documented photographically to double in size.RESULTS: Histopathologic sections revealed a deeply pigmented mass with cytologic features typical of a melanocytoma.CONCLUSION: Although iris melanocytoma is generally a stationary lesion, it can show progressive growth. Such

Jerry A Shields; Ralph C Eagle; Carol L Shields; Leonard B Nelson

2002-01-01

413

Spatial Light Modulator Would Serve As Electronic Iris  

NASA Technical Reports Server (NTRS)

In proposed technique for controlling brightness of image formed by lens, spatial light modulator serves as segmented, electronically variable aperture. Offers several advantages: spatial light modulator controlled remotely and responds faster than motorized iris or other remotely controlled mechanical iris. Unlike iris, modulator also configured so as not to vary depth of field appreciably. Unlike lead lanthanum zirconate titanate crystal, spatial light modulator does not require high voltage.

Gutow, David A.

1991-01-01

414

Non-ideal iris segmentation using graph cuts  

Microsoft Academic Search

A non-ideal iris segmentation approach using graph cuts is presented. Unlike many existing algorithms for iris localization which extensively utilize eye geometry, the proposed approach is predominantly based on image intensities. In a step-wise procedure, first eyelashes are segmented from the input images using image texture, then the iris is segmented using grayscale information, followed by a post-processing step that

Shrinivas J. Pundlik; Damon L. Woodard; Stanley T. Birchfield

2008-01-01

415

HyspIRI Low Latency Concept and Benchmarks  

NASA Technical Reports Server (NTRS)

Topics include HyspIRI low latency data ops concept, HyspIRI data flow, ongoing efforts, experiment with Web Coverage Processing Service (WCPS) approach to injecting new algorithms into SensorWeb, low fidelity HyspIRI IPM testbed, compute cloud testbed, open cloud testbed environment, Global Lambda Integrated Facility (GLIF) and OCC collaboration with Starlight, delay tolerant network (DTN) protocol benchmarking, and EO-1 configuration for preliminary DTN prototype.

Mandl, Dan

2010-01-01

416

Iris recognition as a biometric method after cataract surgery  

Microsoft Academic Search

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

Roberto Roizenblatt; Paulo Schor; Fabio Dante; Jaime Roizenblatt; Rubens Belfort Jr

2004-01-01

417

Iris Recognition System Using Wavelet Packet and Support Vector Machines  

Microsoft Academic Search

\\u000a In this paper, iris recognition system using wavelet packet and support vector machines is presented. It specifically uses\\u000a the multiresolution decomposition of 2-D discrete wavelet packet transform for extracting the unique features from the acquired\\u000a iris image. This method of feature extraction is well suited to describe the shape of the iris while allowing the algorithm\\u000a to be translation and

Byungjun Son; Gyundo Kee; Yillbyung Lee

2003-01-01

418

A Concept of Passive Safety, Pressurized Water Reactor System with Inherent Matching Nature of Core Heat Generation and Heat Removal  

Microsoft Academic Search

The reduction of manpower in operation and maintenance by simplification of the system are essential to improve the safety and the economy of future light water reactors. At the Japan Atomic Energy Research Institute (JAERI), a concept of a simplified passive safety reactor system JPSR was developed for this purpose and in the concept minimization of developing work and conservation

Yoshio MURAO; Fumimasa ARAYA; Takamichi IWAMURA; Keisuke OKUMURA

1995-01-01

419

The Importance of Being Random: Statistical Principles of Iris Recognition  

NSDL National Science Digital Library

Professor John Daugman of the University of Cambridge Computer Laboratory is the author of this paper on iris recognition. It examines the characteristics of the human iris from a statistical perspective in order to estimate the requirements for accurate identification. Many complex issues of pattern recognition are addressed, such as the problems of isolating the iris and maintaining accuracy regardless of the eye's position. Professor Daugman's home page has numerous other research papers, as well as a general introduction and overviews of basic iris recognition concepts.

Daugman, John.

2001-01-01

420

Limbus Impact on Off-angle Iris Degradation  

SciTech Connect

The accuracy of iris recognition depends on the quality of data capture and is negatively affected by several factors such as angle, occlusion, and dilation. Off-angle iris recognition is a new research focus in biometrics that tries to address several issues including corneal refraction, complex 3D iris texture, and blur. In this paper, we present an additional significant challenge that degrades the performance of the off-angle iris recognition systems, called the limbus effect . The limbus is the region at the border of the cornea where the cornea joins the sclera. The limbus is a semitransparent tissue that occludes a side portion of the iris plane. The amount of occluded iris texture on the side nearest the camera increases as the image acquisition angle increases. Without considering the role of the limbus effect, it is difficult to design an accurate off-angle iris recognition system. To the best of our knowledge, this is the first work that investigates the limbus effect in detail from a biometrics perspective. Based on results from real images and simulated experiments with real iris texture, the limbus effect increases the hamming distance score between frontal and off-angle iris images ranging from 0.05 to 0.2 depending upon the limbus height.

Karakaya, Mahmut [ORNL; Barstow, Del R [ORNL; Santos-Villalobos, Hector J [ORNL; Thompson, Joseph W [ORNL; Bolme, David S [ORNL; Boehnen, Chris Bensing [ORNL

2013-01-01

421

Modeling and design of a new core-moderator assembly and neutron beam ports for the Penn State Breazeale Nuclear Reactor (PSBR)  

NASA Astrophysics Data System (ADS)

This study is for modeling and designing a new reactor core-moderator assembly and new neutron beam ports that aimed to expand utilization of a new beam hall of the Penn State Breazeale Reactor (PSBR). The PSBR is a part of the Radiation Science and Engineering Facility (RSEC) and is a TRIGA MARK III type research reactor with a movable core placed in a large pool and is capable to produce 1MW output. This reactor is a pool-type reactor with pulsing capability up to 2000 MW for 10-20 msec. There are seven beam ports currently installed to the reactor. The PSBR's existing core design limits the experimental capability of the facility, as only two of the seven available neutron beam ports are usable. The finalized design features an optimized result in light of the data obtained from neutronic and thermal-hydraulics analyses as well as geometrical constraints. A new core-moderator assembly was introduced to overcome the limitations of the existing PSBR design, specifically maximizing number of available neutron beam ports and mitigating the hydrogen gamma contamination of the neutron beam channeled in the beam ports. A crescent-shaped moderator is favored in the new PSBR design since it enables simultaneous use of five new neutron beam ports in the facility. Furthermore, the crescent shape sanctions a coupling of the core and moderator, which reduces the hydrogen gamma contamination significantly in the new beam ports. A coupled MURE and MCNP5 code optimization analysis was performed to calculate the optimum design parameters for the new PSBR. Thermal-hydraulics analysis of the new design was achieved using ANSYS Fluent CFD code. In the current form, the PSBR is cooled by natural convection of the pool water. The driving force for the natural circulation of the fluid is the heat generation within the fuel rods. The convective heat data was generated at the reactor's different operating powers by using TRIGSIMS, the fuel management code of the PSBR core. In the CFD modeling, the amount of heat generated by the fuel is assumed to be transferred totally into the coolant. Therefore, the surface heat flux is applied to the fuel cladding outer surface by considering the depleted fuel composition of each individual fuel rod under a reference core loading condition defined as; 53H at 1MW full power. In order to model the entire PSBR reactor, fine mesh discretization was achieved with 22 millions structured and unstructured computational meshes. The conductive heat transfer inside the fuel rods was ignored in order to decrease the computational mesh requirement. Since the PSBR core operates in the subcooled nucleate boiling region, the CFD simulation of new PSBR design was completed utilizing an Eulerian-Eulerian multiphase flow formulation and RPI wall boiling model. The simulation results showed that the new moderator tank geometry results in secondary flow entering into the core due to decrease in the cross-flow area. Notably, the radial flow improves the local heat transfer conditions by providing radial-mixing in the core. Bubble nucleation occurs on the heated fuel rods but bubbles are collapsing in the subcooled fluid. Furthermore, the bulk fluid properties are not affected by the bubble formation. Yet, subcooled boiling enhances the heat transfer on the fuel rods. Five neutron beam ports are designed for the new reactor. The geometrical configuration, filter and collimator system designs of each neutron beam ports are selected based on the requirements of the experimental facilities. A cold neutron beam port which utilizes cold neutrons from three curved guide tubes is considered. Therefore, there will be seven neutron beams available in the new facility. The neutronic analyses of the new beam port designs were achieved by using MCNP5 code and Burned Coupled Simulation Tool for the PSBR. The MCNP simulation results showed that thermal neutron flux was increased by a factor of minimum 1.23 times and maximum 2.68 times in the new beam port compared to the existing BP4 design. Besides total gamma dose was decreased by a factor

Ucar, Dundar

422

Verification of maximum radial power peaking factor due to insertion of FPM-LEU target in the core of RSG-GAS reactor  

NASA Astrophysics Data System (ADS)

Verification of Maximum Radial Power Peaking Factor due to insertion of FPM-LEU target in the core of RSG-GAS Reactor. Radial Power Peaking Factor in RSG-GAS Reactor is a very important parameter for the safety of RSG-GAS reactor during operation. Data of radial power peaking factor due to the insertion of Fission Product Molybdenum with Low Enriched Uranium (FPM-LEU) was reported by PRSG to BAPETEN through the Safety Analysis Report RSG-GAS for FPM-LEU target irradiation. In order to support the evaluation of the Safety Analysis Report incorporated in the submission, the assessment unit of BAPETEN is carrying out independent assessment in order to verify safety related parameters in the SAR including neutronic aspect. The work includes verification to the maximum radial power peaking factor change due to the insertion of FPM-LEU target in RSG-GAS Reactor by computational method using MCNP5and ORIGEN2. From the results of calculations, the new maximum value of the radial power peaking factor due to the insertion of FPM-LEU target is 1.27. The results of calculations in this study showed a smaller value than 1.4 the limit allowed in the SAR.

Setyawan, Daddy; Rohman, Budi

2014-09-01

423

Three Mile Island Unit-2 core status summary: a basis for tool development for reactor disassembly and defueling  

SciTech Connect

The accident at Three Mile Island Unit-2 (TMI-2) on March 28, 1979 caused extensive damage to the core. A variety of analyses were performed using three general approaches to determine the extent of core damage. First, thermal-hydraulic events were reconstructed using available data, thermal-hydraulic principles, and computer analyses. Second, determinations of the hydrogen generated yielded estimates of the amount of zircaloy oxidized and embrittled. Third, the type and quantity of fission products released during the accident were used to estimate the location of core damage and the fuel temperatures which were achieved. Uncertainties exist in each type of determination due to the equivocal nature of the data. This paper reviews and summarizes the core damage assessments which have been made, identifies the minimum and maximum bounds of damage, and establishes a reference description for the current status of the core.

Croucher, D.W.

1981-05-01

424

Evaluation of the thermal-hydraulic operating limits of the HEU-LEU transition cores for the MIT Research Reactor  

E-print Network

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 ...

Wang, Yunzhi (Yunzhi Diana)

2009-01-01

425

Estimate of radiation release from MIT reactor with low enriched uranium (LEU) core during maximum hypothetical accident  

E-print Network

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 ...

Plumer, Kevin E. (Kevin Edward)

2011-01-01

426

TITAN : an advanced three dimensional coupled neutronicthermal-hydraulics code for light water nuclear reactor core analysis  

E-print Network

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 ...

Griggs, D. P.

1984-01-01

427

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

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.

Parish, T.A.

1995-03-02

428

Results of a Neutronic Simulation of HTR-Proteus Core 4.2 using PEBBED and other INL Reactor Physics Tools: FY-09 Report  

SciTech Connect

The Idaho National Laboratory’s deterministic neutronics analysis codes and methods were applied to the computation of the core multiplication factor of the HTR-Proteus pebble bed reactor critical facility. A combination of unit cell calculations (COMBINE-PEBDAN), 1-D discrete ordinates transport (SCAMP), and nodal diffusion calculations (PEBBED) were employed to yield keff and flux profiles. Preliminary results indicate that these tools, as currently configured and used, do not yield satisfactory estimates of keff. If control rods are not modeled, these methods can deliver much better agreement with experimental core eigenvalues which suggests that development efforts should focus on modeling control rod and other absorber regions. Under some assumptions and in 1D subcore analyses, diffusion theory agrees well with transport. This suggests that developments in specific areas can produce a viable core simulation approach. Some corrections have been identified and can be further developed, specifically: treatment of the upper void region, treatment of inter-pebble streaming, and explicit (multiscale) transport modeling of TRISO fuel particles as a first step in cross section generation. Until corrections are made that yield better agreement with experiment, conclusions from core design and burnup analyses should be regarded as qualitative and not benchmark quality.

Hans D. Gougar

2009-08-01

429

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)

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.

Bachrata, A.; Fichot, F.; Quintard, M.; Repetto, G.; Fleurot, J.

2012-05-01

430

Non-local equilibrium two-phase flow model with phase change in porous media and its application to reflooding of a severely damaged reactor core  

SciTech Connect

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 {sup d}ebris bed{sup .} 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 deg. C or 700 deg. 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.

Bachrata, A.; Fichot, F.; Quintard, M.; Repetto, G.; Fleurot, J. [Institut de Radioprotection et de Surete Nucleaire, Cadarache (France); Universite de Toulouse (France); INPT, UPS (France); IMFT - Institut de Mecanique des Fluides de Toulouse, Allee Camille Soula, F-31400 Toulouse (France) and CNRS (France); IMFT, F-31400 Toulouse (France); Institut de Radioprotection et de Surete Nucleaire, Cadarache (France)

2012-05-15

431

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

E-print Network

Oak Ridge National Laboratory OTOC Once-Through-and-Out Cycle P&T Partitioning and Transmutation PWR Pressurized Water Reactor SNF Spent Nuclear Fuel THTR Thorium High Temperature Reactor TRISO Tri-structural Isotropic TRU Transuranium Nuclide... and thermal spectrum ?.. 77 17 Key focus of Chapter IV ???????????????????? 86 18 Fuel assembly block?????????????????????.... 88 19 Fuel assembly block dimensions????????????????? 89 20 TRISO fuel structure?????????????????????.. 92 21...

Alajo, Ayodeji Babatunde

2011-08-08

432

IRIS TOXICOLOGICAL REVIEW OF METHYL ETHYL KETONE [AND IRIS SUMMARY] (EXTERNAL REVIEW DRAFT)  

EPA Science Inventory

The U.S. EPA has conducted an external peer review of the scientific basis supporting the health hazard and dose-response assessment for methyl ethyl ketone that will appear on the Agency's online data base, the Integrated Risk Information System (IRIS). Peer review is meant to ...

433

Iris on the Move: Acquisition of Images for Iris Recognition in Less Constrained Environments  

Microsoft Academic Search

Iris recognition is one of the most powerful techniques for biometric identification ever developed. Commercial systems based on the algorithms developed by John Daugman have been available since 1995 and have been used in a variety of practical applications. However, all currently available systems impose substantial constraints on subject position and motion during the recognition process. These constraints are largely

J. R. Matey; O. Naroditsky; K. Hanna; R. Kolczynski; D. J. LoIacono; S. Mangru; M. Tinker; T. M. Zappia; W. Y. Zhao

2006-01-01

434

The cellular and molecular biology of the iris, an overlooked tissue: the iris and pseudoexfoliation glaucoma.  

PubMed

Located between the cornea and the lens, the Iris is fully immersed in aqueous humor. During exfoliation syndrome, a disorder of the elastic fibers, an abnormal fibrillar material (XFM) is deposited on the anterior lens capsule underneath the pigment epithelium of the Iris. Release of this material to the aqueous humor reaches the trabecular meshwork where its presence is associated with elevated intraocular pressure. Ultrastructural studies suggest that the XFM material is produced by the lens capsule, lens epithelial and iris pigment epithelial cells (IPE). The involvement of the IPE in pseudoexfoliation glaucoma has not been extensively addressed. Immunohistochemistry studies have shown higher levels of LOXL1 and clusterin in the IPE extracellular space of specimens from exfoliation patients. But studies using IPE cells to understand the formation of the XFM in vitro and/or in vivo are scarce. A focus on the Iris and its IPE cells would be key for the elucidation of XFM and the understanding of the development of pseudoexfoliation glaucoma. PMID:25275904

Borrás, Terete

2014-01-01

435

An efficient algorithm in extracting human iris Morphological features  

Microsoft Academic Search

The interface of computer technologies and biology is having a huge impact on society. Human recognition research projects promises new life to many security-consulting firms and personal identification system manufacturers. Iris recognition is considered to be the most reliable biometric authentication system. Very few iris recognition algorithms were commercialized. The method proposed in this paper differed from the existing work

Mohamed A. Mohamed; M. E. A. Abou-Elsoud; M. M. Eid

2009-01-01

436

Onboard Instrument Processing Concepts for the HyspIRI Mission  

E-print Network

/ice applications. Keywords imaging spectroscopy, hyperspectral imaging, onboard processing, data reduction I. INTRODUCTION HyspIRI [HyspIRI] is a proposed mission to carry a VSWIR hyperspectral instrument with 220 bands-sampling (a large eruption may cover 100 pixels compared to 1.5x106 pixels/s acquired). Onboard algorithms

Schaffer, Steven

437

Pigmented Iris Malignant Melanomas in New Zealand White Rabbits  

Microsoft Academic Search

Harding-Passey mouse melanoma, maintained in tissue culture for 3 months, was injected into the anterior chamber of anesthetized New Zealand white rabbits. The rabbits were immunosuppressed with depot steroids and developed pigmented iris lesions seen with biomicroscopy by days 6–8. The lesions continued to grow for as long as 29 days, at which time most of the iris was replaced

Norman T. Felberg; Joseph B. Michelson

1979-01-01

438

Utility of a Biopsy in Suspicious Pigmented Iris Tumors  

Microsoft Academic Search

Background: In the presence of pigmented iris lesions evocative of malignant melanoma and implying oncological treatment, a foregoing biopsy to exclude a benign lesion may seem a reasonable approach. After examining patient files, the utility of such a diagnostic approach was explored. Material and Methods: Retrospective, consecutive histopathologic case series of 10 pigmented iris tumor specimens excised since 1993. Histopathologic

Ann Schalenbourg; Sylvie Uffer; Leonidas Zografos

2008-01-01

439

Mechanism and Clinical Significance of Prostaglandin-Induced Iris Pigmentation  

Microsoft Academic Search

The new glaucoma drugs latanoprost, isopropyl unoprostone, travoprost, and bimatoprost cause increased pigmentation of the iris in some patients. The purpose of the present article is to survey the available preclinical and clinical data on prostaglandin-induced iris pigmentation and to assess the phenomenon from a clinical perspective. Most of the data have been obtained with latanoprost, and it appears that

Johan W Stjernschantz; Daniel M Albert; Dan-Ning Hu; Filippo Drago; Per J Wistrand

2002-01-01

440

Iris colour in passerine birds: why be bright-eyed?  

Microsoft Academic Search

An initial survey of iris coloration in passerine birds (Aves: Passeriformes) showed that a brightly pigmented iris is much more common in southern African and Australian birds than in those from Europe, temperate North America, and Venezuela. However, the only statistical correlation reflected the distribution of particular bird families in these regions. Ten family-level groups considered to represent monophyletic taxa

A. J. F. K. Craig; P. E. Hulley

2004-01-01

441

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. ...

442

Iris-based biometric recognition using dyadic wavelet transform  

Microsoft Academic Search

A biometric identification system, based on the processing of the human iris by the dyadic wavelet transform, has been introduced. The procedure to obtain an iris signature of 256 bits has been described, as well as the feature extraction block and the verification system. The results have shown that the system can achieve high rates of security.

C. Sanchez-Avila; R. Sanchez-Reillo; D. de Martin-Roche

2002-01-01

443

Efficient iris recognition by characterizing key local variations  

Microsoft Academic Search

Unlike other biometrics such as fingerprints and face, the distinct aspect of iris comes from randomly distributed fea- tures. This leads to its high reliability for personal identification, and at the same time, the difficulty in effectively representing such details in an image. This paper describes an efficient algorithm for iris recognition by characterizing key local variations. The basic idea

L. Ma; Tieniu Tan; Yunhong Wang; Dexin Zhang

2004-01-01

444

A machine-vision system for iris recognition  

Microsoft Academic Search

This paper describes a prototype system for personnel verification based on automated iris recognition. The motivation for this endevour stems from the observation that the human iris provides a particularly interesting structure on which to base a technology for noninvasive biometric measurement. In particular, it is known in the biomedical community that irises are as distinct as fingerprints or patterns

Richard P. Wildes; Jane C. Asmuth; Gilbert L. Green; Steven C. Hsu; Raymond J. Kolczynski; James R. Matey; Sterling E. Mcbride

1996-01-01

445

Iris recognition with enhanced depth-of-field image acquistion  

Microsoft Academic Search

Automated iris recognition is a promising method for noninvasive verification of identity. Although it is noninvasive, the procedure requires considerable cooperation from the user. In typical acquisition systems, the subject must carefully position the head laterally to make sure that the captured iris falls within the field-of-view of the digital image acquisition system. Furthermore, the need for sufficient energy at

Joseph van der Gracht; V. P. Pauca; Harsha Setty; Ramkumar Narayanswamy; Robert J. Plemmons; Sudhakar Prasad; Todd Torgersen

2004-01-01

446

Comparison of Compression Algorithms' Impact on Iris Recognition Accuracy  

Microsoft Academic Search

The impact of using dierent lossy compression algorithms on the matching accuracy of iris recognition systems is investigated. In particular, we relate rate-distortion performance as measured in PSNR to the matching scores as obtained by a concrete recognition system. JPEG2000 and SPIHT are correctly predicted by PSNR to be well suited compression algorithms to be employed in iris recognition systems.

Stefan Matschitsch; Martin Tschinder; Andreas Uhl

2007-01-01

447

Human eye iris recognition using the mutual information  

Microsoft Academic Search

This article presents the new biometric electro-optical measuring method supported by PC for identification of a person by its eye iris image recognition. The aim of this approach is to show the ability of mutual information to such recognition. Couples of the comparative human iris images were geometrically aligned by maximization of their mutual information and subsequently recognized. The mutual

M. Dobes; L. Machala; P. Tichavsky; J. Pospisil

2004-01-01