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1

Optimum Strategy for Ex-Core Dosimeters/monitors in the IRIS Reactor  

NASA Astrophysics Data System (ADS)

International Reactor Innovative and Secure (IRIS) is a medium-power (~300 MWe) advanced light water reactor that features an integral primary system configuration to enhance safety. Steam generators are located inside the pressure vessel above the core, forming a thick (~1.68 m) annular region, that extends into an equally thick downcomer surrounding the core. As a result, neutron fluence at the pressure vessel and in the cavity is reduced by 5-6 orders of magnitude relative to present loop-type Pressurized Water Reactors (PWRs). Reduction of the RPV fluence eliminates embrittlement concerns, but introduces new challenges for the ex-core flux monitors. This paper proposes using advanced flux monitors, such as SiC semiconductor neutron detectors, and examines their optimum placement in the downcomer region. Furthermore, the requirements on neutron dosimetry/monitors considered for the IRIS-reactor are common to Generation-IV Integral Primary System Reactors (IPSRs).

Petrovi?, Bojan; Ruddy, Frank H.; Lombardi, Carlo

2003-06-01

2

First Core and Refueling Options for IRIS  

SciTech Connect

The International Reactor Innovative and Secure (IRIS) is being developed by an international consortium of industry, laboratory, university and utility establishments, led by Westinghouse. The IRIS design addresses key requirements associated with advanced reactors, including improved safety, enhanced proliferation resistance, competitive electricity production cost, and improved waste management. IRIS is a modular, small/medium size (100 to 335 MWe) PWR with integral vessel configuration. Its design is based on proven LWR technology, so that no new technology development is needed and near term deployment is possible. At the same time, aim was to introduce improvements as compared to present PWRs. These opposing requirements resulted in an evolutionary approach to fuel and core design, balancing new features and the need to avoid extensive testing and demonstration programs. A path forward was devised by selecting the current fuel technology for the first IRIS core, but keeping future upgrades possible through the variable moderation fuel assembly design. This paper describes this approach and discusses core fueling options that enable achieving four-year and eight-year core lifetime. (authors)

Petrovic, Bojan; Carelli, Mario D. [Westinghouse Electric Company (United States); Greenspan, Ehud; Milosevic, Miodrag; Vujic, Jasmina [Univ. California Berkeley, Berkeley CA 94720 (United States); Padovani, Enrico; Ganda, Francesco [Politecnico di Milano, Piazza Leonardo da Vinci, 32, 20133 Milano (Italy)

2002-07-01

3

Preliminary Safety Analysis for the IRIS Reactor  

Microsoft Academic Search

A deterministic analysis of the IRIS safety features has been carried out by means of the best-estimate code RELAP (ver. RELAP5 mod3.2). First, the main system components were modeled and tested separately, namely: the Reactor Pressure Vessel (RPV), the modular helical-coil Steam Generators (SG) and the Passive (natural circulation) Emergency Heat Removal System (PEHRS). Then, a preliminary set of accident

M. E. Ricotti; A. Cammi; A. Cioncolini; C. Lombardi; A. Cipollaro; F. Orioto; L. E. Conway; A. C. Barroso

2002-01-01

4

Iris small break loca phenomena identification and ranking table (PIRT)  

Microsoft Academic Search

The international reactor innovative and secure (IRIS) is a modular pressurized water reactor with an integral configuration (all primary system components – reactor core, internals, pumps, steam generators, pressurizer, and control rod drive mechanisms – are inside the reactor vessel). The IRIS plant conceptual design was completed in 2001 and the preliminary design is currently underway. The pre-application licensing process

T. K. Larson; F. J. Moody; G. E. Wilson; W. L. Brown; C. Frepoli; J. Hartz; B. G. Woods; L. Oriani

2007-01-01

5

NEUTRONIC REACTOR CORE INSTRUMENT  

DOEpatents

A multi-purpose instrument for measuring neutron flux, coolant flow rate, and coolant temperature in a nuclear reactor is described. The device consists essentially of a hollow thimble containing a heat conducting element protruding from the inner wall, the element containing on its innermost end an amount of fissionsble materinl to function as a heat source when subjected to neutron flux irradiation. Thermocouple type temperature sensing means are placed on the heat conducting element adjacent the fissionable material and at a point spaced therefrom, and at a point on the thimble which is in contact with the coolant fluid. The temperature differentials measured between the thermocouples are determinative of the neutron flux, coolant flow, and temperature being measured. The device may be utilized as a probe or may be incorporated in a reactor core. (AE C)

Mims, L.S.

1961-08-22

6

NEUTRONIC REACTOR CORE  

DOEpatents

An improved core for a gas-cooled power reactor which admits gas coolant at high temperatures while affording strong integral supporting structure and efficient moderation of neutrons is described. The multiplicities of fuel elements constituting the critical amassment of fissionable material are supported and confined by a matrix of metallic structure which is interspersed therebetween. Thermal insulation is interposed between substantially all of the metallic matrix and the fuel elements; the insulation then defines the principal conduit system for conducting the coolant gas in heat-transfer relationship with the fuel elements. The metallic matrix itseif comprises a system of ducts through which an externally-cooled hydrogeneous liquid, such as water, is circulated to serve as the principal neutron moderant for the core and conjointly as the principal coolant for the insulated metallic structure. In this way, use of substantially neutron transparent metals, such as aluminum, becomes possible for the supporting structure, despite the high temperatures of the proximate gas. The Aircraft Nuclear Propulsion program's "R-1" reactor design is a preferred embodiment.

Thomson, W.B.; Corbin, A. Jr.

1961-07-18

7

INPRO economic assessment of the IRIS nuclear reactor for deployment in Brazil  

Microsoft Academic Search

This paper presents the results of the economic assessment of the International Reactor Innovative and Secure (IRIS) for deployment in Brazil using the assessment methodology developed under the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO), co-ordinated by the International Atomic Energy Agency (IAEA). INPRO initiated in 2001 and has the main objective of helping to ensure that

Orlando João Agostinho Gonçalves Filho

2011-01-01

8

Lateral restraint assembly for reactor core  

Microsoft Academic Search

In a nuclear reactor including a reactor vessel defining a shielded core cavity having a reactor core extending vertically along a longitudinal axis and being located internally of the cavity, the reactor core is described which has layers of reflector blocks defining an outer peripheral surface for the core spaced from the vessel and being supported in a manner permitting

W. Gorholt; R. K. Luci

1986-01-01

9

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.

Cooke, Franklin E. (San Jose, CA)

1992-01-01

10

IRIS Reactor a Suitable Option to Provide Energy and Water Desalination for the Mexican Northwest Region  

SciTech Connect

The Northwest region of Mexico has a deficit of potable water, along this necessity is the region growth, which requires of additional energy capacity. The IRIS reactor offers a very suitable source of energy given its modular size of 300 MWe and it can be coupled with a desalination plant to provide the potable water for human consumption, agriculture and industry. The present paper assess the water and energy requirements for the Northwest region of Mexico and how the deployment of the IRIS reactor can satisfy those necessities. The possible sites for deployment of Nuclear Reactors are considered given the seismic constraints and the closeness of the sea for external cooling. And in the other hand, the size of the desalination plant and the type of desalination process are assessed accordingly with the water deficit of the region.

Alonso, G.; Ramirez, R.; Gomez, C.; Viais, J.

2004-10-03

11

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

SciTech Connect

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

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

2004-10-06

12

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

13

Upgrade of the Annular Core Pulse Reactor  

Microsoft Academic Search

The Annular Core Pulse Reactor (ACPR) is a TRIGA type reactor which has been in operation at Sandia Laboratories since 1967. The reactor is utilized in a wide variety of experimental programs which include radiation effects, neutron radiography, activation analysis, and fast reactor safety. During the past two years, the ACPR has become an important experimental facility for the United

Reuscher

1976-01-01

14

IRIS Simplified LERF Model  

SciTech Connect

Westinghouse is currently conducting the pre-application licensing of the International Reactor Innovative and Secure (IRIS). One of the key aspects of the IRIS design is its safety-by-designTM philosophy and within this framework the PRA is being used as an integral part of the design process. The most ambitious risk-related goal for IRIS is to reduce the Emergency Planning Zone (EPZ) to within the exclusion area by demonstrating that the off-site doses are consistent with the US Protective Action Guidelines (PAGs) for initiation of emergency response so that the required protective actions would be limited to the exclusion area. As a first step, a model has been developed to provide a first order approximation of the Large Early Release Frequency (LERF) as a surrogate predictor of the off-site doses. A key-aspect of the LERF model development is the characterization of the possible paths of release. Four main categories have been historically pointed out: (1) Core Damage (CD ) sequences with containment bypass, (2) CD sequences with containment isolation failure, (3) CD sequences with containment failure at low pressure and (4) CD sequences with containment failure at high pressure. They have been reevaluated to account for the IRIS design features.

Maioli, A.; Finnicum, D.J.; Kumagai, Y.

2004-10-06

15

Wire core reactor for nuclear thermal propulsion  

Microsoft Academic Search

Studies have been performed of a compact high-performance nuclear rocket reactor that incorporates a tungsten alloy wire fuel element. This reactor, termed the wire core reactor, can deliver a specific impulse of 1,000 s using an expander cycle and a nozzle expansion ratio of 500 to 1. The core is constructed of layers of 0.8-mm-dia fueled tungsten wires wound over

Richard B. Harty; Robert G. Brengle

1993-01-01

16

Wire core reactor for nuclear thermal propulsion  

NASA Astrophysics Data System (ADS)

Studies have been performed of a compact high-performance nuclear rocket reactor that incorporates a tungsten alloy wire fuel element. This reactor, termed the wire core reactor, can deliver a specific impulse of 1,000 s using an expander cycle and a nozzle expansion ratio of 500 to 1. The core is constructed of layers of 0.8-mm-dia fueled tungsten wires wound over alternate layers of spacer wires, which forms a rugged annular lattice. Hydrogen flow in the core is annular, flowing from inside to outside. In addition to the concepts compact size and good heat transfer, the core has excellent power-flow matching features and can resist vibration and thermal stresses during star-up and shutdown.

Harty, Richard B.; Brengle, Robert G.

1993-01-01

17

Seismic analysis of nuclear reactor core  

Microsoft Academic Search

A 3-D homogenization model is proposed to study the overall dynamic behaviour of a nuclear reactor core under seismic excitation. Finite element method is employed for detailed analysis. As the fluid–structure interaction in the core gives rise to a non-conservative force, the mass and stiffness matrices of the system are non-symmetrical. Both Wilson-? direct integration method and QZ method were

R. J. Zhang; W. Q. Wang; S. H. Hou; C. K. Chan

2001-01-01

18

Lifetime embrittlement of reactor core materials.  

National Technical Information Service (NTIS)

Over a core lifetime, the reactor materials Zircaloy-2, Zircaloy-4, and hafnium may become embrittled due to the absorption of corrosion- generated hydrogen and to neutron irradiation damage. Results are presented on the effects of fast fluence on the fra...

P. H. Kreyns W. F. Bourgeois P. L. Charpentier B. F. Kammenzind D. G. Franklin

1994-01-01

19

MODULAR CORE UNITS FOR A NEUTRONIC REACTOR  

DOEpatents

A modular core unit for use in a nuclear reactor is described. Many identical core modules can be placed next to each other to make up a complete core. Such a module includes a cylinder of moderator material surrounding a fuel- containing re-entrant coolant channel. The re-entrant channel provides for the circulation of coolant such as liquid sodium from one end of the core unit, through the fuel region, and back out through the same end as it entered. Thermal insulation surrounds the moderator exterior wall inducing heat to travel inwardly to the coolant channel. Spaces between units may be used to accommodate control rods and support structure, which may be cooled by a secondary gas coolant, independently of the main coolant. (AEC)

Gage, J.F. Jr.; Sherer, D.B.

1964-04-01

20

The Economics of IRIS  

Microsoft Academic Search

IRIS (International Reactor Innovative and Secure) is a small to medium advanced light water cooled modular reactor being developed by an international consortium led by Westinghouse\\/BNFL. This reactor design is specifically aimed at utilities looking to install new (or replacement) nuclear capacity to match market demands, or at developing countries for their distributed power needs. To determine the optimal configuration

K. Miller; D. Paramonov

2002-01-01

21

A reactor core barrel lift at Byron  

SciTech Connect

A core barrel lift operation for weld inspection at Byron Nuclear Power Plant Unit 1 is summarized in the article. The lift of the lower reactor internals was accomplished with robots, cameras, and 2 overhead craned operators. Lift time and return time were 35 minutes each, and the total dose received for the job was 87 person-millirem. Similar operations at Wolf Creek and Zion are also very briefly described.

Michal, R.

1996-10-01

22

Neutron radiography facility annular core pulse reactor  

Microsoft Academic Search

Neutron radiography capabilities which are an integral part of the Annular Core Pulse Reactor (ACPR) experiment facilities have been extended and improved. Neutron radiography provides an important non-destructive testing technique. Ordinary x- or gamma-radiography are not entirely satisfactory for the non-destructive testing of such items as explosives, valve seals, sealing materials, and electrical potting compounds. The non-linearity in neutron absorption

B. F. Estes; J. S. Philbin; F. V. Thome

1976-01-01

23

A vectorized heat transfer model for solid reactor cores  

Microsoft Academic Search

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

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

1990-01-01

24

A vectorized heat transfer model for solid reactor cores  

Microsoft Academic Search

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 (HTGR). In addition, nuclear thermal rockets share a number of similarities with terrestrial HTGRs and would be amendable to similar types of analyses. In these reactors, the heat transfer in the solid core

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

1990-01-01

25

NEUTRONIC REACTOR OPERATIONAL METHOD AND CORE SYSTEM  

DOEpatents

Homogeneous neutronic reactor systems are described wherein an aqueous fuel solution is continuously circulated through a spherical core tank. The pumped fuel solution-is injected tangentially into the hollow spherical interior, thereby maintaining vigorous rotation of the solution within the tank in the form of a vortex; gaseous radiolytic decomposition products concentrate within the axial vortex cavity. The evolved gas is continuously discharged through a gas- outlet port registering with an extremity of the vortex cavity. and the solution stream is discharged through an annular liquid outlet port concentrically encircling the gas outlet by virtue of which the vortex and its cavity are maintained precisely axially aligned with the gas outlet. A primary heat exchanger extracts useful heat from the hot effluent fuel solution before its recirculation into the core tank. Hollow cylinders and other alternative core- tank configurations defining geometric volumes of revolution about a principal axis are also covered. AEC's Homogeneous Reactor Experiment No. 1 is a preferred embodiment.

Winters, C.E.; Graham, C.B.; Culver, J.S.; Wilson, R.H.

1960-07-19

26

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

27

Nuclear design analysis of a multicavity gas core reactor system  

Microsoft Academic Search

The Innovative Nuclear Space Power Institute (INSPI) at the University of Florida has undertaken extensive research to establish the scientific feasibility and engineering validation of gaseous core reactor and energy conversion systems that have core power densities of a kilowatt per cubic centimeter and reactor masses of a kilogram per thermal megawatt for burstpower space applications. Gaseous core fission concepts

M. M. Panicker; E. T. Dugan; S. Anghaie

1987-01-01

28

Post Impact Behavior of Mobile Reactor Core Containment Systems.  

National Technical Information Service (NTIS)

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 mate...

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

1972-01-01

29

Plateau iris.  

PubMed

The term plateau iris was first coined in 1958 to describe the iris configuration of a patient. Two years later the concept of plateau iris was published. In 1977, the plateau iris configuration was classically defined as presurgical changes of an eye with a relative normal central anterior chamber depth, flat iris by conventional biomicroscopy, but displaying an extremely narrow or closed angle on gonioscopic examination. On the other hand, the plateau iris syndrome was defined as an acute glaucoma crisis in one eye with a relative normal central anterior chamber depth and patent iridotomy on direct examination, presenting angle closure confirmed by gonioscopic examination after mydriasis. In 1992, the anatomic aspects of plateau iris were studied using ultrasound biomicroscopy. Finally, plateau iris has been considered an anatomic variant of iris structure in which the iris periphery angulates sharply forward from its insertion point and then again angulates sharply and centrally backward, along with an anterior positioning of the ciliary processes seen on ultrasound biomicroscopy. The clinical treatment of plateau iris syndrome is carried out with topical use of pilocarpine. However, the definitive treatment should be fulfilled by performing an argon laser peripheral iridoplasty. PMID:19039479

Diniz Filho, Alberto; Cronemberger, Sebastião; Mérula, Rafael Vidal; Calixto, Nassim

30

IRIS: Proceeding Towards the Preliminary Design  

SciTech Connect

The IRIS (International Reactor Innovative and Secure) project has completed the conceptual design phase and is moving towards completion of the preliminary design, scheduled for the end of 2002. Several other papers presented in this conference provide details on major aspects of the IRIS design. The three most innovative features which uniquely characterize IRIS are, in descending order of impact: 1. Safety-by-design, which takes maximum advantage of the integral configuration to eliminate from consideration some accidents, greatly lessen the consequence of other accident scenarios and decrease their probability of occurring; 2. Optimized maintenance, where the interval between maintenance shutdowns is extended to 48 months; and 3. Long core life, of at least four years without shuffling or partial refueling. Regarding feature 1, design and analyses will be supplemented by an extensive testing campaign to verify and demonstrate the performance of the integral components, individually as well as interactive systems. Test planning is being initiated. Test results will be factored into PRA analyses under an overall risk informed regulation approach, which is planned to be used in the IRIS licensing. Pre-application activities with NRC are also scheduled to start in mid 2002. Regarding feature 2, effort is being focused on advanced online diagnostics for the integral components, first of all the steam generators, which are the most critical component; several techniques are being investigated. Finally, a four year long life core design is well underway and some of the IRIS team members are examining higher enrichment, eight to ten year life cores which could be considered for reloads. (authors)

Carelli, M. [Westinghouse Electric Company (United States); Miller, K. [BNFL UK (United Kingdom); Lombardi, C. [Politecnico di Milano, Piazza Leonardo da Vinci, 32, 20133 Milano (Italy); Todreas, N. [Massachusetts Institute of Technology, 77 massachusetts avenue, cambridge, ma 02139-4307 (United States); Greenspan, E. [Univ. California Berkeley, Berkeley CA 94720 (United States); Ninokata, H. [Tokyo Institute of Technology, 2-12-1 O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Lopez, F. [Bechtel Power Corp (United States); Cinotti, L. [Ansaldo Nuclear Division, C.so Perrone, 25, Genova 16161 (Italy); Collado, J. [Equipos Nucleares SA - ENSA (Spain); Oriolo, F. [Universita di Pisa, Dipartimento di Ingegneria Meccanica Nucleare e della Produzione, Via Diotisalvi 2, 56126 Pisa (Italy); Alonso, G. [Instituto Nacional de Investigaciones Nucleares, Km 36.5 Carretera Mexico-Toluca, Ocoyoacac 52045, Edo. de Mexico (Mexico); Morales, M. [NUCLEP, Itaguai (Brazil); Boroughs, R. [Tennessee Valley Authority - TVA (United States); Barroso, A. [CNEN, Comissao Nacional de Energia Nuclear, Rua General Severiano 90, Rio de Janeiro, RJ-22-294-900 (Brazil); Ingersoll, D. [Oak Ridge National Laboratory - ORNL, 1 Bethel Valley Rd, Oak Ridge, TN 37830 (United States); Cavlina, N. [Faculty of Electrical Engineering and Computing, University of Zagreb, Unska 3, HR-10000 Zagreb (Croatia)

2002-07-01

31

IRIS: Proceeding Towards the Preliminary Design  

Microsoft Academic Search

The IRIS (International Reactor Innovative and Secure) project has completed the conceptual design phase and is moving towards completion of the preliminary design, scheduled for the end of 2002. Several other papers presented in this conference provide details on major aspects of the IRIS design. The three most innovative features which uniquely characterize IRIS are, in descending order of impact:

M. Carelli; K. Miller; C. Lombardi; N. Todreas; E. Greenspan; H. Ninokata; F. Lopez; L. Cinotti; J. Collado; F. Oriolo; G. Alonso; M. Morales; R. Boroughs; A. Barroso; D. Ingersoll; N. Cavlina

2002-01-01

32

Specific Mass Estimates for A Vapor Core Reactor With MHD  

Microsoft Academic Search

This study investigated the development of a system concept for space power generation and nuclear electric propulsion based on a vapor core reactor (VCR) with magnetohydrodynamic (MHD) power conversion system, coupled to a magnetoplasma-dynamic (MPD) thruster. The VCR is a liquid-vapor core reactor concept operating with metallic uranium or uranium tetrafluoride (UFâ) vapor as the fissioning fuel and alkali metals

Travis Knight; Blair Smith; Samim Anghaie

2002-01-01

33

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

34

Lifetime embrittlement of reactor core materials  

SciTech Connect

Over a core lifetime, the reactor materials Zircaloy-2, Zircaloy-4, and hafnium may become embrittled due to the absorption of corrosion- generated hydrogen and to neutron irradiation damage. Results are presented on the effects of fast fluence on the fracture toughness of wrought Zircaloy-2, Zircaloy-4, and hafnium; Zircaloy-4 to hafnium butt welds; and hydrogen precharged beta treated and weld metal Zircaloy-4 for fluences up to a maximum of approximately 150 x 10{sup 24} n/M{sup 2} (> 1 Mev). While Zircaloy-4 did not exhibit a decrement in K{sub IC} due to irradiation, hafnium and butt welds between hafnium and Zircaloy-4 are susceptible to embrittlement with irradiation. The embrittlement can be attributed to irradiation strengthening, which promotes cleavage fracture in hafnium and hafnium-Zircaloy welds, and, in part, to the lower chemical potential of hydrogen in Zircaloy-4 compared to hafnium, which causes hydrogen, over time, to drift from the hafnium end toward the Zircaloy-4 end and to precipitate at the interface between the weld and base-metal interface. Neutron radiation apparently affects the fracture toughness of Zircaloy-2, Zircaloy-4, and hafnium in different ways. Possible explanations for these differences are suggested. It was found that Zircaloy-4 is preferred over Zircaloy-2 in hafnium-to- Zircaloy butt-weld applications due to its absence of a radiation- induced reduction in K{sub IC} plus its lower hydrogen absorption characteristics compared with Zircaloy-2.

Kreyns, P.H..; Bourgeois, W.F.; Charpentier, P.L.; Kammenzind, B.F.; Franklin, D.G. [Bettis Atomic Power Lab., West Mifflin, PA (United States); White, C.J. [Knolls Atomic Power Lab., Schenectady, NY (United States)

1994-08-01

35

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; Holcomb, David Eugene [ORNL; Varma, Venugopal Koikal [ORNL

2012-01-01

36

Steam Generator of the International Reactor Innovative and Secure  

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

Cinotti, L.; Bruzzone, M.; Meda, N.; Corsini, G. [Ansaldo Nucleare S.p.A., c.so Perrone, 25 - 16161 - Genova (Italy); Lombardi, C.V.; Ricotti, M. [Politecnico di Milano, Piazza Leonardo da Vinci, 32, 20133 Milano (Italy); Conway, L.E. [Westinghouse Electric Company (United States)

2002-07-01

37

Intrinsically secure fast reactors with dense cores  

Microsoft Academic Search

Secure safety, resistance to weapons material proliferation and problems of long-lived wastes remain the most important “painful points” of nuclear power.Many innovative reactor concepts have been developed aimed at a radical enhancement of safety. The promising potential of innovative nuclear reactors allows for shifting accents in current reactor safety “strategy” to reveal this worth. Such strategy is elaborated focusing on

Igor Slessarev

2007-01-01

38

Fission Rate Measurements in Fuel Plate Type Assembly Reactor Cores.  

National Technical Information Service (NTIS)

The methods, materials and equipment have been developed to allow extensive and precise measurement of fission rate distributions in water moderated, U-Al fuel plate assembly type reactor cores. Fission rate monitors are accurately positioned in the react...

J. W. Rogers

1988-01-01

39

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

40

Design of the Clinch River Breeder Reactor Plant heterogeneous core  

SciTech Connect

The original core design for the Clinch River Breeder Reactor Plant (CRBRP) was a homogeneous core, as are essentially all of present day liquid metal fast breeder reactors. A study was undertaken to determine if a rearrangement of the core into a heterogeneous configuration, with fertile elements interspersed within the fueled zone, would improve the breeding ratio significantly without excessive adverse effects on other aspects of the design. The result of the study was that the heterogeneous concept improved not only the breeding ratio and doubling time, but also the control assembly worth, core restraint response, and the fuel cycle cost. This paper describes the design evolution and the major effects of the change from a homogeneous to a heterogeneous core in the Clinch River Breeder Reactor Plant.

Dickson, P.W. Jr.; Arnold, W.H.

1982-01-01

41

IRIS Responsiveness to Generation IV Road-map Goals  

SciTech Connect

The DOE Generation IV road-map process is in its second and final year. Almost one hundred concepts submitted from all over the world have been reviewed against the Generation IV goals of resources sustainability; safety and reliability; and, economics. Advanced LWRs are taken as the reference point. IRIS (International Reactor Innovative and Secure), a 100-335 MWe integral light water reactor being developed by a vast international consortium led by Westinghouse, is one on the concepts being considered in the road-map and is perhaps the most visible representative of the concept set known as Integral Primary System Reactors (IPSR). This paper presents how IRIS satisfies the prescribed goals. The first goal of resource sustainability includes criteria like utilization of fuel resources, amount and toxicity of waste produced, environmental impact, proliferation and sabotage resistance. As a thermal reactor IRIS does not have the same fuel utilization as fast reactors. However, it has a significant flexibility in fuel cycles as it is designed to utilize either UO{sub 2} or MOX with straight burn cycles of 4 to 10 years, depending on the fissile content. High discharge burnup and Pu recycling result in good fuel utilization and lower waste; IRIS has also attractive proliferation resistance characteristics, due to the reduced accessibility of the fuel. The safety and reliability goal include reliability, workers' exposure, robust safety features, models with well characterized uncertainty, source term and mechanisms of energy release, robust mitigation of accidents. IRIS is significantly better than advanced LWRs because of its safety by design which eliminates a variety of accidents such as LOCAs, its containment vessel coupled design which maintains the core safely covered during the accident sequences, its design simplification features such as no (or reduced) soluble boron, internal shielding and four-year refueling/maintenance interval which significantly reduce workers' exposure. IRIS has indeed a superb safety which makes it an excellent candidate to fulfill the Generation IV goal of no off-site emergency response. Finally, the economic goal includes various factors contributing to cost of electricity and capital at risk. Significant uncertainties exist on capital cost of all Generation IV concepts and a consensus is not reached how well advantages of modularity, simplicity and standardized, multiple fabrication compare with economies of scale. Still, IRIS is expected to have attractive economics because of its modular, simplified design which can be constructed in a three-year period and its small-to-medium size which significantly reduces the financial risk. In summary, IRIS is positively responsive to all Generation IV goals and is an excellent candidate for further development by its large international consortium. (authors)

Carelli, M.D.; Paramonov, D.V.; Petrovic, B. [Westinghouse Electric Co., 1344 Beulah Road, Pittsburgh, PA 15235 (United States)

2002-07-01

42

MCNP full-core modeling of the advanced test reactor  

Microsoft Academic Search

A full-core Monte Carlo neutron and photon (MCNP) transport model has been completed for the advanced test reactor (ATR) at Idaho National Engineering Laboratory. This new model is a complete three-dimensional model that represents fuel elements, core structures, and target regions in adequate detail. The model can be used in evaluating heating and reaction rates in various target regions of

S. S. Kim; S. N. Jahshan; R. B. Nielson

1993-01-01

43

Investigation of the Core Melt Accident in Light Water Reactors.  

National Technical Information Service (NTIS)

In the thesis the core melt accident, heating up and collapsing of the reactor core were investigated. The most important parameters of influence were found and their effect on the development of the accident were shown. A causal diagram was developed rep...

H. Koerber

1980-01-01

44

Cooling of core debris within the reactor vessel lower head  

Microsoft Academic Search

Under severe accident conditions, the most crucial action for recovery from the accident state is to cool the core debris and prevent or terminate attack on the remaining fission product barriers. One means of preventing attack on the containment structures is to retain the core debris within the reactor vessel. Some accident situations could result in the transport of molten

R. E. Henry; J. P. Burelbach; R. J. Hammersley; C. E. Henry; G. T. Klopp

1993-01-01

45

Cooling of core debris within the reactor vessel lower head  

Microsoft Academic Search

Under severe-accident conditions, the most crucial action for recovery from the accident state is to cool the core debris and prevent or terminate attack on the remaining fission product barriers. One means of preventing attack on the containment structures is to retain the core debris within the reactor vessel. The Three Mile Island Unit 2 (TMI-2) accident demonstrated that this

R. E. Henry; J. P. Burelbach; R. J. Hammersley; C. E. Henry; G. T. Klopp

1991-01-01

46

Seismic response of block-type nuclear reactor core. [HTGR  

Microsoft Academic Search

An analytical model is described that was developed to predict seismic response of large gas-cooled reactor cores. The model is used to investigate scaling laws involved in the design of physical models of such cores, and to make parameter studies.

J. G. Bennett; R. C. Dove; J. L. Merson

1977-01-01

47

Pancake core high conversion light water reactor concept  

Microsoft Academic Search

A new concept is proposed for a high conversion light water reactor (HCLWR) that achieves both high conversion and high burnup while maintaining a negative void reactivity coefficient. This HCLWR has a flat pancake core with thick axial blankets. By using the flat core, a potential problem of HCLWRs, the positive void reactivity coefficient can be reduced by neutron leakage,

Y. Ishiguro; K. Okumura

1989-01-01

48

Development and validation of a fast reactor core burnup code – FARCOB  

Microsoft Academic Search

A large fast breeder reactor requires the accurate estimation of power produced in different parts of the reactor core and blanket during any operating condition for a safe and economic operation through out reactor life time. A fast reactor core simulation code FARCOB based on multigroup diffusion theory has been developed in IGCAR for core simulation of PFBR reactor under

P. Mohanakrishnan

2008-01-01

49

COMSORS: A light water reactor chemical core catcher  

SciTech Connect

The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate lightwater reactor (LWR) core-melt accidents and ensure containment integrity. A special dissolution glass made of lead oxide (PbO) and boron oxide (B{sub 2}O{sub 3}) is placed under the reactor vessel. If molten core debris is released onto the glass, the following sequence happens: (1) the glass absorbs decay heat as its temperature increases and the glass softens; (2) the core debris dissolves into the molten glass; (3) molten glass convective currents create a homogeneous high-level waste (HLW) glass; (4) the molten glass spreads into a wider pool, distributing the heat for removal by radiation to the reactor cavity above or transfer to water on top of the molten glass; and (5) the glass solidifies as increased surface cooling area and decreasing radioactive decay heat generation allows heat removal to exceed heat generation.

Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W. [Oak Ridge National Lab., TN (United States). Chemical Technology Div.; Kenton, M.A. [Creare Inc., Hanover, NH (United States)

1997-02-24

50

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

Microsoft Academic Search

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

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

2003-01-01

51

Core length testable reactor concept neutronic analysis  

SciTech Connect

Development work on thermionic reactor systems has been ongoing in the US since the early 1950s. While significant successes were achieved, progress has been hampered by frequent changes in direction and funding instabilities (as has been true for many high technology initiatives). The recent Air Force thermionics initiative (1991) represents the latest in thermionics reactor development in the US. This Air Force initiative called for the development of thermionics reactors with the output power of about 40 kWe, and which incorporated the features of testability, fabricability, low development cost, high level of safety and reliability, and survivability. Several concepts were analyzed to define a design that would meet all the requirements set forth by the Air Force. This report describes the methodology used, the different designs analyzed and reasons for the evolution of the design, and presents the results for the different concepts.

Hanan, N.A.; Bhattacharyya, S.K.

1992-09-01

52

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

53

Uranium droplet nuclear reactor core with MHD generator  

NASA Astrophysics Data System (ADS)

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

Anghaie, Samim; Kumar, Ratan

54

Modification of the Core Cooling System of TRIGA 2000 Reactor  

SciTech Connect

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

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

2010-06-22

55

Modification of the Core Cooling System of TRIGA 2000 Reactor  

NASA Astrophysics Data System (ADS)

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

Umar, Efrizon; Fiantini, Rosalina

2010-06-01

56

Investigation of activity release during light water reactor core meltdown  

Microsoft Academic Search

A test facility was developed for the determination of activity release and of aerosol characteristics under realistic light water reactor core melting conditions. It is composed of a high-frequency induction furnace, a ThOâ crucible system, and a collection apparatus consisting of membrane and particulate filters. Thirty-gram samples of a representative core material mixture (corium) were melted under air, argon, or

H. Albrecht; V. Matschoss; H. Wild

1978-01-01

57

An economic optimization of pressurized light water reactor cores  

NASA Astrophysics Data System (ADS)

Two reactor cores (1000 MWe and 600 MWe) are optimized with respect to power cost. The power cost is minimized while retaining the thermal-hydraulic margins of the reference core. Constant thermal-hydraulic margins result in similar accident thermal-hydraulic transient behavior of the cores developed during the optimization study. The cost components impacted by the optimization are once-through fuel cycle, capital, and administrative/manpower costs. The variables in the optimization are pin diameter, moderator to fuel (H/U) ratio, core length, and the number of fuel pins in the core. A sequential quadratic programming approach is employed to solve the nonlinear optimization problem with constraints. The fuel cycle costs are evaluated by the use of the linear reactivity model, and capital costs are adjusted by suitable modifications to the nuclear energy cost database reference costs. The results of the analysis shows that for fixed assembly parameters (i.e., pin diameter, H/U ratio, and core length), the optimum core is one that operates at the thermal-hydraulic limits. Cores optimized with unconstrained assembly characteristics contain a larger number of smaller pins at a higher H/U ratio. This follows the trend in current reactor designs. While the lifetime power cost savings for the optimized core are less than 4 million dollars (versus a present day total cost of 6.9 billion dollars), the optimization analysis shows that higher thermal-hydraulic margins can be attained with minimum power cost increases. With increased emphasis on reactor safety, significantly higher safety margins may therefore be achieved without a significant power cost increase. The optimized configurations were found to be relatively insensitive to fuel cycle cost component variations.

Pfeifer, Holger

58

Two stochastic optimization algorithms applied to nuclear reactor core design  

Microsoft Academic Search

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

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

2006-01-01

59

Structural homogenized analysis for a nuclear reactor core  

Microsoft Academic Search

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

R. J. Zhang

1998-01-01

60

Gas core reactor (neutronics)-theoretical modeling and experimental verification  

Microsoft Academic Search

The development of a sound scientific data base that includes key information in the areas of neutronics, thermophysical properties, and materials for cyclic gaseous core reactors has been the objective of a lengthy theoretical\\/experimental research program at the University of Florida. The most recently completed phase of this program includes theoretical neutronics modeling and experimental verification. Static and dynamic neutronic

E. T. Dugan; E. E. Carroll; N. J. Diaz; H. M. Forehand

1985-01-01

61

Device for Supporting the Core of a Fast Neutron Reactor.  

National Technical Information Service (NTIS)

The object of the invention is a supporting device for the core of a fast reactor cooled by a molten sodium circulation comprising a tank, a flooring mounted to this tank, a grid equipped with studs for the entry of the fuel elements, the grid which is su...

J. Lleres J. P. Martin M. Perona R. Venot

1977-01-01

62

Gamma thermometer based reactor core liquid level detector  

SciTech Connect

A system is provided which employs a modified gamma thermometer for determining the liquid coolant level within a nuclear reactor core. The gamma thermometer which normally is employed to monitor local core heat generation rate (reactor power), is midified by thermocouple junctions and leads to obtain an unambiguous indication of the presence or absence of coolant liquid at the gamma thermometer location. A signal processor generates a signal based on the thermometer surface heat transfer coefficient by comparing the signals from the thermocouples at the thermometer location. The generated signal is a direct indication of loss of coolant due to the change in surface heat transfer when coolant liquid drops below the thermometer location. The loss of coolant indication is independent of reactor power at the thermometer location. Further, the same thermometer may still be used for the normal power monitoring function.

Burns, T.J.

1983-09-20

63

Support arrangements for core modules of nuclear reactors. [PWR  

DOEpatents

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

Bollinger, L.R.

1983-11-03

64

Support arrangement for core modules of nuclear reactors  

DOEpatents

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

Bollinger, Lawrence R. (Schenectady, NY)

1987-01-01

65

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

66

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

67

Thermal-Hydraulics in Uncovered Core of Light Water Reactor in Severe Core Damage Accident, (II)  

Microsoft Academic Search

The SEFDAN is a computer program to analyze the one-dimensional thermal-hydraulics of a partially uncovered core of a light water reactor in a severe degraded-cooling event. In order to verify the code and to obtain better understanding of the severe core damage process, SEFDAN has been applied to analyses of the thermal response of fuel rods in the Power Burst

Fumiya TANABE; Ken MURAMATSU; Tohru SUDA

1986-01-01

68

DABIE: A Data Banking System of Integral Experiments for Reactor Core Characteristics Computer Codes. User's Manual.  

National Technical Information Service (NTIS)

A data banking system of integral experiments for reactor core characteristics computer codes, DABIE, has been developed to lighten the burden on searching so many documents to obtain experiment data required for verification of reactor core characteristi...

K. Matsumoto Y. Naito S. Ohkubo H. Aoyanagi

1987-01-01

69

CFD analysis of core melt spreading on the reactor cavity floor using ANSYS CFX code  

Microsoft Academic Search

In the very unlikely event of a severe reactor accident involving core melt and reactor pressure vessel failure, it is important to provide an accident management strategy that would allow the molten core material to cool down, resolidify and bring the core debris to a coolable state for Light Water Reactors (LWRs). One approach to achieve a coolable state is

Wan-Sik Yeon; Kwang-Hyun Bang; Youngjo Choi; Yong Soo Kim; Jaegon Lee

70

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

SciTech Connect

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

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

2010-12-23

71

Pancake core high conversion light water reactor concept  

SciTech Connect

A new concept is proposed for a high conversion light water reactor (HCLWR) that achieves both high conversion and high burnup while maintaining a negative void reactivity coefficient. This HCLWR has a flat pancake core with thick axial blankets. By using the flat core, a potential problem of HCLWRs, the positive void reactivity coefficient can be reduced by neutron leakage, and a fuel assembly of very tight lattice pitch can be used. The leakage neutrons are utilized in the axial blankets to enhance the conversion ratio.

Ishiguro, Y.; Okumura, K.

1989-03-01

72

The Economics of IRIS  

SciTech Connect

IRIS (International Reactor Innovative and Secure) is a small to medium advanced light water cooled modular reactor being developed by an international consortium led by Westinghouse/BNFL. This reactor design is specifically aimed at utilities looking to install new (or replacement) nuclear capacity to match market demands, or at developing countries for their distributed power needs. To determine the optimal configuration for IRIS, analysis was undertaken to establish Generation Costs ($/MWh) and Internal Rate of Return (IRR %) to the Utility at alternative power ratings. This was then combined with global market projections for electricity demand out to 2030, segmented into key geographical regions. Finally this information is brought together to form insights, conclusions and recommendations regarding the optimal design. The resultant analysis reveals a single module sized at 335 MWe, with a construction period of 3 years and a 60-year plant life. Individual modules can be installed in a staggered fashion (3 equivalent to 1005 MWe) or built in pairs (2 sets of twin units' equivalent to 1340 MWe). Uncertainty in Market Clearing Price for electricity, Annual Operating Costs and Construction Costs primarily influence lifetime Net Present Values (NPV) and hence IRR % for Utilities. Generation Costs in addition are also influenced by Fuel Costs, Plant Output, Plant Availability and Plant Capacity Factor. Therefore for a site based on 3 single modules, located in North America, Generations Costs of 28.5 $/MWh are required to achieve an IRR of 20%, a level which enables IRIS to compete with all other forms of electricity production. Plant size is critical to commercial success. Sustained (lifetime) high factors for Plant Output, Availability and Capacity Factor are required to achieve a competitive advantage. Modularity offers Utilities the option to match their investments with market conditions, adding additional capacity as and when the circumstances are right. Construction schedule needs to be controlled. There is a clear trade-off between reducing financing charges and optimising revenue streams. (authors)

Miller, K. [British Nuclear Fuels - BNFL (United Kingdom); Paramonov, D. [Westinghouse Electric Company (United States)

2002-07-01

73

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

74

A reactor core on-line monitoring program - COMP  

SciTech Connect

A program named COMP is developed for on-line monitoring PWRs' in-core power distribution in this paper. Harmonics expansion method is used in COMP. The Unit 1 reactor of Daya Bay Nuclear Power Plant (Daya Bay NPP) in China is considered for verification. The numerical results show that the maximum relative error between measurement and reconstruction results from COMP is less than 5%, and the computing time is short, indicating that COMP is capable for online monitoring PWRs. (authors)

Wang, C. [State Nuclear Power Software Development Center, Beijing, 100029 (China); School of Nuclear Science and Technology, Xi'an Jiaotong Univ. (China); Building 1, Compound No.29, North Third Ring Road, Xicheng District, Beijing, 100029 (China); Wu, H.; Cao, L. [School of Nuclear Science and Technology, Xi'an Jiaotong Univ. (China)

2012-07-01

75

100KE REACTOR CORE REMOVAL PROJECT ALTERNATIVE ANALYSIS WORKSHOP REPORT  

Microsoft Academic Search

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

HARRINGTON RA

2010-01-01

76

Uranium droplet nuclear reactor core with MHD generator  

Microsoft Academic Search

An innovative concept employing liquid uranium droplets as fuel in an ultrahigh-temperature vapor core reactor (UTVR) magnetohydrodynamic (MHD) generator power system for space power generation has been studied. Metallic vapor in superheated form acts as a working fluid for a closed-Rankine-type thermodynamic cycle. Usage of fuel and working fluid in this form assures certain advantages. The major technical issues emerging

Samim Anghaie; Ratan Kumar

1991-01-01

77

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

78

Piezoelectric material for use in a nuclear reactor core  

SciTech Connect

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

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

2012-05-17

79

Piezoelectric material for use in a nuclear reactor core  

NASA Astrophysics Data System (ADS)

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

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

2012-05-01

80

In-reactor testing of the closed cycle gas core reactor: The Nuclear Light Bulb concept  

Microsoft Academic Search

The Nuclear Light Bulb (NLB) concept is an advanced closed cycle space propulsion rocket engine design that offers unprecidented performance characteristics in terms of specific impulse (greater than 1800 s) and thrust (greater than 445 kN). The NLB is a gas-core nuclear reactor making use of thermal radiation from a high temperature U-plasma core to heat the hydrogen propellant to

R. O. Gauntt; S. A. Slutz; G. A. Harms; T. S. Latham; W. C. Roman; R. J. Rodgers

1992-01-01

81

Development of an automated core model for nuclear reactors  

SciTech Connect

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 used as input to the TRAC transient-analysis code for subsequent safety analysis of the reactor at any point in its operating lifetime. The basic capability to perform steady-state reactor-core analysis already existed in the combination of the HELIOS lattice-physics code and the NESTLE advanced nodal code. In this project, the automated package was completed by (1) obtaining cross-section libraries for HELIOS, (2) validating HELIOS by comparing its predictions to results from critical experiments and from the MCNP Monte Carlo code, (3) validating NESTLE by comparing its predictions to results from numerical benchmarks and to measured data from operating reactors, and (4) developing a linkage code to transform HELIOS output into NESTLE input.

Mosteller, R.D.

1998-12-31

82

Thermal radiation in gas core nuclear reactors for space propulsion  

NASA Astrophysics Data System (ADS)

A diffusive model of the radial transport of thermal radiation out of a cylindrical core of fissioning plasma is presented. The diffusion approximation is appropriate because the opacity of uranium is very high at the temperatures of interest (greater than 3000 K). We make one additional simplification of assuming constant opacity throughout the fuel. This allows the complete set of solutions to be expressed as a single function. This function is approximated analytically to facilitate parametric studies of the performance of a test module of the nuclear light bulb gas-core nuclear-rocket-engine concept, in the Annular Core Research Reactor at Sandia National Laboratories. Our findings indicate that radiation temperatures in range of 4000-6000 K are attainable, which is sufficient to test the high specific impulse potential (approximately 2000 s) of this concept.

Slutz, Stephen A.; Gauntt, Randall O.; Harms, Gary A.; Latham, Thomas; Roman, Ward; Rodgers, Richard J.

1994-05-01

83

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

84

Thermal-Hydraulics in Uncovered Core of Light Water Reactor in Severe Core Damage Accident, (IV)  

Microsoft Academic Search

Analysis of the TMI-2 core damage behavior has been performed with the SEFDAN code. The scope of the analysis is by the time of restarting the reactor coolant pump RCP-2B at 2 h 54 min into the accident. The analysis indicates that fuel temperature would have reached the melting point of UO2 in the upper-most part of the most central

Fumiya TANABE; Tohru SUDA

1987-01-01

85

RECENT DEVELOPMENTS OF THE IRIS PROJECT OF INTEREST FOR LATIN AMERICA  

SciTech Connect

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

Carelli, M.D.; Petrovic, B.

2004-10-03

86

IRIS Final Technical Progress Report  

SciTech Connect

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

M. D. Carelli

2003-11-03

87

A solid reactor core thermal model for nuclear thermal rockets  

NASA Astrophysics Data System (ADS)

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.

Rider, William J.; Cappiello, Michael W.; Liles, Dennis R.

1991-01-01

88

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

Microsoft Academic Search

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

Mitsuru KAMBE

2002-01-01

89

Performance evaluation\\/analysis of Pakistan Research Reactor1 (PARR1) current core configuration  

Microsoft Academic Search

Neutronic and thermal hydraulic analyses have been carried out for current core of Pakistan Research Reactor-1 (PARR-1). Comparison was made between calculated and measured key neutronic parameters. Reactor core parameters important for reactor operation and safety have been calculated. Calculated neutronic parameters include: excess reactivity, shut down margin, control rod worth, peak power density location, criticality position, peaking factors, neutron

Tayyab Mahmood; Ishtiaq Hussain Bokhari; Masood Iqbal; Tariq Mahmood; Naseer Ahmed; Muhammad Israr

2011-01-01

90

Gaseous core reactor concept for both low- and burst-power applications  

Microsoft Academic Search

Extensive theoretical and experimental investigations at the Innovative Nuclear Space Power Institute (INSPI) at the University of Florida have shown that gaseous\\/vapor core reactors (GCR) have promising features for space electric power and propulsion applications. The GCR concept presented in this summary consists of two reactor systems: a large central high (burst)-power gaseous core reactor (BPGCR) chamber surrounded by an

M. M. Panicker; E. T. Dugan

1990-01-01

91

Evaluation of molybdenum and its alloys. [Reactor core heat pipes  

SciTech Connect

The choice of pure molybdenum as the prime candidate material for space reactor core heat pipes is critically examined. Pure molybdenum's high ductile-brittle transition temperature appears to be its major disadvantage. The candidate materials examined in detail for this application include low carbon arc-cast molybdenum, TZM-molybdenum alloy, and molybdenum-rhenium alloys. Published engineering properties are collected and compared, and it appears that Mo-Re alloys with 10 to 15% rhenium offer the best combination. Hardware is presently being made from electron beam melted Mo-13Re to test this conclusion.

Lundberg, L.B.

1981-01-01

92

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

Microsoft Academic Search

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

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

2004-01-01

93

Power Distributions in Fresh and Depleted LEU and HEU Cores of the MITR Reactor.  

National Technical Information Service (NTIS)

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 po...

E. H. Wilson J. G. Stevens L. Hu N. E. Horelik P. E. Dunn T. H. Newton

2012-01-01

94

Pressurized water reactor inherent core protection by primary system thermohydraulics  

SciTech Connect

Current light water reactors (LWRs) depend for the protection of core integrity on a multitude of active systems and components, such as instrumentation, cables, electronic logics, relays, actuators, etc., and on human judgment. This approach to safety has led to a complex and expensive plant design in which all parts of the plant where these systems are present must be protecte against damage due to, e.g., earthquake. It has also failed to persuade the public about the safety of the reactors because of the existing (but very small) probability of multiple failures leading to core meltdown. With the process inherent ultimate safety (PIUS) approach, this dependence on active systems is eliminated. The safety is now no longer a result of their intervention but is built into the thermohydraulics of the primary system itself. The PIUS primary system response to a number of severe anticipated transients without scram (ATWS) is described, as studied by means of a specially devel oped computer simulation program. The method is shown by which the thermohydraulic self-protection properties of the primary system terminates these ATWS transients, which could have severe consequences in a conventional LWR, with neither the core nor the rest of the plant suffering any damage (beyond the initial failure assumed). This has important economic consequences. The surveillance and control systems used to run the plant and the buildings in which they are housed can be designed as for a fossil plant, since they no longer have the ultimate responsibility for nuclear safety. The ensuing design simplification pays for the more expensive pressure vessel and primary system. Inherent safety is obtained as a bonus.

Babala, D.; Hannerz, K.

1985-08-01

95

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

NASA Astrophysics Data System (ADS)

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

Glinik, Rafail A.

1993-06-01

96

Seismic response of a block-type nuclear reactor core. [HTGR  

Microsoft Academic Search

An analytical model is developed to predict seismic response of large gas-cooled reactor cores. The model is used to investigate scaling laws involved in the design of physical models of such cores, and to make parameter studies.

R. C. Dove; J. G. Bennett; J. L. Merson

1976-01-01

97

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.

98

In-reactor testing of the closed cycle gas core reactor: The Nuclear Light Bulb concept  

NASA Astrophysics Data System (ADS)

The Nuclear Light Bulb (NLB) concept is an advanced closed cycle space propulsion rocket engine design that offers unprecidented performance characteristics in terms of specific impulse (greater than 1800 s) and thrust (greater than 445 kN). The NLB is a gas-core nuclear reactor making use of thermal radiation from a high temperature U-plasma core to heat the hydrogen propellant to very high temperatures (greater than 4000 K). Analyses performed in support of the design of in-reactor tests that are planned to be performed in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories in order to demonstrate the technical feasibility of this advanced concept are described. The tests will examine the stability of a hydrodynamically confined fissioning U-plasma under steady and transient conditions. Testing will also involve study of propellant heating by thermal radiation from the plasma and materials performance in the nuclear environment of the NLB. The analyses presented include neutronic performance studies and U-plasma radiation heat-transport studies of small vortex-confined fissioning U-plasma experiments that are irradiated in the ACRE. These analyses indicate that high U-plasma temperatures (4000 to 9000 K) can be sustained in the ACRE for periods of time on the order of 5 to 20 s. These testing conditions are well suited to examine the stability and performance requirements necessary to demonstrate the feasibility of this concept.

Gauntt, R. O.; Slutz, S. A.; Harms, G. A.; Latham, T. S.; Roman, W. C.; Rodgers, R. J.

1992-10-01

99

Fast Reactor Core Concepts for Minor Actinide Transmutation Using Hydride Fuel Targets  

Microsoft Academic Search

Fast reactor core concepts are studied which reduce long-term radiotoxicity of nuclear waste by using minor actinides (MAs) in the form of zirconium-hydride fuel targets. A systematic parameter survey is carried out to investigate the fundamental characteristics of MA transmutation and the core safety parameters such as sodium void reactivity in a 1,000 MWe-class fast reactor core. Two core concepts

Toshio SANDA; Koji FUJIMURA; Kaoru KOBAYASHI; Katsuyuki KAWASHIMA; Michio YAMAWAKI; Kenji KONASHI

2000-01-01

100

Internal Control Rod Drive Mechanisms, Design Options for IRIS  

Microsoft Academic Search

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

Lawrence E. Conway; Bojan Petrovic

2004-01-01

101

Nuclear design of a vapor core reactor for space nuclear propulsion  

Microsoft Academic Search

Neutronic analysis methodology and results are presented for the nuclear design of a vapor core reactor for space nuclear propulsion. The Nuclear Vapor Thermal Reactor (NVTR) Rocket Engine uses modified NERVA geometry and systems which the solid fuel replaced by uranium tetrafluoride vapor. The NVTR is an intermediate term gas core thermal rocket engine with specific impulse in the range

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

1993-01-01

102

MEASUREMENT OF AIR-BLAST EFFECTS FROM SIMULATED NUCLEAR-REACTOR-CORE EXCURSIONS  

Microsoft Academic Search

Results are discussed from a method of determining the transient blast ; loading on a containment shell during a simulated reactor core excursion. Three ; methods of energy release were studiedi suddenly releasing nitrogen gas at high ; static pressure; igniting a confined propellant; and detonating an explosive in ; an air-or water-filled simulated reactor core vessel. (W.D.M.);

W. C. Olson; R. J. Larson; H. Goldstein

1959-01-01

103

Diagnostics of vibrations in induction motor-pump system used for reactor core cooling system  

Microsoft Academic Search

Analysis and diagnostics of vibration in industrial systems play a significant rule to prevent severe damages. This paper proposes a non-linear model to simulate the torsional vibration in the reactor core cooling system. Simulation results of an operating reactor core cooling system set with the actual parameters are presented to validate the accuracy and reliability of the proposed analytical method.

S. A. Qutb; A. M. Abdel-Hamid; A. Mansour; S. E. Soliman

2006-01-01

104

Individual pebble temperature peaking factor due to local pebble arrangement in a pebble bed reactor core  

Microsoft Academic Search

Scientists at the German AVR pebble bed nuclear reactor discovered that the surface temperature of some of the pebbles in the AVR core were at least 200K higher than previously predicted by reactor core analysis calculations. The goal of this research paper is to determine whether a similar unexpected fuel temperature increase of 200K can be attributed solely or mostly

Vladimir Sobes; Benoit Forget; Andrew Kadak

2011-01-01

105

Improved core design of the high temperature supercritical-pressure light water reactor  

Microsoft Academic Search

A new coolant flow scheme has been devised to raise the average coolant core outlet temperature of the High Temperature Supercritical-Pressure Light Water Reactor (SCLWR-H). A new equilibrium core is designed with this flow scheme to show the feasibility of an SCLWR-H core with an average coolant core outlet temperature of 530°C.In previous studies, the average coolant core outlet temperature

A. Yamaji; K. Kamei; Y. Oka; S. Koshizuka

2005-01-01

106

A Plant Control System Development Approach for IRIS  

Microsoft Academic Search

The plant control system concept for the International Reactor Innovative and Secure (IRIS) will make use of integrated control, diagnostic, and decision modules to provide a highly automated intelligent control capability. The plant control system development approach established for IRIS involves determination and verification of control strategies based on whole-plant simulation; identification of measurement, control, and diagnostic needs; development of

R. T. Wood; C. R. Brittain; J. A. March-Leuba; L. E. Conway; L. Oriani

107

Dual Purpose Gamma Thermometer for Use as a Reactor Power Level and Core Coolant Level Detector for Pressurized Water Reactors.  

National Technical Information Service (NTIS)

This citation summarizes a one-page announcement of technology available for utilization. A modified gamma thermometer has been studied for use in a pressurized water reactor as a local power level detector and core coolant level detector. Utilization of ...

1982-01-01

108

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

109

CHAP2 heat-transfer analysis of the Fort St. Vrain reactor core  

Microsoft Academic Search

The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh,

J. F. Kotas; K. R. Stroh

1983-01-01

110

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

Microsoft Academic Search

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

E. H. Wilson; N. E. Horelik; F. E. Dunn; Newton T. H. Jr; L. Hu; J. G. Stevens

2012-01-01

111

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

Microsoft Academic Search

Research is being conducted on gas core reactors for space nuclear power to establish the scientific feasibility and engineering validation of a reactor and energy conversion system that can significantly improve specific power, dynamic performance and system efficiency. Rapid achievement of burst mode (GWe) operation at core power densities of 1 kW\\/mL and reactor masses of a kg\\/MWt are research

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

1987-01-01

112

CHAP-2 Heat-Transfer Analysis of the Fort St. Vrain Reactor Core.  

National Technical Information Service (NTIS)

The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model us...

J. F. Kotas K. R. Stroh

1983-01-01

113

MEASUREMENT OF AIR BLAST EFFECTS FROM SIMULATED NUCLEAR REACTOR CORE EXCURSIONS  

Microsoft Academic Search

Tests were conducted to evaluate methods of simulating on a small scale, ; the effect of nuclear reactor runaway'' on a containment shell surrounding the ; reactor. Reactor core vessels, simulated by small pressure tanks, were burst by ; chemical reactions of various rates, and the resulting pressure-time histories ; were recorded by piezoelectric air blast gages placed at various

R. J. Larson; W. C. Olson

1957-01-01

114

Core performance of fast reactors for actinide recycling using metal, nitride, and oxide fuels  

Microsoft Academic Search

Core performance analyses are conducted for fast reactors that accept and recycle the plutonium and minor actinides (MAs) recovered from light water reactor (LWR) spent fuel, together with the plutonium and MAs from the fast reactors` own production. Metal, nitride, and oxide are the fuel materials used to compare the neutronic and safety parameters and to discuss acceptable minor actinide

Takeshi Yokoo; Akihiro Sasahara; Tadashi Inoue; Jungmin Kang; Atsuyuki Suzuki

1996-01-01

115

Commercialized fast reactor cycle systems and reactor core performance of the promising fast reactors  

Microsoft Academic Search

The Feasibility Study on Commercialized Fast Reactor (FR) Cycle Systems is under progress in order to propose prominent FR cycle systems that will respond to the diverse needs of society in the future. The design studies on various FR system concepts have been achieved and then the evaluations of potential to achieve the development targets have been also carried out.

Shoji Kotake; Yoshihiko Sakamoto; Yutaka Sagayama

2005-01-01

116

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

117

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

118

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

Code of Federal Regulations, 2013 CFR

...core cooling systems for light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING...core cooling systems for light-water nuclear power reactors....

2013-01-01

119

Parametric Study of the Colloid Reactor Rocket Engine. Computer Program for Predicting Colloid Core Fuel Vaporization Loss Rates.  

National Technical Information Service (NTIS)

A parametric study was performed of the Colloid Core Reactor Rocket Engine. This report describes the computer program written to calculate fuel vaporization loss rates in the Colloid Core Reactor. The theoretical development of the model and the organiza...

1973-01-01

120

Development of long-lived radionuclide transmutation technology - Development of a code system for core analysis of the transmutation reactors.  

National Technical Information Service (NTIS)

The objective of this study is to develop a code system for core analysis= of the critical transmutation reactors utilizing fast neutrons. Core characteristics of the transmutation reactors were identified and four codes, HANCELL for pincell calculation, ...

N. Z. Cho Y. H. Kim T. H. Kim C. K. Jo C. J. Park

1996-01-01

121

Core Modification to Improve Irradiation Efficiency of the Experimental Fast Reactor Joyo  

Microsoft Academic Search

Core modification has been investigated to further increase the core burnup and to improve the irradiation efficiency of the experimental fast reactor Joyo. This modification enables the core to accommodate more irradiation test subassemblies that have lower fissile material contents compared with the driver fuel. The design calculations showed that the replacement of the radial reflector elements made of stainless

Shigetaka MAEDA; Masaya YAMAMOTO; Tomonori SOGA; Takashi SEKINE; Takafumi AOYAMA

2011-01-01

122

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

Microsoft Academic Search

High Temperature Gas Reactors (HTGRs) cooled by helium have the capability to develop high core outlet temperatures. The upper temperature bound of HTGRs designed and operated to-date is approximately 950 deg. C. But, the goal for the Next Generation Nuclear Plant (NGNP) is a mixed mean core outlet temperature of 1000 deg. C. The most limiting core design criteria governing

M. LaBar; M. Richards; A. Shenoy

2004-01-01

123

Thermal-Hydraulic Characteristics of Double Flat Core HCLWR (High Conversion Light Water Reactor).  

National Technical Information Service (NTIS)

A thermal-hydraulic characteristics of double flat core high conversion light water reactor (HCLWR) is described. The concept of flat core proposed by Ishiguro et al. is to achieve negative void reactivity coefficient in tight lattice core, and at the sam...

J. Sugimoto T. Iwamura T. Okubo Y. Murao

1989-01-01

124

Heat transfer and core neutronics considerations of the heat pipe cooled thermionic reactor  

NASA Astrophysics Data System (ADS)

The authors summarize the results of detailed neutronic and thermal-hydraulic evaluations of the heat pipe cooled thermionic (HPTI) reactor design, identify its key design attributes, and quantify its performance characteristics. The HPTI core uses modular, liquid-metal core heat transfer assemblies to replace the liquid-metal heat transport loop employed by in-core thermionic reactor designs of the past. The nuclear fuel, power conversion, heat transport, and heat rejection functions are all combined into a single modular unit. The reactor/converter assembly uses UN fuel pins to obtain a critical core configuration with in-core safety rods and reflector controls added to complete the subassembly. By thermally bonding the core heat transfer assemblies during the reactor core is coupled neutronically, thermally, and electrically into a modular assembly of individual power sources with cross-tied architecture. A forward-facing heat pipe radiator assembly extends from the reactor head in the shape of a frustum of a cone on the opposite side of the power system from the payload. Important virtues of the concept are the absence of any single-point failures and the ability of the core to effectively transfer the TFE waste heat load laterally to other in-core heat transfer assemblies in the event of multiple failures in either in-core and radiator heat pipes.

Determan, W. R.; Lewis, Brian

125

Core design studies for a 1000 MW{sub th} advanced burner reactor.  

SciTech Connect

This paper describes the core design and performance characteristics of 1000 MW{sub th} Advanced Burner Reactor (ABR) core concepts with a wide range of TRU conversion ratio. Using ternary metal alloy and mixed oxide fuels, reference core designs of a medium TRU conversion ratio of {approx}0.7 were developed by trade-off between burnup reactivity loss and TRU conversion ratio. Based on these reference core concepts, TRU burner cores with low and high TRU conversion ratios were developed by changing the intra-assembly design parameters and core configurations. Reactor performance characteristics were evaluated in detail, including equilibrium cycle core performances, reactivity feedback coefficients, and shutdown margins. The results showed that by employing different assembly designs, a wide range of TRU conversion ratios from {approx}0.2 to break-even can be achieved within the same core without introducing significant performance and safety penalties.

Kim, T. K.; Yang, W. S.; Grandy, C.; Hill, R.; Nuclear Engineering Division

2009-04-01

126

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

127

Core performance of new concept passive-safety reactor “kamado” - safety, burn-up and uranium resource problem -  

Microsoft Academic Search

New concept of a passive-safety reactor “KAMADO” has a negligible possibility of core melting and flexibility of total reactor power. The reactor core of KAMADO consists of fuel elements of graphite blocks, which have UO2 fuel rods and cooling water holes. These fuel elements are located in a reactor water pool of atmospheric pressure (1 atm) and low temperature (<

Tetsuo Matsumura; Takanori Kameyama; Yasushi Nauchi; Izumi Kinoshita

2005-01-01

128

Status of axial heterogeneous liquid-metal fast breeder reactor core design studies and research and development  

Microsoft Academic Search

The current status of axial heterogeneous core (AHC) design development in Japan, which consists of an AHC core design in a pool-type demonstration fast breeder reactor (DFBR) and research and development activities supporting AHC core design, is presented. The DFBR core design objectives developed by The Japan Atomic Power Company include (a) favorable core seismic response, (b) core compactness, (c)

H. Nakagawa; T. Inagaki; H. Yoshimi; K. Shirakata; Y. Watari; M. Suzuki; K. Inoue

1988-01-01

129

Calculated Neutron and Gamma-Ray Spectra across the Prismatic Very High Temperature Reactor Core  

NASA Astrophysics Data System (ADS)

Neutron and gamma-ray flux spectra are calculated using the MCNP5 computer code and a one-sixth core model of a prismatic Very High Temperature Reactor based on the General Atomics Gas Turbine-Modular Helium Reactor. Spectra are calculated in the five inner reflector graphite block rings, three annular active core fuel rings, three outer graphite reflector block rings, and the core barrel. The neutron spectra are block and fuel pin averages and are calculated as a function of temperature and burnup. Also provided are the total, fast, and thermal radial profile fluxes and core barrel dpa rates.

Sterbentz, James W.

2009-08-01

130

IRIS Process (2008)  

EPA Science Inventory

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

131

IRIS Process (2009 Update)  

EPA Science Inventory

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

132

IRIS Process (Pre-2004)  

EPA Science Inventory

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

133

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

134

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

135

Analysis of in-Core Dynamics in Pressurized Water Reactors with Application to Parameter Monitoring.  

National Technical Information Service (NTIS)

The behavior of the phase relationship between neutron flux and core-exit temperature fluctuations in a pressurized water reactor (PWR) is studied as a function of the moderator temperature coefficient of reactivity ( alpha /sub c/). PWR operational data ...

B. R. Upadhyaya F. J. Sweeney D. J. Shieh O. Glockler

1987-01-01

136

Critical Heat Flux (CHF) Characteristics of High Conversion Pressurized Water Reactor with Double Flat Core.  

National Technical Information Service (NTIS)

Thermal-hydraulic feasibility of a high conversion pressurized water reactor (HCPWR) with a double flat core was studied from a view point of minimum departure from nucleate boiling ratio (MDNBR). The proposed HCPWR improves uranium utilization under the ...

T. Iwamura J. Sugimoto T. Okubo Y. Murao

1989-01-01

137

Preparations to ship the TMI2 damaged reactor core  

Microsoft Academic Search

The March 1979 accident at Three Mile Island Unit 2 (TMI-2) resulted in a severely damaged core. Entries into that core using various tools and inspection devices have shown a significant void, large amounts of rubble, partially intact fuel assemblies, and some resolidified molten materials. The removal and disposition of that core has been of considerable public, regulatory, and governmental

R. C. Schmitt; G. J. Quinn

1985-01-01

138

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

Microsoft Academic Search

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

Mehrdad Boroushaki; Mohammad B. Ghofrani; Caro Lucas

2002-01-01

139

Analysis of Mixed Oxide Fuel Loaded Cores in the Heavy Water Reactor FUGEN  

Microsoft Academic Search

Uranium-plutonium mixed oxide (MOX) fuel cores in the heavy water reactor, FUGEN, were analyzed using the Advanced Thermal Reactor (ATR) type core design code system WIMS-ATR\\/POLESTAR and the accuracy of this code system also has been evaluated by means of operational data through the 34 burnup cycles and on-site ?-scanning data. The root mean square errors of calculated thermal neutron

Tsukasa OHTANI; Takashi IIJIMA; Yoshitake SHIRATORI

2003-01-01

140

Analysis of Mixed Oxide Fuel Loaded Cores in the Heavy Water Reactor FUGEN  

Microsoft Academic Search

Uranium-plutonium mixed oxide (MOX) fuel cores in the heavy water reactor, FUGEN, were analyzed using the Advanced Thermal Reactor (ATR) type core design code system WIMS-ATR\\/POLESTAR and the accuracy of this code system also has been evaluated by means of operational data through the 34 burnup cycles and on-site ? -scanning data. The root mean square errors of calculated thermal

Tsukasa OHTANI; Takashi IIJIMA; Yoshitake SHIRATORI

2003-01-01

141

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

142

Method of and apparatus for measuring the power distribution in nuclear reactor cores  

SciTech Connect

The invention disclosed is the method of exact calibration of gamma ray detectors called gamma thermometers prior to acceptance for installation into a nuclear reactor core. This exact calibration increases the accuracy of determining the power distribution in the nuclear reactor core. The calibration by electric resistance heating of the gamma thermometer consists of applying an electric current along the controlled heat path of the gamma thermometer and then measuring the temperature difference along this controlled heat path as a function of the amount of power generated by the electric resistance heating. Then, after the gamma thermometer is installed into the nuclear reactor core and the reactor core is operating at power producing conditions, the gamma ray heating of the detector produces a temperature difference along the controlled heat path. With the knowledge of this temperature difference, the calibration characteristic determined by the prior electric resistance heating is employed to accurately determine the local rate of gamma ray heating. The accurate measurement of the gamma heating rate at each location of a set of locations throughout the nuclear reactor core is the basis for accurately determining the power distribution within the nuclear reactor core.

Leyse, R.H.

1983-07-12

143

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

Microsoft Academic Search

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

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

2010-01-01

144

Criticality aspects of nuclear power reactor cores in the case of emerging nuclear fuels  

Microsoft Academic Search

Reactor cores of PWR and LMFBR, loaded with different commercial and emerging nuclear fuels, have been simulated and compared at BOI with respect to criticality with and without chemical shim, control rods and sodium. The different cases considered, within each of the reactor types, are grouped together according to their fissile content, when compared on the basis of the neutron

G. Nicolaou

2010-01-01

145

Evaluating the core damage frequency of a TRIGA research reactor using risk assessment tool software  

Microsoft Academic Search

After all preventive and mitigative measures considered in the design of a nuclear reactor, the installation still represents a residual risk to the outside world. Probabilistic safety assessment (PSA) is a powerful method to survey the safety of nuclear reactors. In this study the occurrence frequency of different types of core damage states (CDS) which may potentially arise in Tehran

Shahabeddin Kamyab; Mohammadreza Nematollahi

146

Three pass core design proposal for a high performance light water reactor  

Microsoft Academic Search

The paper describes a novel core concept for a nuclear reactor cooled with supercritical water, in which the coolant is heated up from 280°C at the reactor inlet to 500°C at the outlet in four steps: a first heat-up step is provided by heat transfer from fuel assemblies to the moderator water in gaps and moderator boxes, a second step

T. Schulenberg; J. Starflinger; J. Heinecke

2008-01-01

147

Comparison of two models for a pebble bed modular reactor core coupled to a Brayton cycle  

Microsoft Academic Search

The pebble bed modular reactor (PBMR) plant is a promising concept for inherently safe nuclear power generation. This paper presents two dynamic models for the core of a high temperature reactor (HTR) power plant with a helium gas turbine. Both the PBMR and its power conversion unit (PCU) based on a three-shaft, closed cycle, recuperative, inter-cooled Brayton cycle have been

Ayelet Walter; Alexander Schulz; Günter Lohnert

2006-01-01

148

Nuclear mechanism for TRU burning oxide fueled core in Advanced Recycling Reactor  

Microsoft Academic Search

This paper presents an approximation approach to predict the core characteristics based on parametric survey and an analysis of nuclear mechanism in a conceptual nuclear design for enhanced transuranics (TRU) burning mixed oxide fueled and sodium cooled fast reactor which can be realized in the near future. The design study of Advanced Recycling Reactor was conducted in the context of

Kazumi Ikeda; Hiroshi Sekimoto

2010-01-01

149

Optimized core design and fuel management of a pebble-bed type nuclear reactor  

Microsoft Academic Search

The core design of a pebble-bed type Very High Temperature Reactor (VHTR) is optimized, aiming for an increase of the coolant outlet temperature to 1000 C, while retaining its inherent safety features. The VHTR has been selected by the international Generation IV research initiative as one of the six most promising nuclear reactor concepts that are expected to enter service

B. Boer

2009-01-01

150

Operating experience of natural circulation core cooling in boiling water reactors  

Microsoft Academic Search

General Electric (GE) has proposed an advanced boiling water reactor, the Simplified Boiling Water Reactor (SBWR), which will utilize passive, gravity-driven safety systems for emergency core coolant injection. The SBWR design includes no recirculation loops or recirculation pumps. Therefore the SBWR will operate in a natural circulation (NC) mode at full power conditions. This design poses some concerns relative to

C. Kullberg; K. Jones; C. Heath

1993-01-01

151

Investigation of Core Thermohydraulics in Fast Reactors - Interwrapper Flow During Natural Circulation  

Microsoft Academic Search

A proper assessment of core thermohydraulics under natural circulation conditions is important so that the full potential of the inherent, passive feature of a fast reactor can be used. When the heat exchangers of the decay heat removal system are operated in the upper plenum of a reactor vessel, cold sodium exiting the heat exchangers may penetrate into the gap

H. Kamide; K. Hayashi; T. Isozaki; M. Nishimura

2001-01-01

152

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

Microsoft Academic Search

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

David W. Nigg; Devin A. Steuhm

2011-01-01

153

Nodal Diffusion Burnable Poison Treatment for Prismatic Reactor Cores  

Microsoft Academic Search

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)

A. M. Ougouag; R. M. Ferrer

2010-01-01

154

Optimum utilization of nuclear fuel with gas and vapor core reactors  

Microsoft Academic Search

Gas and Vapor Core Reactors (G\\/VCR) are externally reflected and moderated nuclear energy systems fueled by stable uranium compound in gaseous or vapor phase. In G\\/VCR systems the functions of fuel and coolant are combined and the reactor outlet temperature is not constrained by solid fuel-cladding temperature limitations. G\\/VCRs can potentially provide the highest reactor and cycle temperature among all

Samim Anghaie; Travis W. Knight; Rob Norring; Blair M. Smith

2005-01-01

155

Conceptual Design Study of 180 MWt Small-Sized Reduced-Moderation Water Reactor Core  

Microsoft Academic Search

Conceptual design of a Small-sized Reduced-Moderation Water Reactor (S-RMWR) core, which has the thermal output of 180 MW, the conversion ratio of 1.0 and the void reactivity coefficient of negative value, has been constructed. S-RMWR is a technology demonstration reactor which also conducts material and fuel testing for commercial use of Reduced-Moderation Water Reactor (RMWR) in large-scale power plants. It

Yoshihiro NAKANO; Tsutomu OKUBO; Sadao UCHIKAWA

2006-01-01

156

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

157

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

Microsoft Academic Search

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

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

1992-01-01

158

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

Microsoft Academic Search

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

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

1986-01-01

159

Vaporization of core materials in postulated severe light water reactor accidents  

Microsoft Academic Search

The vaporization of core materials other than fission products during a postulated severe light water reactor accident is treated by chemical thermodynamics. The core materials considered were (a) the control rod materials, silver, cadmium, and indium; (b) the structural materials, iron, chromium, nickel, and manganese; (c) cladding material, zirconium and tin; and (d) the fuel, uranium oxide. Thermodynamic data employed

D. Cubicciotti; B. R. Sehgal

1984-01-01

160

Thermal-hydraulic mixing in the split-core ANS reactor design  

Microsoft Academic Search

A design has been proposed for the advanced neutron source (ANS) reactor that incorporates a split core, one purpose of which is to create a mixing plenum between the upper and lower cores. It was hoped that in addition to introducing various desirable neutronics features, such as decreasing the fast neutron flux contamination of thermal and cold neutron beams located

Dorning

1988-01-01

161

Review of experimental results of light water reactor core melt progression  

Microsoft Academic Search

This paper reports on results from integral-effects core melt progression experiments and from the examination of the damaged core of the Three Mile Island Unit 2 (TMI-2) reactor which are reviewed to gain insight on key severe accident phenomena. The experiments and the TMI-2 accident represent a wide variety of conditions and physical scales, yet several important phenomena appear to

R. R. Hobbins; D. A. Petti; O. J. Osetek; D. L. Hagrman

1991-01-01

162

2ND Reactor Core of the NS Otto Hahn. Design, Operation Experience, Developments.  

National Technical Information Service (NTIS)

Details of the design of the 2nd reactor core are given, followed by a brief report summarising the operating experience gained with this 2nd core, as well as by an evaluation of measured data and statements concerning the usefulness of the knowledge gain...

H. J. Manthey H. Kracht

1979-01-01

163

Structural failure analysis of reactor vessels due to molten core debris  

Microsoft Academic Search

Maintaining structural integrity of the reactor vessel during a postulated core melt accident is an important safety consideration in the design of the vessel. This paper addresses the failure predictions of the vessel due to thermal and pressure loadings from the molten core debris depositing on the lower head of the vessel. Different loading combinations were considered based on a

1993-01-01

164

Multiplexed fibre Bragg grating sensors for in-core thermometry in nuclear reactors  

Microsoft Academic Search

In-core temperature measurement is a critical issue for the safe operation of nuclear reactors. Classical thermocouples require shielded connections and are known to drift under high neutron fluence. As an alternative, we propose to take advantage of the multiplexing capabilities of Fibre Bragg Grating (FBG) temperature sensors to perform the in-core temperature measurements. We first report on our irradiation experiments

A. I. Gusarov; O. Deparis; P. Megret; M. Blondel; A. Delchambre

165

[Vascular anomaly of the iris].  

PubMed

A 48-year-old man presented with a vascular anomaly of the iris in the left eye. Slit-lamp microscopy revealed dilated and tortuous vessels of the iris between 12 and 4 o'clock. Fluorescein angiography confirmed a diagnosis of arteriovenous (AV) malformation of the iris. The vessel originated at the iris base, passed to the pupillary margin and returned to the base. Such AV-malformations of the iris are very rare, benign vascular anomalies that have to be distinguished from other, potentially malignant pathologies of the iris (e. g. tortuous vessels in iris melanoma). PMID:22526009

Ponto, K A; Mirshahi, A

2012-07-01

166

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

167

In-reactor testing of the closed cycle gas core reactor-the nuclear light bulb concept  

Microsoft Academic Search

The Nuclear Light Bulb (NLB) concept is an advanced closed cycle space propulsion rocket engine design that offers unprecidented performance characteristics in terms of specific impulse (>~1800 s) and thrust (>~445 kN). The NLB is a gas-core nuclear reactor making use of thermal radiation from a high temperature U-plasma core to heat the hydrogen propellant to very high temperatures (~4000

Randall O. Gauntt; Stephen A. Slutz; Gary A. Harms; Thomas S. Latham; Ward C. Roman; Richard J. Rodgers

1993-01-01

168

In-reactor testing of the closed cycle gas core reactor—the nuclear light bulb concept  

Microsoft Academic Search

The Nuclear Light Bulb (NLB) concept is an advanced closed cycle space propulsion rocket engine design that offers unprecidented performance characteristics in terms of specific impulse (?1800 s) and thrust (?445 kN). The NLB is a gas-core nuclear reactor making use of thermal radiation from a high temperature U-plasma core to heat the hydrogen propellant to very high temperatures (?4000

Randall O. Gauntt; Stephen A. Slutz; Gary A. Harms; Thomas S. Latham; Ward C. Roman; Richard J. Rodgers

1993-01-01

169

Survey of dust production in pebble bed reactor cores  

Microsoft Academic Search

Graphite dust produced via mechanical wear from the pebbles in a pebble bed reactor is an area of concern for licensing. Both the German pebble bed reactors produced graphite dust that contained activated elements. These activation products constitute an additional source term of radiation and must be taken under consideration during the conduct of accident analysis of the design. This

Joshua J. Cogliati; Abderrafi M. Ougouag; Javier Ortensi

2011-01-01

170

Survey of Dust Production in Pebble Bed Reactors Cores  

SciTech Connect

Graphite dust produced via mechanical wear from the pebbles in a pebble bed reactor is an area of concern for licensing. Both the German pebble bed reactors produced graphite dust that contained activated elements. These activation products constitute an additional source term of radiation and must be taken under consideration during the conduct of accident analysis of the design. This paper discusses the available literature on graphite dust production and measurements in pebble bed reactors. Limited data is available on the graphite dust produced from the AVR and THTR-300 pebble bed reactors. Experiments that have been performed on wear of graphite in pebble-bed-like conditions are reviewed. The calculation of contact forces, which are a key driving mechanism for dust in the reactor, are also included. In addition, prior graphite dust predictions are examined, and future areas of research are identified.

Joshua J. Cogliati; Abderafi M. Ougouag; Javier Ortensi

2011-06-01

171

The Encapsulated Nuclear Heat Source (ENHS) Reactor Core Design  

SciTech Connect

A once-for-life, uniform composition, blanket-free and fuel-shuffling-free reference core has been designed for the Encapsulated Nuclear Heat Source (ENHS) to provide the design goals of a nearly zero burnup reactivity swing throughout {approx}20 yr of full-power operation up to the peak discharge burnup of more than 100 GWd/t HM. What limits the core life is radiation damage to the HT-9 structural material. The temperature coefficients of reactivity are all negative, except for that of the coolant expansion. However, the negative reactivity coefficient associated with the radial expansion of the core structure can compensate for the coolant thermal expansion. The void coefficient is positive but of no safety concern because the boiling temperature of lead or lead-bismuth is so high that there is no conceivable mechanism for the introduction of significant void fraction into the core. The core reactivity coefficients, reactivity worth, and power distributions are almost constant throughout the core life.It was found possible to design such once-for-life cores using different qualities of Pu and transuranics as long as U is used as the primary fertile material. It is also feasible to design ENHS cores using nitride rather than metallic fuel. Relative to the reference metallic fuel core, nitride fuel cores offer up to {approx}25% higher discharge burnup and longer life, up to {approx}38% more energy per core, a significantly more negative Doppler reactivity coefficient, and less positive coolant expansion and coolant void reactivity coefficient but a somewhat smaller negative fuel expansion reactivity coefficient. The pitch-to-diameter ratio (1.45 of the nitride fuel cores using enriched N) is larger than that (1.36) for the reference metallic fuel core, implying a reduction of the coolant friction loss, thus enabling an increase in the power level that can be removed from the core by natural circulation cooling.It is also possible to design Pu-U(10Zr) fueled ENHS-type cores using Na as the primary coolant with either Na or Pb-Bi secondary coolants. The Na-cooled cores feature a tighter lattice and are therefore more compact but have spikier power distribution, more positive coolant temperature reactivity coefficients, and smaller reactivity worth of the control elements.

Hong, Ser Gi [Korea Atomic Energy Research Institute (Korea, Republic of); Greenspan, Ehud [University of California, Berkeley (United States); Kim, Yeong Il [Korea Atomic Energy Research Institute (Korea, Republic of)

2005-01-15

172

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

SciTech Connect

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

Takahisa Yamamoto; Koshi Mitachi; Masatoshi Nishio [Toyohashi University of Technology, 1-1, Hibarigaoka, Tempaku-cho, Toyohashi-shi Aichi, 4418580 (Japan)

2006-07-01

173

Core melt spreading on a reactor containment floor  

Microsoft Academic Search

The ex-vessel core melt spreading, cooling and stabilization is proposed for a nuclear power plant containment design. Clearly, the retention and coolability of the decay-heated core debris is very much the focal point in the proposed new and advanced designs so that, in the postulated event of a severe accident, the containment integrity is maintained and the risk of radioactivity

T. N Dinh; M. J Konovalikhin; B. R Sehgal

2000-01-01

174

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

Microsoft Academic Search

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

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

2010-01-01

175

The IRIS Data Management  

NASA Astrophysics Data System (ADS)

The IRIS network (International Research on the Interior of the Sun) has been running since August 1987 with one to three active sites, and data analysis is now the major task to undertake. After describing how the IRIS data are generated, we present a four step analysis flowchart, starting with raw uncalibrated data, and moving toward ‘satellite-like’ continuous calibrated, timed, and merged time series which the helioseismological analysis will use. A task-by-task description is presented.

Gelly, Bernard

1991-05-01

176

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

177

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

178

Exploiting iris dynamics  

NASA Astrophysics Data System (ADS)

The human iris is a circular curtain over the light entrance pupil which is controlled directly by the intensity of blue light from photosensitive ganglions in the retina within the eye. The human iris dynamic is remarkable in that it is capable of shrinking concentrically along the radial direction by a factor 4 from 8mm to 2mm, and constantly oscillates in 1/2 second periodicity. Pupil dilation and contraction causes the iris texture to undergo nonlinear deformation with discrete components and minutia features. Thus, iris recognition must be scale invariant due to the pupil dynamics. We propose the Mandelbrot fractal dimension count of minutia iris details, at different intensity thresholds, in dilation-invariant wedge-boxes, formed at specific angular sizes, but spatially varying over 4 90° quadrants due to the cellular growth under the gravity. Despite the concentric dynamic, we have sought an invariant fractal dimensionality in the circular direction and discovered the non-isotropic effect, departed from the simple Richardson fractal law. Furthermore, we choose an optimum Rayleigh criterion ?/D matching the robust fine resolution scale for the given lens aperture D and the illumination wavelength ? for a potential application from a distant, with the help of comprehensive biometric including iris.

Hsu, Charles; Szu, Harold

2010-04-01

179

The Encapsulated Nuclear Heat Source (ENHS) Reactor Core Design  

Microsoft Academic Search

A once-for-life, uniform composition, blanket-free and fuel-shuffling-free reference core has been designed for the Encapsulated Nuclear Heat Source (ENHS) to provide the design goals of a nearly zero burnup reactivity swing throughout â20 yr of full-power operation up to the peak discharge burnup of more than 100 GWd\\/t HM. What limits the core life is radiation damage to the HT-9

Ser Gi Hong; Ehud Greenspan; Yeong Il Kim

2005-01-01

180

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

181

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

2011-09-01

182

Optimizing a three-element core design for the Advanced Neutron Source Reactor  

SciTech Connect

Source of neutrons in the proposed Advanced Neutron Source facility is a multipurpose research reactor providing 5-10 times the flux, for neutron beams, of the best existing facilities. Baseline design for the reactor core, based on the ``no new inventions`` rule, was an assembly of two annular fuel elements similar to those used in the Oak Ridge and Grenoble high flux reactors, containing highly enriched U silicide particles. DOE commissioned a study of the use of medium- or low-enriched U; a three-element core design was studied as a means to provide extra volume to accommodate the additional U compound required when the fissionable {sup 235}U has to be diluted with {sup 238}U to reduce the enrichment. This paper describes the design and optimization of that three-element core.

West, C.D.

1995-12-31

183

Analysis of the Seismic Response of a Fast Reactor Core. Effects of the Vessel Core Seismic Interaction and Applications of the Results.  

National Technical Information Service (NTIS)

This report deals with the methods to apply for a correct evaluation of the reactor core seismic response. Reference is made to up-to-date design data concerning the PEC core, taking into account the presence of the core-restraint plate located close to t...

A. Martelli S. Chiocchio R. Melloni P. G. Muratori S. Rizzi

1984-01-01

184

Core conversion anaylses for the Portuguese Research Reactor  

Microsoft Academic Search

Design and safety analyses are presented for conversion of the Portuguese Research Reactor (RPI) from the use of HEU fuel to the use of LEU fuel. The analyses were performed jointly by the RERTR Program at the Argonne National Laboratory (ANL) and the Instituto Tecnologico e Nuclear (ITN). The LEU fuel assembly design uses USi-Al dispersion fuel with 4.8 g

J. E. Matos; J. G. Stevens; E. E. Feldman; J. A. Stillman; F. E. Dunn; K. Kalimullah; J. G. Marques; N. P. Barradas; A. R. Ramos; A. Kling

2006-01-01

185

Physics of rewetting in water reactor emergency core cooling  

Microsoft Academic Search

Surface rewetting is essential for the re-establishment of normal and ; safe temperature levels following dryout in rod clusters or boiler tubes, or ; following postulated loss-of-coolant accidents in water reactors. Rewetting ; experiments have been performed with tubes and rods with a wide range of ; materials and experimental conditions (surface temperatures 300 to 8O0 deg C, ; constant

R. B. Duffey; D. T. C. Porthouse

1973-01-01

186

Modification of the Penn State reactor to allow transverse and rotational core motion  

SciTech Connect

At Pennsylvania State University (Penn State), nuclear engineering students have the opportunity to perform experiments in reactor physics, work with reactor and radiation instrumentation, and operate a nuclear reactor. These activities are carried out at the Penn State Breazeale Reactor (PSBR), a General Atomics Mark III TRIGA research reactor. Unfortunately, these activities alone cannot fully support the facility. The PSBR is mandated by Penn State to provide a portion of its operating budget by selling services to users outside as well as within Penn State. To increase the marketability of PSBR, an upgrade program was started to increase the quality and versatility of operation. The PSBR is the longest operating university reactor in the United States. The first phase of the upgrade program began in 1992. The quality of operation was increased by replacing a 1965-vintage console with a more reliable digital control and monitoring system. The current phase of the upgrade program is expected to increase the versatility of operation by modifying the reactor to allow transverse and rotational core motion. The addition of two more degrees of motion to the reactor core will increase the capability of the facility to meet the needs of current and future users. This upgrade is being financed by a grant from the US Department of Energy and matching funds from Penn State.

Hughes, D.E.

1994-12-31

187

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

Microsoft Academic Search

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

Mathias Enohnyaket; Jonas Ekman

2009-01-01

188

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

189

Demonstration of the reactivity constraint approach on SNL's annual core research reactor  

Microsoft Academic Search

This paper reports on the initial demonstration of the reactivity constraint approach and its implementing algorithm, the MIT-CSDL Non-Linear Digital Controller, on the annual core research reactor (ACCR) that is operated by the Sandia National Laboratories. This demonstration constituted the first use of reactivity constraints for the closed-loop, digital control of reactor power on a facility other than the Massachusetts

J. A. Bernard; K. S. Kwok; F. J. Wyant; F. V. Thome

1989-01-01

190

Solid-Core, Gas-Cooled Reactor for Space and Surface Power  

SciTech Connect

The solid-core, gas-cooled, Submersion-Subcritical Safe Space (S and 4) reactor is developed for future space power applications and avoidance of single point failures. The Mo-14%Re reactor core is loaded with uranium nitride fuel in enclosed cavities, cooled by He-30%Xe, and sized to provide 550 kWth for seven years of equivalent full power operation. The beryllium oxide reflector disassembles upon impact on water or soil. In addition to decreasing the reactor and shadow shield mass, Spectral Shift Absorber (SSA) materials added to the reactor core ensure that it remains subcritical in the worst-case submersion accident. With a 0.1 mm thick boron carbide coating on the outside surface of the core block and 0.25 mm thick iridium sleeves around the fuel stacks, the reflector outer diameter is 43.5 cm and the combined reactor and shadow shield mass is 935.1 kg. With 12.5 atom% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide intersititial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating, the S and 4 reactor has a slightly smaller reflector outer diameter of 43.0 cm, and a total reactor and shield mass of 901.7 kg. With 8.0 atom% europium-151 added to the fuel, 2.0 mm diameter europium-151 sesquioxide interstitial pins, and a 0.1 mm thick europium-151 sesquioxide coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respect0011ive.

King, Jeffrey C.; El-Genk, Mohamed S. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM 87131 (United States); Chemical and Nuclear Engineering Dept., University of New Mexico, Albuquerque, NM 87131 (United States)

2006-01-20

191

Deposit suppression in the core of water-cooled nuclear reactors  

Microsoft Academic Search

In pressurized water-cooled nuclear reactors, the formation of deposits on surfaces in the core of the reactor, e.g. on fuel sheaths, is suppressed by maintaining in the circulating pressurized water, a high concentration of ammonia ranging from about 120 to about 200 mg Nhâ\\/kg water. Crevice corrosion of the fuel sheaths is avoided, even under localized boiling conditions, since the

Burrill

1982-01-01

192

Research of the power frequency magnetic fields distribution around dry-type air-core reactor  

Microsoft Academic Search

The strong magnetic fields generated by reactors, which not only interfere other electric equipmentspsila regularly running, but also are potentially harmful to substation employeespsila health, are the main electromagnetic contamination in substations. In this paper, a mixed finite-element (FE) model for dry-type air-core reactor, which considers the effect of eddy current, is proposed for calculation of magnetic fields. According to

Zhang Yan; Wang Quandi; Li Bin; Zhao Yi; Liu Qingsheng; Hazem H. Refai

2008-01-01

193

Benchmark Tests of Radiation Transport Computer Codes for Reactor Core and Shield Calculations  

Microsoft Academic Search

Aiming at providing test problems that may be used to verify an adequate performance of the current version of a neutron and ?-ray transport computer code used for reactor core or shield calculations, we summarize the input data and the calculated results for three benchmark problems.The 1st problem deals with a 1-dimensional small spherical reactor for use to test 1-dimensional

Takumi ASAOKA; Norio ASANO; Hisashi NAKAMURA; Hiroshi MIZUTA; Hiroshi CHICHIWA; Tadahiro OHNISHI; Shun-ichi MIYASAKA; Atsushi ZUKERAN; Tsuneo TSUTSUI; Toichiro FUJIMURA; Satoru KATSURAGI

1978-01-01

194

Verification of the ORIGEN2 code analysis for the TMI-2 reactor core  

SciTech Connect

Accurate definition of the fission product inventories produced in the TMI-2 reactor prior to the accident on March 29, 1979 are of considerable interest to many organizations including the Department of Energy which is shipping the damage reactor core to the Idaho National Engineering Laboratory and conducting the TMI-2 reactor examination program, and General Public Utilities which is defueling the reactor. Numerous fission product inventory calculations have been performed for the TMI-2 core, including an ORIGEN2 analysis by EG G which uses 1239 nodes to define burnup in the reactor core. To provide a verification of the predicted fission product inventories, a measurement study was performed using pellets from various core regions. Measurements were performed for transuranics, burnup monitors, noble gases, principal gamma ray emitters, {sup 129}I, and {sup 90}Sr. Comparisons between the experimental results and the code analyses are presented with an evaluation of the associated uncertainties. Also, a discussion is presented of the probable causes of the observed differences between the code and measured values. 10 refs., 1 fig., 2 tabs.

Akers, D.W.; Schnitzler, B.G.

1988-01-01

195

Impact of the Core Minor Actinide Content on Fast Reactors Reactivity Coefficients  

SciTech Connect

A major challenge for future Fast Reactors could be the recycling of minor actinides (MA) in the core fuel, in order to minimize wastes and contribute to meet both the sustainability objective and the reduction of the burden on a geological disposal. Although the most outstanding issues will be found in the development and validation of the appropriate fuels, the presence of MA in the core can potentially deteriorate the core reactivity coefficients. In the present paper we will show however that there is no well defined physical limit to the amount of MA in the core fuel, but that a careful physics analysis can indicate the most appropriate measures to reduce the MA impact on the reactivity coefficients, and in particular, for Na cooled reactors, on the Na void reactivity coefficient.

G. Palmiotti; M. Salvatores

2011-03-01

196

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

197

Loss-of-fluid test findings in pressurized water reactor core's thermal-hydraulic behavior  

Microsoft Academic Search

This paper summarizes the pressurized water reactor (PWR) core's thermal-hydraulic behavior findings from experiments performed at the Loss-of-Fluid Test (LOFT) Facility at the Idaho National Engineering Laboratory. The potential impact of these findings on the safety and economics of PWR's generation of electricity is also discussed. Reviews of eight important findings in the core's physical behavior and in experimental methods

1983-01-01

198

Core conversion anaylses for the Portuguese Research Reactor.  

SciTech Connect

Design and safety analyses are presented for conversion of the Portuguese Research Reactor (RPI) from the use of HEU fuel to the use of LEU fuel. The analyses were performed jointly by the RERTR Program at the Argonne National Laboratory (ANL) and the Instituto Tecnologico e Nuclear (ITN). The LEU fuel assembly design uses U{sub 3}Si{sub 2}-Al dispersion fuel with 4.8 g U/cm{sup 3} and is very similar to the HEU fuel design. The results of neutronic studies, steady-state thermal-hydraulic analyses, accident analyses, and revisions to the Operating Limits and Conditions demonstrate that the RPI reactor can be operated safely with the new LEU fuel assemblies. Delivery of the LEU fuel is expected around the end of 2006, with conversion in early 2007. The HEU fuel is planned to be returned to the US in 2008.

Matos, J. E.; Stevens, J. G.; Feldman, E. E.; Stillman, J. A.; Dunn, F. E.; Kalimullah, K.; Marques, J. G.; Barradas, N. P.; Ramos, A .R.; Kling, A.; Inst. Tecnologico e Nuclear

2006-01-01

199

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

200

Demonstration of the reactivity constraint approach on SNL's annual core research reactor  

SciTech Connect

This paper reports on the initial demonstration of the reactivity constraint approach and its implementing algorithm, the MIT-CSDL Non-Linear Digital Controller, on the annual core research reactor (ACCR) that is operated by the Sandia National Laboratories. This demonstration constituted the first use of reactivity constraints for the closed-loop, digital control of reactor power on a facility other than the Massachusetts Institute of Technology's (MIT's) research reactor (MITR-II). Also, because the ACRR and the MITR-II are of very different design, these trials established the generic nature of the reactivity constraint approach.

Bernard, J.A.; Kwok, K.S.; Wyant, F.J.; Thome, F.V.

1989-01-01

201

Iris und Kammerwasser-Zirkulation  

Microsoft Academic Search

The anterior surface of the human iris was studied by transmission and scanning electron microscopy. Melanocytes, spongy grouped cells looking like fibrocytes, and bundles of collagen fibers between the cells were found. Holes and pores of different size reach down to the iris stroma. No closed-banded endothelium was found on the anterior surface of the iris. It is assumed that

Claus E. Dieterich; Rudolf Wither; Helmut E. Franz

1971-01-01

202

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

203

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

204

In-reactor testing of the closed cycle gas core reactor-the nuclear light bulb concept  

NASA Astrophysics Data System (ADS)

The Nuclear Light Bulb (NLB) concept is an advanced closed cycle space propulsion rocket engine design that offers unprecidented performance characteristics in terms of specific impulse (>~1800 s) and thrust (>~445 kN). The NLB is a gas-core nuclear reactor making use of thermal radiation from a high temperature U-plasma core to heat the hydrogen propellant to very high temperatures (~4000 K). The following paper describes analyses performed in support of the design of in-reactor tests that are planned to be performed in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories in order to demonstrate the technical feasibility of this advanced concept. The tests will examine the stability of a hydrodynamically confined fissioning U-plasma under steady and transient conditions. Testing will also involve study of propellant heating by thermal radiation from the plasma and materials performance in the nuclear environment of the NLB. The analyses presented here include neutronic performance studies and U-plasma radiation heat-transport studies of small vortex-confined fissioning U-plasma experiments that are irradiated in the ACRR. These analyses indicate that high U-plasma temperatures (4000 to 9000 K) can be sustained in the ACRR for periods of time on the order of 5 to 20 s. These testing conditions are well suited to examine the stability and performance requirements necessary to demonstrate the feasibility of this concept.

Gauntt, Randall O.; Slutz, Stephen A.; Harms, Gary A.; Latham, Thomas S.; Roman, Ward C.; Rodgers, Richard J.

1993-01-01

205

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

206

Multipurpose Advanced 'inherently' Safe Reactor (MARS): Core design studies  

SciTech Connect

In the year 2005, in collaboration with CEA, the University of Rome 'La Sapienza' investigated a new core model with the aim at increasing the performances of the reference one, by extending the burn-up to 60 GWD/t in the case of multi-loading strategy and investigating the characteristics and limitations of a 'once-through' option, in order to enhance the proliferation resistance. In the first part of this paper, the objectives of this study and the methods of calculation are briefly described, while in the second part the calculation results are presented. (authors)

Golfier, H. [DM2S/SERMA/LCA, CEA Saclay, 91191 Gif-sur-Yvette Cedex (France); Caterino, S. [Univ. of Rome La Sapienza, Dept. of Nuclear Engineering and Energy Conversion, Corso Vittorio Emanuele II, 244 00186 Rome (Italy); Poinot, C.; Delpech, M.; Mignot, G. [DM2S/SERMA/LCA, CEA Saclay, 91191 Gif-sur-Yvette Cedex (France); Naviglio, A.; Gandini, A. [Univ. of Rome La Sapienza, Dept. of Nuclear Engineering and Energy Conversion, Corso Vittorio Emanuele II, 244 00186 Rome (Italy)

2006-07-01

207

Knowledge-based system for core operation management of boiling water reactors  

Microsoft Academic Search

A knowledge-based system, which supports designers in the core operation management of boiling water reactors, has been developed. The system consists of (1) the knowledge base to store heuristics of experienced designers in the forms of rules, functions and frames: (2) the inference program with automatic deletion and generation of those data which have to be modified: and (3) rule

Takaharu Fukuzaki; Yutaka Wada; Yasuhiro Kobayashi; Hidefumi Matsuura; Takanori Shimoshige

1988-01-01

208

MAIN FEATURES OF THE CORE MELT STABILIZATION SYSTEM OF THE EUROPEAN PRESSURIZED WATER REACTOR (EPR)  

Microsoft Academic Search

For the European Pressurized Water Reactor (EPR) a fourth level of defense-in-depth has been introduced to limit the consequences of a postulated severe accident with core melting. This requires to strengthen the containment and to implement measures which can either prevent high loads on the containment structures or mitigate effects of severe accidents in a way to maintain the containment

Dietmar Bittermann; Manfred Fischer; Markus Nie

209

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

210

Surface radiation properties between nuclear reactor materials in contact. [LMFBR core disruptive accident  

Microsoft Academic Search

This technical report summarizes a portion of the results on surface radiation properties of reactor materials, which is part of the program on Attenuation of Radiological Consequences From Core Disruptive Accidents by Radiative Transfer. In analyzing radiant energy exchange between the fuel and coolant, it is necessary to know the reflection absorption, and emission characteristics of the fuel and its

S. H. Chan; H. H. Tseng

1978-01-01

211

In-core flux monitoring system for a boiling water reactor  

Microsoft Academic Search

From international nuclear industries fair; Basel, Switzerland (16 Oct ; 1972). A complete incore flux instrumentation system for a boiling water reactor ; is described. This system comprises a pulse channel and Campbell channel, both ; of which are withdrawn from the core after operation, and ninety-six fixed power ; range detectors together with a moving detector for flux scanning

A. C. Bartlett; L. Nelson-Jones

1972-01-01

212

A metal fuel fast reactor core for the Self-Consistent Nuclear Energy System  

Microsoft Academic Search

Feasibility of transmutation of the major long-lived FPs (I, Pd, Tc, Sn, Se, Zr, Cs) while maintaining fuel breeding capability for the Self-Consistent Nuclear Energy System is evaluated based on the actinide recycle metal fuel core of a fast reactor. It is shown that I, Pd, Tc, Sn, Se, and Zr can be transmuted simultaneously by an aid of the

Masaki Saito; Hiroshi Sekimoto

2002-01-01

213

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

214

Effect of Gas Entrainment on Thermal-Hydraulic Performance of Sodium Cooled Reactor Core  

Microsoft Academic Search

With the view to determining whether or not gas entrained in the sodium coolant could cause overheating of a fast reactor core, the following items were studied:1. The effect of gas entrainment on the coolant flow rate and on coolant temperature rise.2. The effect of gas entrainment on the coolant heat transfer coefficient and film temperature drop.Equations were derived to

Masao HORI; A. J. FRIEDLAND

1970-01-01

215

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

Microsoft Academic Search

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

W. B. Scott; D. G. Harrison; R. A. Libby; R. D. Tokarz; R. D. Wooton; R. S. Denning; R. W. Jr. Tayloe

1990-01-01

216

Spring design for use in the core of a nuclear reactor  

DOEpatents

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

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

1993-01-01

217

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

Microsoft Academic Search

An investigation was conducted using a 1.2 MW RF induction heater facility to aid in developing the technology necessary for designing a self critical fissioning uranium plasma core reactor. Pure, high temperature uranium hexafluoride (UF6) was injected into an argon fluid mechanically confined, steady state, RF heated plasma while employing different exhaust systems and diagnostic techniques to simulate and investigate

W. C. Roman

1979-01-01

218

EDF'S PWR Power Plants: Anomalies Concerning the Reactor Core Instrumentation System.  

National Technical Information Service (NTIS)

This report presents the problems of fatigue and leaks found on the internal core instrumentation thimbles of several French PWR power plants, as also the solutions chosen according the reactor has already or not been operating. (ERA citation 12:040969)

1985-01-01

219

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

Microsoft Academic Search

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

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

1981-01-01

220

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

Microsoft Academic Search

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

K. C. Schulz; G. T. Yahr

1995-01-01

221

Neutronics Analysis of an Open-Cycle High-Impulse Gas Core Reactor Concept.  

National Technical Information Service (NTIS)

A procedure was developed to calculate the critical fuel mass, including the effects of propellant pressure, for coaxial-flow gas-core reactors operating at 196,600 newtons thrust and 4400 seconds specific impulse. Data were generated for a range of cavit...

C. L. Whitmarsh

1972-01-01

222

Transition phase of the whole-core demonstration at the Oak Ridge Research Reactor  

SciTech Connect

The transition from operation of the Oak Ridge Research Reactor with high-enrichment uranium (HEU) fuel to operation with low-enrichment uranium (LEU) fuel is nearing completion. The systematics of the replacement of the HEU fuel with the LEU fuel are discussed. The results of the core physics measurements that have been conducted during the transition phase are described.

Hobbs, R.W.; Bretscher, M.M.; Cornella, R.J.; Snelgrove, J.L.

1986-01-01

223

Heterogeneous gas core reactor and dual fluid closed cycle power conversion system  

Microsoft Academic Search

The basic objective of the present research program is the conceptual analysis and preliminary assessment of the Heterogeneous Gas Core Reactor (HGCR) concept and Dual-Fluid power conversion system. This interim report presents a summary of the results obtained to date, and serves as a basis for an early evaluation of the system's nuclear non-proliferating advantages, its enhanced (over LWR's and

N. J. Diaz; E. T. Dugan; C. C. Oliver; R. A. Gater

1977-01-01

224

Startup and control of out-of-core thermionic space reactors.  

National Technical Information Service (NTIS)

An analysis of out-of-core thermionic space reactor (OTR) startup and control has been performed. The reference ITR chosen for this study is a 75 kWt version of the STAR-C (GA Technologies 1987). The applicability of point kinetics was first verified for ...

M. G. Houts D. D. Lanning

1991-01-01

225

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

Microsoft Academic Search

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

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

2002-01-01

226

Metal fueled long life fast reactor cores with inherent safety features  

SciTech Connect

A large fast reactor core concept is proposed that has inherent safety characteristics against both the Unprotected Loss of Flow (ULOF) event and the Unprotected Transient of Over-Power (UTOP) event, where commonly used zirconium alloy metal fuel (U-Pu- Zr) is adopted to achieve a long life cycle length up to 5 years. The burn-up reactivity of the core which is equivalent to the maximum insertion reactivity in the UTOP due to the control rod run-out event at the rated power, is reduced to less than 1 $ by introducing minor actinides to the fuel, while the sodium void reactivity is suppressed to be negative by applying a step core concept, where the inner core height is lower than the outer core height, and by deleting the upper axial blanket. (authors)

Yokoyama, Tsugio [AITEL Corporation: 8-Shinsugita-cho, Isogo-ku Yokohama 235-8523 (Japan); Ninokata, Hisashi; Endo, Hiroshi [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology 2-12-1 O-okayama, Meguro-ku, Tokyo, 152-8550 (Japan)

2007-07-01

227

Gamma-thermometer-based reactor-core liquid-level detector. [PWR  

SciTech Connect

A system is provided which employs a modified gamma thermometer for determining the liquid coolant level within a nuclear reactor core. The gamma thermometer which normally is employed to monitor local core heat generation rate (reactor power), is modified by thermocouple junctions and leads to obtain an unambiguous indication of the presence or absence of coolant liquid at the gamma thermometer location. A signal processor generates a signal based on the thermometer surface heat transfer coefficient by comparing the signals from the thermocouples at the thermometer location. The generated signal is a direct indication of loss of coolant due to the change in surface heat transfer when coolant liquid drops below the thermometer location. The loss of coolant indication is independent of reactor power at the thermometer location. Further, the same thermometer may still be used for the normal power monitoring function.

Burns, T.J.

1981-06-16

228

METHOD OF FABRICATING A URANIUM-ZIRCONIUM HYDRIDE REACTOR CORE  

DOEpatents

A method is described of evenly dispersing uranlum metal in a zirconium hydride moderator to produce a fuel element for nuclear reactors. According to the invention enriched uranium hydride and zirconium hydride powders of 200 mesh particle size are thoroughly admixed to form a mixture containing 0.1 to 3% by weight of U/sup 235/ hydride. The mixed powders are placed in a die and pressed at 100 tons per square inch at room temperature. The resultant compacts are heated in a vacuum to 300 deg C, whereby the uranium hydride deoomposes into uranium metal and hydrogen gas. The escaping hydrogen gas forms a porous matrix of zirconium hydride, with uramum metal evenly dispersed therethrough. The advantage of the invention is that the porosity and uranium distribution of the final fuel element can be more closely determined and controlled than was possible using prior methods of producing such fuel ele- ments.

Weeks, I.F.; Goeddel, W.V.

1960-03-22

229

Critical evaluation of molybdenum and its alloys for use in space reactor core heat pipes  

SciTech Connect

The choice of pure molybdenum as the prime candidate material for space reactor core heat pipes is examined, and the advantages and disadvantages of this material are brought into focus. Even though pure molybdenum heat pipes have been built and tested, this metal's high ductile-brittle transition temperature and modest creep strength place significant design restrictions on a core heat pipe made from it. Molybdenum alloys are examined with regard to their promise as potential replacements for pure molybdenum. The properties of TZM and molybdenum-rhenium alloys are examined, and it appears that Mo-Re alloys with 10 to 15 wt % rhenium offer the most advantage as an alternative to pure molybdenum in space reactor core heat pipes.

Lundberg, L.B.

1981-01-01

230

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

231

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

232

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

233

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

234

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

235

Methods and performance of the three-dimensional pressurized water reactor core dynamics SIMTRAN on-line code  

Microsoft Academic Search

New reactor physics and computation methods have been developed in the three-dimensional pressurized water reactor (PWR) core dynamics SIMTRAN code for on-line surveillance and prediction. The accuracy of the coupled neutronic thermal-hydraulic solution is improved, and its scope is extended to provide, mainly, the calculation of the fission reaction rates at the in-core mini detectors, the responses at the ex-core

J. M. Aragones; C. Ahnert; O. Cabellos

1996-01-01

236

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

SciTech Connect

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

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

1988-04-01

237

Flow Test for the Full Scale Core Mockup of the KUHFR (Kyoto University High Flux Reactor), (1). Velocity Distribution and Pressure Loss in the Core.  

National Technical Information Service (NTIS)

The flow test using a full scale mockup of the core of the Kyoto University high flux reactor (KUHFR) has been performed as a part of the research and development of the KUHFR which is planned by the Research Reactor Institute, Kyoto University. The press...

K. Mishima Y. Araki Y. Ishikawa

1981-01-01

238

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

SciTech Connect

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

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

2006-07-01

239

Nuclear design of a vapor core reactor for space nuclear propulsion  

NASA Astrophysics Data System (ADS)

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

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

1993-01-01

240

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

SciTech Connect

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

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

2012-07-01

241

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

242

Dominance of convective heat transport in the core of TFTR (Tokamak Fusion Test Reactor) supershot plasmas  

SciTech Connect

Using perturbations in electron density and temperature induced by small Helium gas puffs in TFTR (Tokamak Fusion Test Reactor) the dominance of convective heat transport in the core (r/a < 0.4) of supershot plasmas has been demonstrated in a new way. The TRANSP transport code was used to calculate the time-dependent particle and heat fluxes. Perturbations in the calculated convective and total electron heat fluxes were compared. They demonstrate that the conductive component decreases moving into the supershot core, and the convective component dominates in the supershot core. These results suggest a different transport drive in the supershot core compared to that in the rest of the supershot plasma.

Kissick, M.W.; Efthimion, P.C.; Mansfield, D.K.; Callen, J.D.; Bush, C.E.; Park, H.K.; Schivell, J.; Synakowski, E.J.; Taylor, G.

1993-08-01

243

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

244

Sensitivity Degradation Characteristics of In-core Neutron Detector for Heavy Water Reactor, Fugen NPP  

SciTech Connect

Fugen nuclear power plant is a 165 MWe, heavy water-moderated, boiling light water-cooled, pressure tube-type reactor developed by JNC, which is the world's first thermal neutron power reactor to utilize mainly Uranium and Plutonium mixed oxide (MOX) fuel. Fugen has been loaded a total of 726 MOX fuel assemblies since the initial core in 1978. Each in-core neutron detector assembly of Fugen composed of four Local Power Monitors (LPM) is located at sixteen positions in the area of heavy water moderator in the core and monitors its power distribution during operation. The thermal neutron flux of Fugen is relatively higher than that of Boiling Water Reactor (BWR), therefore LPM, which is comprised of a fission chamber, degrades more quickly than that of BWR. An Improved Long-life LPM (LLPM) pasted inner surface wall of the chamber with {sup 234}U/{sup 235}U at a ratio of 4 to 1 had been developed through the irradiation test at Japan Material Test Reactor (JMTR). The {sup 234}U is converted to {sup 235}U with absorption of neutron, and compensates the consumption of {sup 235}U. LPM has been loaded to the initial core of Fugen since 1978. JNC had evaluated its sensitivity degradation characteristics through the accumulated irradiation data and the parametric survey for {sup 234}U and {sup 235}U. Based on the experience of evaluation for sensitivity degradation, JNC has applied shuffling operation of LPM assemblies during an annual inspection outage to reduce the operating cost. This operation realizes the reduction of replacing number of LPM assemblies and volume of radioactive waste. This paper describes the sensitivity degradation characteristics of in-core neutron detector and the degradation evaluation methods established in Fugen. (authors)

Tsuyoshi Okawa; Naoyuki Yomori [Japan Nuclear Cycle Development Institute (Japan)

2002-07-01

245

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

2013-01-01

246

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

247

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

NASA Astrophysics Data System (ADS)

In-core temperature measurement is a critical issue for the safe operation of nuclear reactors. Classical thermocouples require shielded connections and are known to drift under high neutron fluence. As an alternative, we propose to take advantage of the multiplexing capabilities of fiber Bragg grating (FBG) temperature sensors. Our experiments show that sensitivity to radiation depends on both the radiation field and the grating characteristics. For some FBGs installed in an air-cooled graphite- moderated nuclear reactor the difference between the measurements and the readings of calibrated backup thermocouples was within the measurement uncertainty. In the worst case, the difference saturated after 30 h of reactor operation at about 5 degree(s)C. To reach megagray per hour level gamma-dose rates and 1019 neutron/cm2 fluences, we irradiated multiplexed FBG sensors in a material testing nuclear reactor. At room temperature, FBG temperature sensors can survive in such radiation conditions, but at 90 degree(s)C a severe degradation is observed. We evidence the possibility to use FBG sensing technology for in-core monitoring of nuclear reactors with specific care under well-specified conditions.

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

2002-06-01

248

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

249

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

250

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

SciTech Connect

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

Samim Anghaie

2002-08-13

251

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.

252

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

253

Development of a detailed core flow analysis code for prismatic fuel reactors  

SciTech Connect

The development of a computer code for the analysis of the detailed flow of helium in prismatic fuel reactors is reported. The code, called BYPASS, solves, a finite difference control volume formulation of the compressible, steady state fluid flow in highly cross-connected flow paths typical of the Modular High-Temperature Gas Cooled Reactor (MHTGR). The discretization of the flow in a core region typically considers the main coolant flow paths, the bypass gap flow paths, and the crossflow connections between them. 16 refs., 5 figs.

Bennett, R.G.

1990-01-01

254

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

255

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

256

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

SciTech Connect

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

Lee, Bom Soon

1994-06-01

257

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

258

A Metal Fuel Core Concept for 1000 MWt Advanced Burner Reactor  

SciTech Connect

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

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

2007-07-01

259

Medical isotope production: A new research initiative for the Annular Core Research Reactor  

SciTech Connect

An investigation has been performed to evaluate the capabilities of the Annular Core Research Reactor and its supporting Hot Cell Facility for the production of {sup 99}Mo and its separation from the fission product stream. Various target irradiation locations for a variety of core configurations were investigated, including the central cavity, fuel and reflector locations, and special target configurations outside the active fuel region. Monte Carlo techniques, in particular MCNP using ENDF B-V cross sections, were employed for the evaluation. The results indicate that the reactor, as currently configured, and with its supporting Hot Cell Facility, would be capable in meeting the current US demand if called upon. Modest modifications, such as increasing the capacity of the external heat exchangers, would permit significantly higher continuous power operation and even greater {sup 99}Mo production ensuring adequate capacity for future years.

Coats, R.L.; Parma, E.J.

1993-12-31

260

A reactor core/containment status evaluation flowchart for determining protective actions in emergencies  

SciTech Connect

In the event of an emergency at a power reactor station, there might not be adequate time or sufficient data to fully assess radiological implications and make protective action recommendations based on projected population exposures. Thus, decision-making guidance is needed that is based on readily available plant indicators, not just on time-consuming dose calculations. In the United States, this guidance must be compatible with the recommended by the Nuclear Regulatory Commission and the Environmental Protection Agency, and it must include predetermined, measurable, site-specific parameters for assessing conditions in the reactor core and containment. The preparation of this real time guidance calls for the selection of suitable parameters and the determination of the values for these parameters that will correspond to different levels of protective action. This process is illustrated in this paper by selecting parameters and determining appropriate values for constructing a Core/Containment Status Evaluation Flowchart for an example power plant.

Glissman, M.A. (Impell Corp., Lincolnshire, IL (United States))

1988-02-01

261

IRI: An international Rawer initiative  

NASA Astrophysics Data System (ADS)

This paper was presented during the special session that was held at the 1993 International Reference Ionosphere (IRI) Workshop in honor of Karl Rawer's 80th birthday. It retraces the steps that led from the start of the IRI project to the present edition of the model highlighting the important role that the honoree played in guiding IRI from infancy to maturity. All summary view graphs are reproduced at the end of the article.

Bilitza, D.

1995-02-01

262

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

263

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

264

Gas Core Reactor with Magnetohydrodynamic Power System and Cascading Power Cycle  

SciTech Connect

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

Smith, Blair M.; Anghaie, Samim [University of Florida (United States)

2004-03-15

265

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

Microsoft Academic Search

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

PARVATHAREDDY B. REDDY; PEDDAPULLAIAH SANNUTI

1975-01-01

266

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

Microsoft Academic Search

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

D. Grigoriadis; M. Varvayanni; N. Catsaros; E. Stakakis

2008-01-01

267

Modular assembly for supporting, straining, and directing flow to a core in a nuclear reactor  

DOEpatents

A reactor core support arrangement for supporting, straining, and providing fluid flow to the core and periphery of a nuclear reactor during normal operation. A plurality of removable inlet modular units are contained within permanent liners in the lower supporting plate of the reactor vessel lower internals. During normal operation (1) each inlet modular unit directs main coolant flow to a plurality of core assemblies, the latter being removably supported in receptacles in the upper portion of the modular unit and (2) each inlet modular unit may direct bypass flow to a low pressure annular region of the reactor vessel. Each inlet modular unit may include special fluid seals interposed between mating surfaces of the inlet modular units and the core assemblies and between the inlet modular units and the liners, to minimize leakage and achieve an hydraulic balance. Utilizing the hydraulic balance, the modular units are held in the liners and the assemblies are held in the modular unit receptacles by their own respective weight. Included as part of the permanent liners below the horizontal support plate are generally hexagonal axial debris barriers. The axial debris barriers collectively form a bottom boundary of a secondary high pressure plenum, the upper boundary of which is the bottom surface of the horizontal support plate. Peripheral liners include radial debris barriers which collectively form a barrier against debris entry radially. During normal operation primary coolant inlet openings in the liner, below the axial debris barriers, pass a large amount of coolant into the inlet modular units, and secondary coolant inlet openings in the portion of the liners within the secondary plenum pass a small amount of coolant into the inlet modular units. The secondary coolant inlet openings also provide alternative coolant inlet flow paths in the unlikely event of blockage of the primary inlet openings. The primary inlet openings have characteristics which limit the entry of debris and minimize the potential for debris entering the primary inlets blocking the secondary inlets from inside the modular unit.

Pennell, William E. (Greensburg, PA)

1977-01-01

268

Fuel and core testing plan for a target fueled isotope production reactor.  

SciTech Connect

In recent years there has been an unstable supply of the critical diagnostic medical isotope 99Tc. Several concepts and designs have been proposed to produce 99Mo the parent nuclide of 99Tc, at a commercial scale sufficient to stabilize the world supply. This work lays out a testing and experiment plan for a proposed 2 MW open pool reactor fueled by Low Enriched Uranium (LEU) 99Mo targets. The experiments and tests necessary to support licensing of the reactor design are described and how these experiments and tests will help establish the safe operating envelop for a medical isotope production reactor is discussed. The experiments and tests will facilitate a focused and efficient licensing process in order to bring on line a needed production reactor dedicated to supplying medical isotopes. The Target Fuel Isotope Reactor (TFIR) design calls for an active core region that is approximately 40 cm in diameter and 40 cm in fuel height. It contains up to 150 cylindrical, 1-cm diameter, LEU oxide fuel pins clad with Zircaloy (zirconium alloy), in an annular hexagonal array on a {approx}2.0 cm pitch surrounded, radially, by a graphite or a Be reflector. The reactor is similar to U.S. university reactors in power, hardware, and safety/control systems. Fuel/target pin fabrication is based on existing light water reactor fuel fabrication processes. However, as part of licensing process, experiments must be conducted to confirm analytical predictions of steady-state power and accident conditions. The experiment and test plan will be conducted in phases and will utilize existing facilities at the U.S. Department of Energy's Sandia National Laboratories. The first phase is to validate the predicted reactor core neutronics at delayed critical, zero power and very low power. This will be accomplished by using the Sandia Critical Experiment (CX) platform. A full scale TFIR core will be built in the CX and delayed critical measurements will be taken. For low power experiments, fuel pins can be removed after the experiment and using Sandia's metrology lab, relative power profiles (radially and axially) can be determined. In addition to validating neutronic analyses, confirming heat transfer properties of the target/fuel pins and core will be conducted. Fuel/target pin power limits can be verified with out-of-pile (electrical heating) thermal-hydraulic experiments. This will yield data on the heat flux across the Zircaloy clad and establish safety margin and operating limits. Using Sandia's Annular Core Research Reactor (ACRR) a 4 MW TRIGA type research reactor, target/fuel pins can be driven to desired fission power levels for long durations. Post experiment inspection of the pins can be conducted in the Auxiliary Hot Cell Facility to observe changes in the mechanical properties of the LEU matrix and burn-up effects. Transient tests can also be conducted at the ACRR to observe target/fuel pin performance during accident conditions. Target/fuel pins will be placed in double experiment containment and driven by pulsing the ACRR until target/fuel failure is observed. This will allow for extrapolation of analytical work to confirm safety margins.

Coats, Richard Lee; Dahl, James J.; Parma, Edward J., Jr.

2010-12-01

269

Ordinal measures for iris recognition.  

PubMed

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 rather than precise measurements of iris image structures. Such a representation may lose some image-specific information, but it achieves a good trade-off between distinctiveness and robustness. We show that ordinal measures are intrinsic features of iris patterns and largely invariant to illumination changes. Moreover, compactness and low computational complexity of ordinal measures enable highly efficient iris recognition. Ordinal measures are a general concept useful for image analysis and many variants can be derived for ordinal feature extraction. In this paper, we develop multilobe differential filters to compute ordinal measures with flexible intralobe and interlobe parameters such as location, scale, orientation, and distance. Experimental results on three public iris image databases demonstrate the effectiveness of the proposed ordinal feature models. PMID:19834142

Sun, Zhenan; Tan, Tieniu

2009-12-01

270

Optimizing a three-element core design for the advanced neutron source reactor  

SciTech Connect

The source of neutrons in the proposed Advanced Neutron Source facility is a multipurpose research reactor providing 5 to 10 times the flux, for neutron beams, of the best existing facilities. The project team constrained the design with the {open_quotes}no new inventions rule,{close_quotes} which states that the design should not rely on the development of new technology to meet the minimum design criteria (although research and development that can lead to further major improvements beyond the minimum requirements is encouraged). The baseline design for the reactor core, based on this objective and within this constraint, was an assembly of two annular fuel elements, similar to those used in the high-flux reactors at Oak Ridge and Grenoble, containing highly enriched (93%) uranium silicide particles. Subsequently, the U.S. Department of Energy commissioned a study of the impact on performance and on cost of using medium - or low-enriched uranium. In the course of that work, a three-element core design was studied as a means to provide extra volume to accommodate the additional uranium compound required when the fissionable {sup 235}U has to be diluted with {sup 238}U to reduce the enrichment. This paper describes the design and optimization of that three-element core.

West, C.D. [Martin Marietta Energy System, Oak Ridge, TN (United States)

1995-12-31

271

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

SciTech Connect

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

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

2008-01-21

272

Dosimetry assessments for the reactor pressure vessel and core barrel in UK PWR plant  

SciTech Connect

Specimens for the Sizewell B reactor pressure vessel (RPV) inservice steels surveillance program are irradiated inside eight capsules located within the reactor pressure vessel and loaded prior to commissioning. The periodic removal of these capsules and testing of their contents provides material properties data at intervals during the lifetime of the plant. Neutron activation measurements and radiation transport calculations play an essential role in assessing the neutron exposure of the specimens and RPV. Following the most recent withdrawal, seven capsules have now been removed covering nine cycles of reactor operation. This paper summarizes the dosimetry results of the Sizewell B surveillance program obtained to date. In addition to an overview of the calculational methodology it includes a review of the measurements. Finally, it describes an extension of the methodology to provide dosimetry recommendations for the core barrel and briefly discusses the results that were obtained. (authors)

Thornton, D.A.; Allen, D.A.; Huggon, A.P.; Picton, D.J.; Robinson, A.T.; Steadman, R.J. [Serco, Rutherford House, Quedgeley, Gloucester, Gl2 4NF (United Kingdom); Seren, T.; Lipponen, M.; Kekki, T. [VTT, Technical Research Centre of Finland, Otakaari 3 K, P.O. BOX 1000, Espoo, FI-02044 (Finland)

2011-07-01

273

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

SciTech Connect

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

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

2011-07-01

274

Thermal hydraulic response of the Advanced Neutron Source Reactor to piping breaks near the core region  

SciTech Connect

This paper describes the application of the RELAP5 thermal hydraulic code to a highly subcooled, plate type reactor typical of many research and production reactor systems. The specific system modeled is the latest design of the Advanced Neutron Source Reactor (ANSR). A discussion of the model as well as the results from several loss-of-coolant accident (LOCA) scenarios is included. The results indicate that this system responds to these accidents by a very rapid depressurization (over a few milliseconds) followed by a pressure recovery due to fluid inertia. In addition, the effect of including a gas pressurized accumulator in the system is addressed. The results show that tracking the pressure response of the system over these short time scales will be a key to accurately predicting the thermal response of the core of the reactor. Further, the break time scale as well as the time scale of the thermal response of the core, presently treated conservatively, will be additional important areas of study. 22 refs.

Chen, N.C.J.; Williams, P.T.; Yoder, G.L.

1992-01-01

275

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

276

Thermohydraulics in a high-temperature gas-cooled reactor prestressed-concrete reactor vessel during unrestricted core-heatup accidents  

Microsoft Academic Search

The hypothetical accident considered for siting considerations in High Temperature Gas-Cooled Reactors (HTGR) is the so called Unrestricted Core Heatup Accident (UCHA), in which all forced circulation is lost at initiation, and none of the auxillary cooling loops can be started. The result is a gradual slow core heatup, extending over days. Whether the liner cooling system (LCS) operates during

P. G. Kroeger; J. Colman; K. Araj

1983-01-01

277

Results of Reactor Materials Experiments Investigating 2-D Core-Concrete Interaction and Debris Coolability  

SciTech Connect

The OECD-sponsored Melt Coolability and Concrete Interaction (MCCI) program is conducting reactor materials experiments and associated analysis to achieve the following objectives: 1) resolution of the ex-vessel debris coolability issue, and 2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two objectives will demonstrate the efficacy of severe accident management guidelines for existing plants and provide the technical basis for better containment designs of future plants. With respect to the second objective, there remain uncertainties in the models that evaluate the lateral vs. axial power split during core-concrete interaction because of a lack of truly two-dimensional experiment data. As a first step in bridging this gap, a large scale Core-Concrete Interaction experiment (CCI-1) has been conducted as part of the MCCI program. This test investigated the interaction of a 400 kg core-oxide melt with a crucible made of siliceous concrete along two walls and the base. The two remaining walls were made of non-ablative magnesium oxide. The initial phase of the test was conducted under dry conditions. After a predefined ablation depth was achieved, the cavity was flooded to obtain data on the coolability of a core melt after core-concrete interaction has progressed for some time. This paper provides a description of the test facility and an overview of results from this test. (authors)

Farmer, M. T.; Lomperski, S. [Nuclear Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States); Basu, S. [U.S. Nuclear Regulatory Commission, MS-T10K8, 11545 Rockville Pike, Rockville, MD 20852 (United States)

2004-07-01

278

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

279

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

280

Using Non-Orthogonal Iris Images for Iris Recognition.  

National Technical Information Service (NTIS)

The iris is the colored portion of the eye that surrounds the pupil and controls the amount of light that can enter the eye. The variations within the patterns of the iris are unique between eyes, which allows for accurate identification of an individual....

R. M. Gaunt

2006-01-01

281

Simplified interassembly heat transfer model for the analysis of liquid-metal fast breeder reactor core restraint systems  

Microsoft Academic Search

A simplified interassembly heat transfer model has been developed to satisfy liquid-metal fast breeder reactor core restraint system analysis needs that explicitly treats steady-state intra-assembly and interassembly heat transfer in core assemblies. The intra-assembly heat transfer inside reactor assemblies is modeled based on application of the subchannel concept together with the use of bulk parameters for coolant velocity and coolant

1979-01-01

282

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

283

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

284

IRIS Process (2004-2008)  

EPA Science Inventory

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

285

MORE ON IRIS YELLOW SPOT  

Technology Transfer Automated Retrieval System (TEKTRAN)

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

286

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

287

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

288

Results from an international intercomparison of fundamental mode benchmark calculations for steam ingress into gas-cooled fast reactor cores  

Microsoft Academic Search

Steam ingress into a gas-cooled fast reactor (GCFR) core may lead to reactivity effects that are undesirable from the point of view of reactor safety. Unfortunately, the amount of reactivity increase caused by a certain steam concentration is usually subject to considerable uncertainty, as has become evident by occasional comparisons between various laboratories for specific examples. Therefore, some time ago,

Kiefhaber

1982-01-01

289

Development and steady state level experimental validation of TASS\\/SMR core heat transfer model for the integral reactor SMART  

Microsoft Academic Search

An advanced integral pressurized water reactor (PWR) of a small size (?330MWt) is being developed by the Korea Atomic Energy Research Institute (KAERI). The purposes of the reactor are a sea water desalination and an electricity generation. To enhance its safety, many advanced design concepts are introduced such as a passive residual removal system and a low power density core.

Seong Wook Lee; Soo Hyung Kim; Young Jong Chung

2009-01-01

290

New methods in iris recognition.  

PubMed

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 trigonometry and projective geometry, allowing off-axis gaze to be handled by detecting it and "rotating" the eye into orthographic perspective; 3) statistical inference methods for detecting and excluding eyelashes; and 4) exploration of score normalizations, depending on the amount of iris data that is available in images and the required scale of database search. Statistical results are presented based on 200 billion iris cross-comparisons that were generated from 632500 irises in the United Arab Emirates database to analyze the normalization issues raised in different regions of receiver operating characteristic curves. PMID:17926700

Daugman, John

2007-10-01

291

Prediction, analysis and solution of the flow inversion phenomenon in a typical MTR-reactor with upward core cooling  

Microsoft Academic Search

Research reactors of power greater than 20MW are usually designed to be cooled with upward coolant flow direction inside the reactor core. This is mainly to prevent flow inversion problems following a pump coast down. However, in some designs and under certain operating conditions, flow inversion phenomenon is predicted. In the present work, the best-estimate Material Testing Reactors Thermal-Hydraulic Analysis

Salah El-Din El-Morshedy

2011-01-01

292

SIMULATE-E: A Nodal Core-Analysis Program for Light-Water Reactors. Computer Code User's Manual.  

National Technical Information Service (NTIS)

This report contains the methods descriptions and user's manual for the light-water reactor nodal core-analysis computer program, SIMULATE-E. SIMULATE yields the distribution of fission power within the core in three-dimensional detail while including the...

D. M. Ver Planck W. R. Cobb R. S. Borland P. L. Versteegen

1983-01-01

293

Review of the state of criticality of the Three Mile Island Unit 2 core and reactor vessel  

Microsoft Academic Search

The events during the early hours of the Three Mile Island Unit 2 (TMI-2) accident on March 28, 1979 caused the fuel in the reactor core to crumble or disintegrate, and then subside into a rubble structure more compact that its normal configuration. The present height of the core is about seven feet, five feet less than its normal configuration

Stratton

1987-01-01

294

DNBR analyses under steady-state and accident conditions for a double-flat-core high conversion light water reactor.  

National Technical Information Service (NTIS)

A double-flat-core high conversion light water reactor (HCLWR-JDF1) has been developed at JAERI aiming at better fuel utilization and higher safety margin. The HCLWR has two pancake type cores piled up with lower, internal and upper axial blankets. Fuel r...

T. Iwamura T. Suemura T. Okubo F. Hiraga Y. Murao

1990-01-01

295

Phenomena occurring in the reactor coolant system during severe core damage accidents  

SciTech Connect

The reactor coolant system (RCS) of a nuclear power plant consists of the reactor pressure vessel and the piping and associated components that are required for the continuous circulation of the coolant which is used to maintain thermal equilibrium throughout the system. In the event of an accident, the RCS also serves as one of several barriers to the escape of radiotoxic material into the biosphere. In contrast to normal operating conditions, severe core damage accidents are characterized by significant temporal and spatial variations in heat and mass fluxes, and by eventual geometrical changes within the RCS. Furthermore, the difficulties in describing the system in the severe accident mode are compounded by the occurrence of chemical reactions. These reactions can influence both the thermal and the mass transport behavior of the system. In addition, behavior of the reactor vessel internals and of materials released from the core region (especially the radioactive fission products) in the course of the accident likewise become of concern to the analyst. This report addresses these concerns. 9 refs., 1 tab.

Malinauskas, A.P.

1989-01-01

296

Gas Core Reactor-MHD Power System with Cascading Power Cycle  

SciTech Connect

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

Smith, Blair M.; Anghaie, Samim; Knight, Travis W. [Innovative Nuclear Space Power and Propulsion Institute, University of Florida, PO Box 116502, Gainesville, FL, 32611 (United States)

2002-07-01

297

Primary cysts of the iris.  

PubMed Central

This paper has presented the author's experience with the evaluation and follow-up of 62 patients with primary cysts of the iris. On the basis of these observations, a classification of iris cyst is proposed. Accordingly, primary iris cysts are divided into epithelial and stromal categories, each having different clinical characteristics. Epithelial cysts arise between the pigmented epithelial layers of the iris and occur at the pupillary margin (central cysts), in the mid-portion of the iris (midzonal cysts) or, more commonly, in the iridociliary sulcus (peripheral cysts). In some cases, the cysts apparently break free from their epithelial attachment and migrate into the anterior chamber of vitreous chamber (dislodged cysts). Primary stromal cysts occur within the iris stroma and are not directly continuous with the posterior epithelium. They apparently arise from ectopic surface epithelium which is trapped in the iris during embryologic development. A study of the natural course and complications of these lesions has shown that the great majority of primary iris cysts, particularly those which arise from the iris pigment epithelial layer, are stationary lesions which rarely progress or cause visual complications. This finding is contradictory to the contemporary belief of certain authorities who stress that many such lesions lead to severe complications with blindness and loss of the eye. The natural course of primary epithelial cysts differs from that of secondary iris cysts which follow surgical or non-surgical trauma. The latter lesions do frequently enlarge and lead to severe complications such as inflammation and glaucoma. Images FIGURE 1 FIGURE 2 A FIGURE 2 B FIGURE 3 FIGURE 4 A FIGURE 4 B FIGURE 5 FIGURE 6 A FIGURE 6 B FIGURE 7 FIGURE 8 FIGURE 9 A FIGURE 9 B FIGURE 10 A FIGURE 10 B FIGURE 10 C FIGURE 10 D FIGURE 11 FIGURE 12 FIGURE 13 FIGURE 14 A FIGURE 14 B FIGURE 15 FIGURE 16 FIGURE 17 FIGURE 18

Shields, J A

1981-01-01

298

Improvement on the prediction accuracy of transmutation properties for fast reactor core using the minor actinides irradiation test data on the Joyo MK-II CORE  

SciTech Connect

For a validation of MA nuclear data and improvement on the prediction accuracy of MA transmutation properties in fast reactor cores, MA sample irradiation test data of Joyo were utilized. Adopting MA cross-sections in JENDL-3.3, result of their evaluations showed good agreement with experimental data. Further, the present study clarified that utilization of these data with cross-section adjustment technique has a potential to reduce uncertainty of MA transmutation properties in fast reactor cores to less than half. (author)

Sugino, Kazuteru [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency - JAEA, 4002, Narita-Cho, Oarai-Machi, Higashi-Ibaraki-Gun, Ibaraki, 311-1393 (Japan)

2007-07-01

299

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

300

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

301

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

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

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

2010-09-01

302

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

SciTech Connect

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

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

1996-02-01

303

Eddy current position indicating apparatus for measuring displacements of core components of a liquid metal nuclear reactor  

DOEpatents

Apparatus for measuring displacements of core components of a liquid metal fast breeder reactor by means of an eddy current probe. The active portion of the probe is located within a dry thimble which is supported on a stationary portion of the reactor core support structure. Split rings of metal, having a resistivity significantly different than sodium, are fixedly mounted on the core component to be monitored. The split rings are slidably positioned around, concentric with the probe and symmetrically situated along the axis of the probe so that motion of the ring along the axis of the probe produces a proportional change in the probes electrical output.

Day, Clifford K. (Richland, WA); Stringer, James L. (Richland, WA)

1977-01-01

304

Status of axial heterogeneous liquid-metal fast breeder reactor core design studies and research and development  

SciTech Connect

The current status of axial heterogeneous core (AHC) design development in Japan, which consists of an AHC core design in a pool-type demonstration fast breeder reactor (DFBR) and research and development activities supporting AHC core design, is presented. The DFBR core design objectives developed by The Japan Atomic Power Company include (a) favorable core seismic response, (b) core compactness, (c) high availability, and (d) lower fuel cycle cost. The AHC concept was selected as a reference pool-type DFBR core because it met these objectives more suitably than the homogeneous core (HOC). The AHC core layouts were optimized emphasizing the reduction of the burnup reactivity swing, peak fast fluence, and power peaking. The key performance parameters resulting from the AHC, such as flat axial power/flux distribution, lower peak fast fluence, lower burnup reactivity swing, etc., were evaluated in comparison with the HOC. The critical experiments at the Japan Atomic Energy Research Institute's Fast Critical Assembly facility demonstrate the key AHC performance characteristics. The large AHC engineering benchmark experiments using the zero-power plutonium reactor and the AHC fuel pin irradiation test program using the JOYO reactor are also presented.

Nakagawa, H.; Inagaki, T.; Yoshimi, H.; Shirakata, K.; Watari, Y.; Suzuki, M.; Inoue, K.

1988-11-01

305

Exploring multispectral iris recognition beyond 900nm  

Microsoft Academic Search

Most iris recognition systems acquire images of the eye in the 700 nm-900 nm range of the electromagnetic spectrum. In this work, the iris is examined at wavelengths beyond 900 nm. The purpose is to understand the iris structure at longer wavelengths and to determine the possibility of performing cross-spectral iris matching. An acquisition system is first designed for imaging

Arun Ross; Raghunandan Pasula; Lawrence Hornak

2009-01-01

306

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

307

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

308

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

309

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

310

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

311

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

312

IRI annual report 1989.  

National Technical Information Service (NTIS)

In this annual report of the Dutch Interfacultary Reactor Institute, summary reports are presented of current research and teaching activities during 1989 of the departments radiochemistry, radiation chemistry, radiation physics and reactor physics, opera...

1990-01-01

313

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

314

Transport of recycled deuterium to the plasma core in the Tokamak Fusion Test Reactor  

SciTech Connect

Fueling of the plasma core by recycling in the Tokamak Fusion Test Reactor [Phys. Plasmas {bold 2}, 2176 (1995)] has been studied. In plasmas fueled by deuterium recycled from the limiter and tritium-only neutral beam injection, the DT neutron rate provides a measure of the deuterium influx into the core plasma. A reduced influx with plasmas using lithium pellet conditioning and with plasmas of reduced major (and minor) radius is found. Modeling with the DEGAS [D. P. Stotler {ital et al.}, Phys. Plasmas {bold 3}, 4084 (1996)] neutrals code shows that the dependence on radius can be related to the penetration of neutrals through the scrape-off layer. {copyright} {ital 1998 American Institute of Physics.}

Skinner, C.H.; Bell, M.G.; Budny, R.V.; Jassby, D.L.; Park, H.; Ramsey, A.T.; Stotler, D.P.; Strachan, J.D. [Plasma Physics Laboratory, Princeton University, P.O. Box 451, Princeton, New Jersey08543 (United States)

1998-04-01

315

Iris image acquirement and preprocessing in biometrics  

NASA Astrophysics Data System (ADS)

As a biometrics, iris recognition is becoming an active topic in recent years. Due to constriction and dilation of iris sphincter muscle affected by illumination, iris images should be captured under steady illumination. The principles of our homemade iris capture device and real product are introduced first. This device can be used to capture fine iris images and store them into the computer. Due to changes of head orientation and binocular vergence at different iris acquirement time, a preprocessing method including iris location, normalization, image enhancement and rotation correction is then applied to achieve rotation, shift and scale invariance. Finally the correlation measurement is used to evaluate the performance of this processing.

Ming, Xing; Li, Zhihui; Liu, Yuanning

2005-12-01

316

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

317

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

318

The scalability of OTR (Out-of-core Thermionic Reactor) space nuclear power systems  

NASA Astrophysics Data System (ADS)

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.

Gallup, Donald R.

1990-03-01

319

Iris Localization via Pulling and Pushing  

Microsoft Academic Search

Iris localization is a critical module in iris recogni- tion because it defines the inner and outer boundaries of iris region used for feature analysis. State-of-the-art iris localization methods need to implement a brute- force search of the large parameter space, which is time-consuming and sensitive to noises. This paper proposes a novel iris localization method based on a spring

Zhaofeng He; Tieniu Tan; Zhenan Sun

2006-01-01

320

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

321

Preliminary Probabilistic Safety Assessment of the IRIS Plant  

SciTech Connect

A preliminary level-1 probabilistic safety assessment of the IRIS plant has been performed. The first focus is on five internal initiating events, such as primary system break (loss-of-coolant accident and steam generator tube rupture) and transients (secondary system line break and loss-of-off-site power). In this study, the event tree for each initiating event was developed and the fault tree analysis of the event tree headings was carried out. In particular, since one of the IRIS safety systems, the passive emergency heat removal system, is unique to the IRIS plant and its reliability is key to the core damage frequency evaluation, it received more extensive fault-tree development. Finally the dominant sequences that lead to severe accidents and the failures in the main and support systems are identified. (authors)

Mizuno, Yuko O.; Ogura, Katsunori; Ninokata, Hisashi [Tokyo Institute of Technology, 2-12-1 O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Conway, Lawrence E. [Westinghouse Electric Company (United States)

2002-07-01

322

Micropropagation of Iris sp.  

PubMed

Irises are perennial plants widely used as ornamental garden plants or cut flowers. Some species accumulate secondary metabolites, making them highly valuable to the pharmaceutical and perfume industries. Micropropagation of irises has successfully been accomplished by culturing zygotic embryos, different flower parts, and leaf base tissues as starting explants. Plantlets are regenerated via somatic embryogenesis, organogenesis, or both processes at the same time depending on media composition and plant species. A large number of uniform plants are produced by somatic embryogenesis, however, some species have decreased morphogenetic potential overtime. Shoot cultures obtained by organogenesis can be multiplied for many years. Somatic embryogenic tissue can be reestablished from leaf bases of in vitro-grown shoots. The highest number of plants can be obtained by cell suspension cultures. This chapter describes effective in vitro plant regeneration protocols for Iris species from different types of explants by somatic embryogenesis and/or organogenesis suitable for the mass propagation of ornamental and pharmaceutical irises. PMID:23179708

Jevremovi?, Sla?ana; Jekni?, Zoran; Suboti?, Angelina

2013-01-01

323

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

324

Aspects of iris image and iris match pair quality  

NASA Astrophysics Data System (ADS)

Iris recognition technology has the potential to broaden to include non-ideal imaging situations, as well as scale to national-level deployments. Hence, the study of population factors, subject intrinsics, and sensing contexts that can individually or collectively impact iris recognition performance assumes increased importance. This presentation summarizes recent research on a number of such "quality variables" and motivates the need for large data sets exhibiting a variety of such non-idealities to adequately characterize performance.

Flynn, Patrick J.; Bowyer, Kevin W.

2010-04-01

325

Combined analysis of neutron and photon flux measurements for the Jules Horowitz reactor core mapping  

SciTech Connect

We study the combined analysis of nuclear measurements to improve the knowledge of the irradiation conditions in the experimental locations of the future Jules Horowitz Reactor (JHR). The goal of the present work is to measure more accurately neutron flux, photon flux and nuclear heating in the reactor. In a Material Testing Reactor (MTR), nuclear heating is a crucial parameter to design the experimental devices to be irradiated in harsh nuclear conditions. This parameter drives the temperature of the devices and of the samples. The numerical codes can predict this parameter but in-situ measurements are necessary to reach the expected accuracy. For this reason, one objective of the IN-CORE program [1] is to study the combined measurements of neutron and photon flux and their cross advanced interpretation. It should be reminded that both neutron and photon sensors are not totally selective as their signals are due to neutron and photon interactions. We intend to measure the neutron flux by three different kinds of sensors (Uranium Fission chamber, Plutonium Fission chamber and Self Powered Neutron Detector), the photon flux by two different sensors (Ionization chamber and Self Powered Gamma Detector) and the nuclear heating by two different ones (Differential calorimeter and Gamma Thermometer). For the same parameter, we expect that the use of different kinds of sensors will allow a better estimation of the aimed parameter by mixing different spectrum responses and different neutron and gamma contributions. An experimental test called CARMEN-1 is scheduled in OSIRIS reactor (CEA Saclay - France) at the end of 2011, with the goal to map irradiation locations in the reactor reflector to get a first validation of the analysis model. This article focuses on the sensor selection for CARMEN-1 experiment and to the way to link neutron and photon flux measurements in view to reduce their uncertainties but also to better assess the neutron and photon contributions to nuclear heating. (authors)

Fourmentel, D.; Villard, J. F.; Lyoussi, A. [DEN Reactor Studies Dept., French Nuclear Energy and Alternative Energies Commission, CEA Cadarache, 13108 Saint Paul-Lez-Durance (France); Reynard-Carette, C. [Laboratoire Chimie Provence LCP UMR 6264, Univ. of Provence, Centre St. Jerome, 13397 Marseille Cedex 20 (France); Bignan, G.; Chauvin, J. P.; Gonnier, C.; Guimbal, P.; Malo, J. Y. [DEN Reactor Studies Dept., French Nuclear Energy and Alternative Energies Commission, CEA Cadarache, 13108 Saint Paul-Lez-Durance (France); Carette, M.; Janulyte, A.; Merroun, O.; Brun, J.; Zerega, Y.; Andre, J. [Laboratoire Chimie Provence LCP UMR 6264, Univ. of Provence, Centre St. Jerome, 13397 Marseille Cedex 20 (France)

2011-07-01

326

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

SciTech Connect

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

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

2012-07-01

327

Calculational-experimental research models for a fast reactor with a heterogeneous core  

SciTech Connect

The physical characteristics of heterogeneous metallic oxide cores were experimentally studied by physical tests of the critical assemblies BFS-46 and BFS-46AZ, which simulate a reactor of the BN-1600 type, into the core of which a fuel assembly with metallic uranium is inserted. A calculational model for the critical assemblies being investigated, showing the zones and their dimensions, is presented. The critical assembly BFS-46AZ is a modification of the basic critical assembly BFS-46 which adds plutonium to the IBZ to simulate its accumulation during reactor operation. The BFS-46 and BFS-46AZ assemblies have identical dimensions for the IBZ and LEZ, and have different HEZ dimensions, necessary to ensure the criticality of each assembly. Plutonium with a /sup 240/Pu content equal to 3.8% is used in the LEZ. The critically parameters are calculated using one-dimensional and two-dimensional models in a 26-group diffusion approximation based on the BNAP-78 system of group constants.

Belov, S.P.; Bobrov, S.B.; Kazanskii, Yu.A.; Kuzin, E.N.; Matveev, V.I.; Novozhilov, A.I.; Chernyi, V.A.

1987-11-01

328

Startup and control of out-of-core thermionic space reactors  

NASA Astrophysics Data System (ADS)

An analysis of out-of-core thermionic space reactor (OTR) startup and control has been performed. The reference ionic thermal reactor (ITR) chosen for this study is a 75 kWt version of the STAR-C (GA Technologies 1987). The applicability of point kinetics was first verified for the reference OTR system. Point kinetics applicability was verified for core length-to-diameter (L/D) ratios of two and four, and for both subcritical-to-critical and critical-to-supercritical transients. A coupled thermal/point kinetics code was then written, and OTR startup was analyzed. The analyses lead to several observations. First, point kinetics is applicable to the reference OTR for all transients considered. Second, to achieve a 900 second startup the reference OTR must operate at powers well above steady-state rated power during startup. Finally, the large thermal inertia of the radial reflector could be used to reduce radiator temperature during the first several hundred seconds of operation. Further research should be performed on transient heat pipe operation and on off-normal thermionic converter operation.

Houts, M. G.; Lanning, D. D.

329

Investigation of Core Thermohydraulics in Fast Reactors - Interwrapper Flow During Natural Circulation  

SciTech Connect

A proper assessment of core thermohydraulics under natural circulation conditions is important so that the full potential of the inherent, passive feature of a fast reactor can be used. When the heat exchangers of the decay heat removal system are operated in the upper plenum of a reactor vessel, cold sodium exiting the heat exchangers may penetrate into the gap regions between fuel subassemblies; this gap flow between the wrapper tubes of subassemblies is called interwrapper flow (IWF). During natural circulation decay heat removal, IWF will significantly modify the flow and temperature distributions in the subassemblies. Sodium experiments were carried out to investigate these phenomena, using a test section consisting of seven subassemblies housed and connected to an upper plenum. The cooling effect of IWF on the fuel subassemblies was evaluated and a new nondimensional parameter was deduced to characterize this effect. On the other hand, IWF reduced the natural circulation flow in the primary loop due to a temperature decrease in the upper part of the core. A balance between the cooling effect and the flow reduction effect is discussed. Three-dimensional analyses were performed to establish an estimation method for IWF. For the temperature decreases due to IWF at the hottest point in the subassemblies there was good agreement between experiments and predictions.

Kamide, H.; Hayashi, K.; Isozaki, T.; Nishimura, M. [Japan Nuclear Cycle Development Institute (Japan)

2001-01-15

330

Iris Matching Based on Personalized Weight Map.  

PubMed

Iris recognition typically involves three steps, namely iris image preprocessing, feature extraction and feature matching. The first two steps of iris recognition have been well studied but the last step is less addressed. Each human iris has its unique visual pattern and local image features also vary from region to region, which leads to significant differences in robustness and distinctiveness among the feature codes derived from different iris regions. However, most state-of-the-art iris recognition methods use a uniform matching strategy, where features extracted from different regions of the same person or the same region for different individuals are considered to be equally important. This paper proposes a personalized iris matching strategy using a class-specific weight map learned from the training images of the same iris class and updated online. The weight map reflects the robustness of an encoding algorithm on different iris regions by assigning an appropriate weight to each feature code for iris matching. Such a weight map trained by sufficient iris templates is convergent and robust against various noise. Extensive and comprehensive experiments demonstrate that the proposed personalized iris matching strategy achieves much better iris recognition performance than uniform strategies, especially for poor-quality iris images. PMID:21173439

Dong, Wenbo; Tan, Tieniu; Sun, Zhenan

2010-12-14

331

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.

332

Modification of the Penn State Reactor to allow transverse and rotational core motion to increase operational versatility  

SciTech Connect

At Penn State the Nuclear Engineering students have the opportunity to perform experiments in reactor physics, work with reactor and radiation instrumentation, and operate a nuclear reactor. These activities are done at the Penn State Breazeale Reactor (PSBR), a General Atomics Mark III TRIGA reactor. Unfortunately this activity alone can not fully support the facility. The PSBR is mandated by Penn State to provide a portion of its operating budget by selling services to users outside as well as inside Penn State. In order to increase the marketability of PSBR an upgrade program was started to increase the quality and versatility of operation. The PSBR is the longest operating university reactor in the United States. The first phase of the upgrade program began in 1992. The quality of operation was increased by replacing a 1965 vintage console with a more reliable digital control and monitoring system. The present phase of the upgrade program is to increase the versatility of operation by modifying the reactor to allow transverse and rotational core motion. Adding two more degrees of motion to the reactor core increases the capability of the facility to meet the needs of present and future users. This upgrade is being financed by a grant from the Department of Energy and matching funds from Penn State. (author)

Hughes, Daniel E. [Penn State University (United States)

1994-07-01

333

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

Microsoft Academic Search

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

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

2010-01-01

334

Core performance and proliferation resistance prospective of a novel natural uranium fueled, heavy water moderated nuclear research reactor  

Microsoft Academic Search

Three-dimensional burnup calculations were carried out to analyze the performance and proliferation resistance prospective of a novel natural uranium fueled, D2O moderated, D2O cooled and graphite reflected nuclear research reactor. The lattice simulation code WIMS-D\\/4 generated microscopic group constants, (?a, ?f, ?tr, ? etc.,), in conjunction with the diffusion theory based reactor core simulation code CITATION was employed in this

Mohammad Javed Khan; Aslam; Nasir Ahmad

2006-01-01

335

Characterization and quantification of an in-core neutron irradiation facility at a TRIGA II research reactor  

NASA Astrophysics Data System (ADS)

Experiments have been performed to characterize the neutron environment at an in-core TRIGA type nuclear research reactor. Steady-state thermal and epithermal neutron environment testing is important for many applications including, materials, electronics and biological cells. A well characterized neutron environment at a research reactor, including energy spectrum and spatial distribution, can be useful to many research communities and for educational research. This paper describes the characterization process and an application of exposing electronics to high neutron fluence.

Aghara, Sukesh; Charlton, William

2006-07-01

336

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY2011 Activities  

Microsoft Academic Search

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

Michael A. Pope

2011-01-01

337

A detailed neutronics comparison of the university of Florida training reactor (UFTR) current HEU and proposed LEU cores  

SciTech Connect

For over 35 years, the UFTR highly-enriched core has been safely operated. As part of the Reduced Enrichment for Research and Test Reactors Program, the core is currently being converted to low-enriched uranium fuel. The analyses presented in this paper were performed to verify that, from a neutronic perspective, a proposed low-enriched core can be operated as safely and as effectively as the highly-enriched core. Detailed Monte Carlo criticality calculations are performed to determine: i) Excess reactivity for different core configurations, ii) Individual integral blade worth and shutdown margin, iii) Reactivity coefficients and kinetic parameters, and iv) Flux profiles and core six-factor formula parameters. (authors)

Dionne, B.; Haghighat, A.; Yi, C.; Smith, R.; Ghita, G.; Manalo, K.; Sjoden, G.; Huh, J.; Baciak, J.; Mock, T.; Wenner, M. [Dept. of Nuclear and Radiological Engineering, Univ. of Florida, Gainesville, FL (United States); Matos, J.; Stillman, J. [Reduced Enrichment for Research and Test Reactors Program, Argonne National Laboratory, Argonne, IL (United States)

2006-07-01

338

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

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

David W. Nigg; Devin A. Steuhm

2011-09-01

339

Moxifloxacin and bilateral acute iris transillumination  

PubMed Central

Recent publications have alerted clinicians to a syndrome of uveitic transilluminating iris depigmentation associated with systemic fluoroquinolones and other antibiotics. Bilateral acute iris transillumination, which is associated with loss of the iris pigment epithelium and results in iris transillumination, differs from the previously described bilateral acute depigmentation of the iris, which is associated with atrophy of the iris stroma without transillumination. We present a case of fluoroquinolone-associated uveitis with anterior segment optical coherence tomography imaging to highlight some observations about this syndrome. We interpret pharmacokinetic data to help explain why oral, but not topical, moxifloxacin may cause fluoroquinolone-associated uveitis.

2013-01-01

340

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

SciTech Connect

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

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

1987-01-01

341

Development of an asymmetric multiple position neutron source (AMPNS) method for monitoring the criticality of the degraded reactor core  

Microsoft Academic Search

An analytical\\/experiment method was developed to monitor the subcritical reactivity and unfold the k\\/sub infinity\\/ distribution of a degraded reactor core. The method uses several fixed neutron detectors and a Cf-252 neutron source placed sequentially in multiple positions in the core. Therefore, it is called the asymmetric multiple position neutron source (AMPNS) method. The AMPNS method employs the nucleonic codes

1984-01-01

342

Investigations of sloshing fluid motions in pools related to recriticalities in liquid-metal fast breeder reactor core meltdown accidents  

Microsoft Academic Search

This paper reports that analyses of unprotected loss-of-flow accidents for medium-size cores of current liquid-metal fast breeder reactors have shown that the accident proceeds into a transition phase where further meltdown is accompanied by recriticalities and secondary excursions. Assuming very pessimistic conditions concerning fuel discharge and blockage formation, a neutronically active whole-core pool of molten m material can form. Neutronic

W. Maschek; C. D. Munz; L. Meyer

1992-01-01

343

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

344

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

345

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

SciTech Connect

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

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

1990-01-01

346

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

347

ACRR (Annular Core Research Reactor) fission product release tests: ST-1 and ST-2  

SciTech Connect

Two experiments (ST-1 and ST-2) have been performed in the Annular Core Research Reactor (ACER) at Sandia National Laboratories (SNLA) to obtain time-resolved data on the release of fission products from irradiated fuels under light water reactor (LWR) severe accident conditions. Both experiments were conducted in a highly reducing environment at maximum fuel temperatures of greater than 2400 K. These experiments were designed specifically to investigate the effect of increased total pressure on fission product release; ST-1 was performed at approximately 0.16 MPa and ST-2 was run at 1.9 MPa, whereas other parameters were matched as closely as possible. Release rate data were measured for Cs, I, Ba, Sr, Eu, Te, and U. The release rates were higher than predicted by existing codes for Ba, Sr, Eu, and U. Te release was very low, but Te did not appear to be sequestered by the zircaloy cladding; it was evenly distributed in the fuel. In addition, in posttest analysis a unique fuel morphology (fuel swelling) was observed which may have enhanced fission product release, especially in the high pressure test (ST-2). These data are compared with analytical results from the CORSOR correlation and the VICTORIA computer model. 8 refs., 8 figs., 2 tabs.

Allen, M.D.; Stockman, H.W.; Reil, K.O.; Grimley, A.J.; Camp, W.J.

1988-01-01

348

Evaluation of radcal gamma thermometers for in-core monitoring of Savannah River Site production reactors  

SciTech Connect

The Savannah River Site (SRS) recently obtained a quantity of Radcal Gamma Thermometer Assemblies (RGTAs) for in-core monitoring of local power in their production reactors. The RGTAs, manufactured by DELTA M Corporation in Oak Ridge, Tennessee, contained seven Self Calibrating Gamma Thermometer (SCGT) sensors within a 7.26 mm diameter, 3.06 m length with a total length of 5.6 m. All RGTAs contained an isolated segmented heater cable for in-situ calibration. Each SCGT sensor was subjected to a 40 point calibration at discrete power levels from 0.5 to 6 watts per gram (w/g) under both joule and cable power. Calibration equations were developed from this to predict reactor power at each sensor. Additionally three units were calibrated at combined joule and cable heating conditions from 0.5 to 2.5 w/g cable and 0.5 to 6 w/g joule. A statistical analysis of all data was used to derive prediction equations that enable SRS engineers to precisely track any changes in sensor calibration throughout the lifetime of the instruments. This paper presents the detailed configuration of the 36 units manufactured for SRS, reviews the calibration results, and discusses the utility and accuracy of the statistically derived prediction equations for in-situ calibration.

McCulloch, R.W.; Crowley, J.L. [DELTA M Corp., Oak Ridge, TN (United States); Croft, W.D. [Westinghouse Savannah River Co., Aiken, SC (United States)

1991-12-31

349

Evaluation of radcal gamma thermometers for in-core monitoring of Savannah River Site production reactors  

SciTech Connect

The Savannah River Site (SRS) recently obtained a quantity of Radcal Gamma Thermometer Assemblies (RGTAs) for in-core monitoring of local power in their production reactors. The RGTAs, manufactured by DELTA M Corporation in Oak Ridge, Tennessee, contained seven Self Calibrating Gamma Thermometer (SCGT) sensors within a 7.26 mm diameter, 3.06 m length with a total length of 5.6 m. All RGTAs contained an isolated segmented heater cable for in-situ calibration. Each SCGT sensor was subjected to a 40 point calibration at discrete power levels from 0.5 to 6 watts per gram (w/g) under both joule and cable power. Calibration equations were developed from this to predict reactor power at each sensor. Additionally three units were calibrated at combined joule and cable heating conditions from 0.5 to 2.5 w/g cable and 0.5 to 6 w/g joule. A statistical analysis of all data was used to derive prediction equations that enable SRS engineers to precisely track any changes in sensor calibration throughout the lifetime of the instruments. This paper presents the detailed configuration of the 36 units manufactured for SRS, reviews the calibration results, and discusses the utility and accuracy of the statistically derived prediction equations for in-situ calibration.

McCulloch, R.W.; Crowley, J.L. (DELTA M Corp., Oak Ridge, TN (United States)); Croft, W.D. (Westinghouse Savannah River Co., Aiken, SC (United States))

1991-01-01

350

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

351

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

SciTech Connect

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

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

2006-07-31

352

Improvement on the prediction accuracy of transmutation properties for fast reactor core using the minor actinides irradiation test data on the Joyo MK-II CORE  

Microsoft Academic Search

For a validation of MA nuclear data and improvement on the prediction accuracy of MA transmutation properties in fast reactor cores, MA sample irradiation test data of Joyo were utilized. Adopting MA cross-sections in JENDL-3.3, result of their evaluations showed good agreement with experimental data. Further, the present study clarified that utilization of these data with cross-section adjustment technique has

Sugino; Kazuteru

2007-01-01

353

Analytical evaluation on dynamical response characteristics of reduced-moderation water reactor with tight-lattice core under natural circulation core cooling  

Microsoft Academic Search

The time-domain analyses with TRAC-BF1 code were performed for clarifying the dynamical response characteristics of the reduced-moderation water reactor (RMWR) with tight-lattice core configuration. The response characteristics were evaluated based on the step response basically utilized for dynamical system evaluation. As for the most fundamental dynamical characteristics, the channel flow response characteristics of single fuel assembly were evaluated. In the

Nobuyuki Ishikawa; Tsutomu Okubo

2009-01-01

354

Further Results On The Iris Effect  

NASA Astrophysics Data System (ADS)

Over the past year, our original work on the iris effect as an important negative feed- back for global climate has been subject to substantial criticism. We briefly review the iris effect in order to explain that it is, in fact, a property associated with cumulus convection, and must, therefore, be scaled by a measure of cumulus activity. What it depends on is the degree to which cumulus towers detrain ice, which, in turn, depends on precipitation efficiency within the cumulus themselves. Its observation depends on considering scales and times which allow for significant numbers of cumulus clouds under a variety of circumstances. Such conditions do not demand global measure- ments. In brief, the original results suggested that precipitation efficiency increased with temperature, and that, therefore, upper level cirrus coverage diminished with temperature. The negative feedback depended on the claim that the infra red impact of upper level cirrus exceeded its impact on the visible. The name, "iris," stems from the analogy between the opening and closing of clear areas in response to temperature to the opening and closing of the eye's iris in response to light. Our current efforts extend earlier efforts and respond to various criticisms: 1. Our data set now extends to 4 full years rather than the 20 months used in the original study. 2. We have considered versions of our full data with degraded spatial and temporal resolution in order to determine exactly what resolution is necessary for a study of this phenomenon. 3. We have devised a better measure of cumulonimbus activity than the simple infrared threshold used originally. 4. We have shown that uncertainties in infrared properties do not substantially alter feedback estimates. 5. We have used CERES data in order to investigate how albedo varies with area of cirrus (scaled by cumulus activity) in order to separate the albedo change due to fluctuations of cirrus areas from mean albedos which are biased by the high albedos associated with thick anvils near cumulus cores. It is, of course, the former which are relevant to the feedback. Results from each of the above studies will be presented.

Lindzen, R. S.; Chou, M.-D.; Hou, A. Y.

355

Startup and control of out-of-core thermionic space reactors  

NASA Astrophysics Data System (ADS)

An analysis of out-of-core thermionic space reactor (OTR) startup and control has been performed. The reference OTR chosen for this study is a 75 kWt version of the STAR-C (GA Technologies 1987). The applicability of point kinetics was first verified for the reference OTR system. Point kinetics applicability was verified for core length-to-diameter (L/D) ratios of two and four, and for both subcritical-to-critical and critical-to-supercritical transients. A coupled thermal/point kinetics code was then written, and OTR startup was analyzed. The analyses lead to several observations. First, point kinetics is applicable to the reference OTR for all transients considered. Second, to achieve a 900-s startup the reference OTR must operate at powers well above steady-state rated power during startup. Finally, the large thermal inertia of the radial reflector could be used to reduce radiator temperature during the first several hundred seconds of operation. Further research should be performed on transient heat pipe operation and on off-normal thermionic converter operation.

Houts, Michael G.; Lanning, Davic D.

1992-01-01

356

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

2010-10-14

357

Development of an inconel self powered neutron detector for in-core reactor monitoring  

NASA Astrophysics Data System (ADS)

The paper describes the development and testing of an Inconel600 (2 mm diameter×21 cm long) self-powered neutron detector for in-core neutron monitoring. The detector has 3.5 mm overall diameter and 22 cm length and is integrally coupled to a 12 m long mineral insulated cable. The performance of the detector was compared with cobalt and platinum detectors of similar dimensions. Gamma sensitivity measurements performed at the 60Co irradiation facility in 14 MR/h gamma field showed values of -4.4×10-18 A/R/h/cm (-9.3×10-24 A/?/cm2-s/cm), -5.2×10-18 A/R/h/cm (-1.133×10-23 A/?/cm2-s/cm) and 34×10-18 A/R/h/cm (7.14×10-23 A/?/cm2-s/cm) for the Inconel, Co and Pt detectors, respectively. The detectors together with a miniature gamma ion chamber and fission chamber were tested in the in-core Apsara Swimming Pool type reactor. The ion chambers were used to estimate the neutron and gamma fields. With an effective neutron cross-section of 4b, the Inconel detector has a total sensitivity of 6×10-23 A/nv/cm while the corresponding sensitivities for the platinum and cobalt detectors were 1.69×10-22 and 2.64×10-22 A/nv/cm. The linearity of the detector responses at power levels ranging from 100 to 200 kW was within ±5%. The response of the detectors to reactor scram showed that the prompt response of the Inconel detector was 0.95 while it was 0.7 and 0.95 for the platinum and cobalt self-powered detectors, respectively. The detector was also installed in the horizontal flux unit of 540 MW Pressurised Heavy Water Reactor (PHWR). The neutron flux at the detector location was calculated by Triveni code. The detector response was measured from 0.02% to 0.07% of full power and showed good correlation between power level and detector signals. Long-term tests and the dynamic response of the detector to shut down in PHWR are in progress.

Alex, M.; Ghodgaonkar, M. D.

2007-04-01

358

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

359

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

Microsoft Academic Search

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

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

2004-01-01

360

Radioactive Gaseous Effluents from the Core of the Agn-201 Reactor at the United States Naval Postgraduate School.  

National Technical Information Service (NTIS)

A qualtitative and quantitative analysis of the core gas generated by the AGN-201 reactor at the United States Naval Postgraduate School was made by analysis of the spectrum of gamma-rays emitted two hours after peak power operations. The principle radioa...

R. T. A. Bredderman

1966-01-01

361

Effect of control rods on seismic response of the core of a large high temperature gas cooled reactor: analytical study  

Microsoft Academic Search

An analytical study to assess the influence of control rods on the seismic response of the core of large High Temperature Gas Cooled Reactors (HTGR) is reported. The study is part of an extensive experimental and analytical program concerned with demonstrating the safety and functionability of the control rods and reserve shutdown systems during and after seismic events. This program

R. E. Bachman; R. P. Kennedy; S. A. Short

1975-01-01

362

Effect of Control Rods on Seismic Response of the Core of a Large High Temperature Gas Cooled Reactor: Analytical Study.  

National Technical Information Service (NTIS)

An analytical study to assess the influence of control rods on the seismic response of the core of large High Temperature Gas Cooled Reactors (HTGR) is reported. The study is part of an extensive experimental and analytical program concerned with demonstr...

R. E. Bachman R. P. Kennedy S. A. Short

1975-01-01

363

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

Microsoft Academic Search

To ensure that the public health and safety are protected under any accident conditions in a Liquid Metal Fast Breeder Reactor (LMFBR), many accidents are analyzed for their potential consequences. The SAS4A code system, described in this paper, provides such an analysis capability, including the ability to analyze low probability events such as the Hypothetical Core Disruptive Accidents (HCDAs). The

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

1987-01-01

364

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

365

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

366

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

SciTech Connect

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. [Bosscha Laboratory, Department of Physics, Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia)

2010-06-22

367

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

SciTech Connect

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

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

2008-05-05

368

A novel approach for iris recognition  

Microsoft Academic Search

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

U. M. Chaskar; M. S. Sutaone

2010-01-01

369

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

370

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

371

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

372

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

373

Effect of Image Compression on Iris Recognition  

Microsoft Academic Search

Iris recognition is a proven, accurate means to identify people. Commercial iris recognition systems are currently employed to allow passengers in some airports to be rapidly processed through security, to allow access to secure areas, and for secure access to computer networks. With the growing employment of iris recognition systems and associated research to support this, the need for large

Robert W. Ives; Bradford L. Bonney; Delores M. Etter

2005-01-01

374

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

SciTech Connect

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

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

2011-03-01

375

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

376

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

377

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

Federal Register 2010, 2011, 2012, 2013

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

2012-06-15

378

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

Federal Register 2010, 2011, 2012, 2013

...Pressurized-Water Reactors.'' This RG is being...new pressurized water reactor (PWR) designs. ADDRESSES...technical questions, contact the individual(s...FURTHER INFORMATION CONTACT: Frank Talbot, telephone...gov, Office of New Reactors, or Mark P....

2013-10-25

379

United States Department of Energy`s reactor core protection evaluation methodology for fires at RBMK and VVER nuclear power plants. Revision 1  

SciTech Connect

This document provides operators of Soviet-designed RBMK (graphite moderated light water boiling water reactor) and VVER (pressurized light water reactor) nuclear power plants with a systematic Methodology to qualitatively evaluate plant response to fires and to identify remedies to protect the reactor core from fire-initiated damage.

NONE

1997-06-01

380

A new approach for cancelable iris recognition  

NASA Astrophysics Data System (ADS)

The iris is a stable and reliable biometric for positive human identification. However, the traditional iris recognition scheme raises several privacy concerns. One's iris pattern is permanently bound with him and cannot be changed. Hence, once it is stolen, this biometric is lost forever as well as all the applications where this biometric is used. Thus, new methods are desirable to secure the original pattern and ensure its revocability and alternatives when compromised. In this paper, we propose a novel scheme which incorporates iris features, non-invertible transformation and data encryption to achieve "cancelability" and at the same time increases iris recognition accuracy.

Yang, Kai; Sui, Yan; Zhou, Zhi; Du, Yingzi; Zou, Xukai

2010-04-01

381

Extending Iris: The VAO SED Analysis Tool  

NASA Astrophysics Data System (ADS)

Iris is a tool developed by the Virtual Astronomical Observatory (VAO) for building and analyzing Spectral Energy Distributions (SEDs). Iris was designed to be extensible, so that new components and models can be developed by third parties and then included at runtime. Iris can be extended in different ways: new file readers allow users to integrate data in custom formats into Iris SEDs; new models can be fitted to the data, in the form of template libraries for template fitting, data tables, and arbitrary Python functions. The interoperability-centered design of Iris and the Virtual Observatory standards and protocols can enable new science functionalities involving SED data.

Laurino, O.; Busko, I.; Cresitello-Dittmar, M.; D'Abrusco, R.; Doe, S.; Evans, J.; Pevunova, O.

2013-10-01

382

CORTAP: a coupled neutron kinetics-heat transfer digital computer program for the dynamic simulation of the high temperature gas cooled reactor core  

Microsoft Academic Search

CORTAP (Core Transient Analysis Program) was developed to predict the dynamic behavior of the High Temperature Gas Cooled Reactor (HTGR) core under normal operational transients and postulated accident conditions. CORTAP is used both as a stand-alone component simulation and as part of the HTGR nuclear steam supply (NSS) system simulation code ORTAP. The core thermal neutronic response is determined by

1977-01-01

383

EXPOSURE SUMMARIES FOR IRIS CHEMICALS.  

EPA Science Inventory

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

384

IRIS: Intermolecular RNA Interaction Search  

Microsoft Academic Search

Here we present IRIS, a method for prediction of RNA-RNA interactions that is based on dy- namic programming and extends current RNA secondary structure prediction approaches. Using this method we have found a number of interesting reflnements to the structures of RNA-RNA complexes that have been studied previously and predicted novel targets for several known reg- ulatory RNAs in E.

Dmitri D. Pervouchine

2004-01-01

385

Dynamic Features for Iris Recognition.  

PubMed

The human eye is sensitive to visible light. Increasing illumination on the eye causes the pupil of the eye to contract, while decreasing illumination causes the pupil to dilate. Visible light causes specular reflections inside the iris ring. On the other hand, the human retina is less sensitive to near infra-red (NIR) radiation in the wavelength range from 800 nm to 1400 nm, but iris detail can still be imaged with NIR illumination. In order to measure the dynamic movement of the human pupil and iris while keeping the light-induced reflexes from affecting the quality of the digitalized image, this paper describes a device based on the consensual reflex. This biological phenomenon contracts and dilates the two pupils synchronously when illuminating one of the eyes by visible light. In this paper, we propose to capture images of the pupil of one eye using NIR illumination while illuminating the other eye using a visible-light pulse. This new approach extracts iris features called "dynamic features (DFs)." This innovative methodology proposes the extraction of information about the way the human eye reacts to light, and to use such information for biometric recognition purposes. The results demonstrate that these features are discriminating features, and, even using the Euclidean distance measure, an average accuracy of recognition of 99.1% was obtained. The proposed methodology has the potential to be "fraud-proof," because these DFs can only be extracted from living irises. PMID:22389153

da Costa, Ronaldo Martins; Gonzaga, Adilson

2012-02-29

386

An analysis of IrisCode.  

PubMed

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 methods have been developed for iris and palmprint identification. However, we lack a detailed analysis of IrisCode. The aim of this paper is to provide such an analysis as a way of better understanding IrisCode, extending the coarse phase representation to a precise phase representation, and uncovering the relationship between IrisCode and other coding methods. Our analysis demonstrates that IrisCode is a clustering algorithm with four prototypes; the locus of a Gabor function is a 2-D ellipse with respect to a phase parameter and can be approximated by a circle in many cases; Gabor function can be considered as a phase-steerable filter and the bitwise hamming distance can be regarded as a bitwise phase distance. We also discuss the theoretical foundation of the impostor binomial distribution. We use this analysis to develop a precise phase representation which can enhance accuracy. Finally, we relate IrisCode and other coding methods. PMID:20083454

Kong, Adams W K; Zhang, David; Kamel, Mohamed S

2010-02-01

387

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

388

Recent research results in iris biometrics  

NASA Astrophysics Data System (ADS)

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

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

2009-05-01

389

Review of the state of criticality of the Three Mile Island Unit 2 core and reactor vessel  

SciTech Connect

The events during the early hours of the Three Mile Island Unit 2 (TMI-2) accident on March 28, 1979 caused the fuel in the reactor core to crumble or disintegrate, and then subside into a rubble structure more compact that its normal configuration. The present height of the core is about seven feet, five feet less than its normal configuration of 12 feet. With the same boron content and some or all of the control rod and burnable poison rod material as the normal core configuration, the collapsed structure is calculated to be more reactive. However, the reactor is assuredly subcritical at present because of the extraordinarily high boron concentration maintained in the coolant water. Four additional and different physical models are discussed briefly in the report to illustrate the margin of subcriticality, to provide a better estimate of the neutron multiplication factor, and to provide some understanding of the criticality effects of the important parameters. Two different finite, cylindrical models of a collapsed core are also presented in this report. The conclusion of this review is that the reactor is now very far subcritical with a boron concentration of 4350 ppM or more, and no conceivable rearrangement of fuel can create a critical state. Careful administrative control to maintain the boron concentration of the reactor coolant close to 5000 ppM, and controls to rigorously exclude addition of unborated water to the primary system, provide additional assurance that subcriticality will be maintained. The immediate corollary is that the defueling of the reactor vessel can proceed as planned, with complete confidence that such operations will remain subcritical. 20 refs.

Stratton, W.R. (GPU Nuclear Corp., Middletown, PA (USA))

1987-04-15

390

Natural nuclear reactor at Oklo and variation of fundamental constants: Computation of neutronics of a fresh core  

SciTech Connect

Using modern methods of reactor physics, we performed full-scale calculations of the Oklo natural reactor. For reliability, we used recent versions of two Monte Carlo codes: the Russian code MCU-REA and the well-known international code MCNP. Both codes produced similar results. We constructed a computer model of the Oklo reactor zone RZ2 which takes into account all details of design and composition. The calculations were performed for three fresh cores with different uranium contents. Multiplication factors, reactivities, and neutron fluxes were calculated. We also estimated the temperature and void effects for the fresh core. As would be expected, we found for the fresh core a significant difference between reactor and Maxwell spectra, which had been used before for averaging cross sections in the Oklo reactor. The averaged cross section of {sub 62}{sup 149}Sm and its dependence on the shift of a resonance position E{sub r} (due to variation of fundamental constants) are significantly different from previous results. Contrary to the results of previous papers, we found no evidence of a change of the samarium cross section: a possible shift of the resonance energy is given by the limits -73{<=}{delta}E{sub r}{<=}62 meV. Following tradition, we have used formulas of Damour and Dyson to estimate the rate of change of the fine structure constant {alpha}. We obtain new, more accurate limits of -4x10{sup -17}{<=}{alpha}{center_dot}/{alpha}{<=}3x10{sup -17} yr{sup -1}. Further improvement of the accuracy of the limits can be achieved by taking account of the core burn-up. These calculations are in progress.

Petrov, Yu. V.; Nazarov, A. I.; Onegin, M. S.; Petrov, V. Yu.; Sakhnovsky, E. G. [St. Petersburg Nuclear Physics Institute, Gatchina, RU-188-300, St. Petersburg (Russian Federation)

2006-12-15

391

Thermohydraulics in a high-temperature gas-cooled reactor prestressed-concrete reactor vessel during unrestricted core-heatup accidents  

SciTech Connect

The hypothetical accident considered for siting considerations in High Temperature Gas-Cooled Reactors (HTGR) is the so called Unrestricted Core Heatup Accident (UCHA), in which all forced circulation is lost at initiation, and none of the auxillary cooling loops can be started. The result is a gradual slow core heatup, extending over days. Whether the liner cooling system (LCS) operates during this time is of crucial importance. If it does not, the resulting concrete decomposition of the prestressed concrete reactor vessel (PCRV) will ultimately cause containment building (CB) failure after about 6 to 10 days. The primary objective of the work described here was to establish for such accident conditions the core temperatures and approximate fuel failure rates, to check for potential thermal barrier failures, and to follow the PCRV concrete temperatures, as well as PCRV gas releases from concrete decomposition. The work was done for the General Atomic Corporation Base Line Zero reactor of 2240 MW(t). Most results apply at least qualitatively also to other large HTGR steam cycle designs.

Kroeger, P.G.; Colman, J.; Araj, K.

1983-01-01

392

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

Microsoft Academic Search

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

B. Boer; A. M. Ougouag

2010-01-01

393

Removal plan for Shippingport pressurized water reactor core 2 blanket fuel assemblies form T plant to the canister storage building  

SciTech Connect

This document presents the current strategy and path forward for removal of the Shippingport Pressurized Water Reactor Core 2 blanket fuel assemblies from their existing storage configuration (wet storage within the T Plant canyon) and transport to the Canister Storage Building (designed and managed by the Spent Nuclear Fuel. Division). The removal plan identifies all processes, equipment, facility interfaces, and documentation (safety, permitting, procedures, etc.) required to facilitate the PWR Core 2 assembly removal (from T Plant), transport (to the Canister storage Building), and storage to the Canister Storage Building. The plan also provides schedules, associated milestones, and cost estimates for all handling activities.

Lata

1996-09-26

394

Assessing the IRIS Professional Development Model: Impact Beyond the Workshops  

NASA Astrophysics Data System (ADS)

The IRIS Education and Outreach (E&O) Program has developed a highly effective, one-day professional development experience for formal educators. Leveraging the expertise of its consortium, IRIS delivers content including: plate tectonics, propagation of seismic waves, seismographs, Earth's interior structure. At the core of the IRIS professional development model is the philosophy that changes in teacher behavior can be affected by increasing teacher comfort in the classroom. Science and research organizations such as IRIS are able to increase teachers' comfort in the classroom by providing professional development which: increases an educator's knowledge of scientific content, provides educators with a variety of high-quality, scientifically accurate activities to deliver content to students, and provides educators with experiences involving both the content and the educational activities as the primary means of knowledge transfer. As reflected in a 2002-2003 academic year assessment program, this model has proven to be effective at reaching beyond participants and extending into the educators' classrooms. 76% of respondents report increasing the amount of time they spend teaching seismology or related topics in their classroom as a result of participating in IRIS professional development experience. This increase can be directly attributed to the workshop as 90% of participants report using at least one activity modeled during the workshop upon returning to their classrooms. The reported mean activity usage by teachers upon was 4.5 activities per teacher. Since the inception of the professional development model in 1999, IRIS E&O has been committed to evaluation. Data derived from assessment is utilized as a key decision making tool, driving a continuous improvement process. As a result, both the model and the assessment methods have become increasingly refined and sophisticated. The alignment of the professional development model within the IRIS E&O Program framework has resulted in a clarified a definition of success and an increased demand for the collection of new data. Currently, the assessment program is testing tools to examine participant learning, measure the transfer of knowledge and resources from professional development into in classrooms, and measure the use of individual activities.

Hubenthal, M.; Braile, L. W.; Taber, J. J.

2003-12-01

395

Self-adaptive iris image acquisition system  

NASA Astrophysics Data System (ADS)

Iris image acquisition is the fundamental step of the iris recognition, but capturing high-resolution iris images in real-time is very difficult. The most common systems have small capture volume and demand users to fully cooperate with machines, which has become the bottleneck of iris recognition's application. In this paper, we aim at building an active iris image acquiring system which is self-adaptive to users. Two low resolution cameras are co-located in a pan-tilt-unit (PTU), for face and iris image acquisition respectively. Once the face camera detects face region in real-time video, the system controls the PTU to move towards the eye region and automatically zooms, until the iris camera captures an clear iris image for recognition. Compared with other similar works, our contribution is that we use low-resolution cameras, which can transmit image data much faster and are much cheaper than the high-resolution cameras. In the system, we use Haar-like cascaded feature to detect faces and eyes, linear transformation to predict the iris camera's position, and simple heuristic PTU control method to track eyes. A prototype device has been established, and experiments show that our system can automatically capture high-quality iris image in the range of 0.6m×0.4m×0.4m in average 3 to 5 seconds.

Dong, Wenbo; Sun, Zhenan; Tan, Tieniu; Qiu, Xianchao

2008-03-01

396

An effective iris recognition system for identification of humans  

Microsoft Academic Search

In this paper, identification and verification approach based on human iris pattern is presented. The system is based on several steps from capturing the iris pattern, determining and localizing the iris boundaries, transformation of localized iris into rectangular and polar components, extracting the features from the image, based on wavelet transformation and at the end the matching of the iris.

Muhaminad Khurrarn Khan; Jiashu Zhang; Shi-Jinn Horng

2004-01-01

397

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

398

Core conversion analyses of the Syrian MNSR reactor from HEU to LEU and MEU fuel with homogeneously mixed burnable poisons  

Microsoft Academic Search

A comprehensive analysis has been performed to investigate the conversion of the Syrian MNSR (miniature neutron source reactor) from current HEU fuel to selected alternatives LEU and MEU fuels. For this purposes the core design calculations related to design and engineering of LEU and MEU fuels have been carried out using the codes WIMSD\\/4 and BORGES-part of the MTR-PC and

N. Ghazi; H. Haj Hassan; A. Hainoun

2009-01-01

399

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

Microsoft Academic Search

An experimental demonstration of fiber optic temperature sensing in the in-core region of Japan Materials Testing Reactor from 250 to 750 degrees C is described. Temperature data could be obtained for two full-power weeks with neutron fluxes of approximately 1014 n\\/cm2\\/s and gamma dose rates of approximately 5 X 103 Gy\\/s. The measurements were based on thermally generated IR light

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

1997-01-01

400

Radiation-induced material changes and susceptibility to intergranular failure of light-water-reactor core internals  

Microsoft Academic Search

Current understanding of radiation-induced material changes that occur in light-water-reactor (LWR) core components is critically reviewed and linked to intergranular failure processes. Although the basic science of radiation damage processes in metals is reasonably well established, accurate prediction of microstructures, microchemistries and mechanical property changes in complex stainless alloys during irradiation at LWR temperatures is not possible at present. Mechanistic

Stephen M. Bruemmer; Edward P. Simonen; Peter M. Scott; Peter L. Andresen; James L. Nelson

1999-01-01

401

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

Microsoft Academic Search

Reliability-centered maintenance (RCM) methods are applied to boiling water reactor plant-specific emergency core cooling system probabilistic risk assessment (PRA) fault trees. The RCM is a technique that is system function-based, for improving a preventive maintenance (PM) program, which is applied on a component basis. Many PM programs are based on time-directed maintenance tasks, while RCM methods focus on component condition-directed

Y. A. Choi; M. A. Feltus

1995-01-01

402

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

403

Influence of high dose irradiation on core structural and fuel materials in advanced reactors.  

National Technical Information Service (NTIS)

The IAEA International Working Group on Fast Reactors (IWGFR) periodically organizes meeting to discuss and review important aspects of fast reactor technology. The fifth meeting held in Obninsk, Russian Federation, 16-19 June 1997, was devoted to the inf...

1998-01-01

404

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

405

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

406

Thermal hydraulic response of the Advanced Neutron Source Reactor to piping breaks near the core region  

Microsoft Academic Search

This paper describes the application of the RELAP5 thermal hydraulic code to a highly subcooled, plate type reactor typical of many research and production reactor systems. The specific system modeled is the latest design of the Advanced Neutron Source Reactor (ANSR). A discussion of the model as well as the results from several loss-of-coolant accident (LOCA) scenarios is included. The

N. C. J. Chen; P. T. Williams; G. L. Yoder

1992-01-01

407

Steady-state thermal hydraulic analysis of the equilibrium core of Pakistan research reactor-1  

Microsoft Academic Search

Steady-state thermal hydraulic analysis of Pakistan Research Reactor-1 (PARR-1) has been carried out. RELAP5\\/Mod 3.4 (a best-estimate system code) was employed. PARR-1 is a swimming pool type research reactor using MTR (Material Testing Reactor) type fuel. It uses low enriched uranium (

I. H. Bokhari; T. Mahmood; K. S. Chaudri

2007-01-01

408

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.

Dharmasena, Aruna; Wallis, Simon

2013-01-01

409

IRIS: Animations of Plate Tectonics  

NSDL National Science Digital Library

This is a collection of animations on dynamic earth processes: plate tectonics, earthquakes, volcanoes, and seismic waves. Users can explore the interaction of Earth's tectonic plates, view models of P and S wave propagation, study how seismographs work, monitor earthquakes and volcanoes, and get instructions for modeling earthquakes in the classroom. This resource is part of IRIS, the Incorporated Research Institutions for Seismology, a consortium of international laboratories and data collection centers.

2011-03-18

410

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

SciTech Connect

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

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

2012-07-01

411

Video-based noncooperative iris image segmentation.  

PubMed

In this paper, we propose a video-based noncooperative iris image segmentation scheme that incorporates a quality filter to quickly eliminate images without an eye, employs a coarse-to-fine segmentation scheme to improve the overall efficiency, uses a direct least squares fitting of ellipses method to model the deformed pupil and limbic boundaries, and develops a window gradient-based method to remove noise in the iris region. A remote iris acquisition system is set up to collect noncooperative iris video images. An objective method is used to quantitatively evaluate the accuracy of the segmentation results. The experimental results demonstrate the effectiveness of this method. The proposed method would make noncooperative iris recognition or iris surveillance possible. PMID:20403786

Du, Yingzi; Arslanturk, Emrah; Zhou, Zhi; Belcher, Craig

2010-04-15

412

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

413

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

NASA Astrophysics Data System (ADS)

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

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

1997-11-01

414

DCT-based iris recognition.  

PubMed

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 the Bath database. On this data, we achieve 100 percent Correct Recognition Rate (CRR) and perfect Receiver-Operating Characteristic (ROC) Curves with no registered false accepts or rejects. Individual feature bit and patch position parameters are optimized for matching through a product-of-sum approach to Hamming distance calculation. For verification, a variable threshold is applied to the distance metric and the False Acceptance Rate (FAR) and False Rejection Rate (FRR) are recorded. A new worst-case metric is proposed for predicting practical system performance in the absence of matching failures, and the worst case theoretical Equal Error Rate (EER) is predicted to be as low as 2.59 x 10(-4) on the available data sets. PMID:17299216

Monro, Donald M; Rakshit, Soumyadip; Zhang, Dexin

2007-04-01

415

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

416

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

417

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

418

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

419

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

420

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.  

National Technical Information Service (NTIS)

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 l...

T. A. Parish

1995-01-01

421

Performance Specification Fuel Drying and Canister Inerting System for Shippingport Pressurized Water Reactor (PWR) Core 2 Blanket Fuel Assemblies Stored Within Shippingport Spent Fuel Canisters.  

National Technical Information Service (NTIS)

This specification establishes the performance requirements and basic design requirements imposed on the fuel drying and canister inerting system for Shippingport Pressurized Water Reactor (PWR) Core 2 blanket fuel assemblies (BFAs) stored within Shipping...

D. M. Johnson

2000-01-01

422

Modeling and analysis framework for core damage propagation during flow-blockage-initiated accidents in the Advanced Neutron Source reactor at Oak Ridge National Laboratory.  

National Technical Information Service (NTIS)

This paper describes modeling and analysis to evaluate the extent of core damage during flow blockage events in the Advanced Neutron Source (ANS) reactor planned to be built at ORNL. Damage propagation is postulated to occur from thermal conduction betwee...

S. H. Kim R. P. Taleyarkhan S. Navarro-Valenti V. Georgevich

1995-01-01

423

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

424

Core Burnup Characteristics of High Conversion Light Water Reactor, (1). Core Analyses for HCLWR-J1 (V/Sub M//V/sub p/ Approx. =0.8).  

National Technical Information Service (NTIS)

In order to evaluate core burnup characteristics of a high conversion light water reactor (HCLWR) with tight pitched lattice, core burnup calculation was made using two dimensional diffusion method. The volume ratio of moderator to fuel is about 0.8 in th...

K. Okumura Y. Ishiguro E. Doi

1988-01-01

425

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

426

Recognising persons by their iris patterns  

NASA Astrophysics Data System (ADS)

Iris recognition provides real-time, high confidence identification of persons by analysis of the random patterns that are visible within the iris of an eye from some distance. Because the iris is a protected, internal, organ whose random texture is epigenetic and stable over the lifespan, it can serve as a living password. Recognition decisions are made with confidence levels high enough to support rapid exhaustive searches through national-sized databases. The principle that underlies these algorithms is the failure of an efficient test of statistical independence involving more than 200 degrees-of-freedom, based on phase sequencing each iris pattern with quadrature 2D wavelets. Different persons always pass this test of statistical independence, but images from the same iris almost always fail this test of independence. Database search speeds are around 1 million persons per second per CPU. Data from 200 billion cross-comparisons between different eyes will be presented in this talk, using a database consisting of 632,500 iris images acquired in the United Arab Emirates in a networked national border-crossing security system which performs, every day, about 9 billion iris comparisons using these algorithms. Current research efforts with this technology aim to make it more tolerant of difficult conditions of iris capture, such as "iris on the move," at a distance, and off-axis.

Daugman, John

2010-04-01

427

Characterization of Neutron Fields in the Experimental Fast Reactor Joyo Mk-Iii Core  

Microsoft Academic Search

In 2003, Joyo MK-III core was upgraded to increase the irradiation testing capability. This paper describes the details of distributions of neutron flux and reaction rate in the MK-III core that was measured by characterization tests during the first two operating cycles. The calculation accuracy of the core management codes HESTIA, TORT and MCNP, was also evaluated by the measured

Shigetaka Maeda; Chikara Ito; Yasushi Ohkawachi; Takashi Sekine; Takafumi Aoyama

2009-01-01

428

Investigation of primary cooling water chemistry following the partial meltdown of Pu–Be neutron source in Tehran Research Reactor Core (TRR)  

Microsoft Academic Search

Effect of Pu–Be neutron source meltdown in core on reactor water chemistry was main aim of this study. Leaving the neutron source in the core after reactor power exceeds a few hundred Watts was the main reason for its partial meltdown.Water chemistry of primary cooling before, during and after of above incident was compared. Activity of some radio-nuclides such as

Reza Gholizadeh Aghoyeh; Hossein Khalafi

2011-01-01

429

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

430

The DF-4 fuel damage experiment in ACRR (Annual Core Research Reactor) with a BWR (Boiling Water Reactor) control blade and channel box  

SciTech Connect

The DF-4 test was an experimental investigation into the melt progression behavior of boiling water reactor (BWR) core components under high temperature severe core damage conditions. In this study 14 zircaloy clad UO{sub 2} fuel rods, and representations of the zircaloy fuel canister and stainless steel/B{sub 4}C control blade were assembled into a 0.5 m long test bundle. The test bundle was fission heated in a flowing steam environment, using the Annular Core Research Reactor at Sandia Laboratories, simulating the environmental conditions of an uncovered BWR core experiencing high temperature damage as a result residual fission product decay heating. The experimental results provide information on the thermal response of the test bundle components, the rapid exothermic oxidation of the zircaloy fuel cladding and canister, the production of hydrogen from metal-steam oxidation, and the failure behavior of the progressively melting bundle components. This information is provided in the form of thermocouple data, steam and hydrogen flow rate data, test bundle fission power data and visual observation of the damage progression. In addition to BWR background information, this document contains a description of the experimental hardware with details on how the experiment was instrumented and diagnosed, a description of the test progression, and a presentation of the on-line measurements. Also in this report are the results of a thermal analysis of the fueled test section of the fueled test section of the experiment demonstrating an overall consistency in the measurable quantities from the test. A discussion of the results is provided. 38 refs., 72 figs., 7 tabs.

Gauntt, R.O.; Gasser, R.D.; Ott, L.J. (Sandia National Labs., Albuquerque, NM (USA))

1989-11-01

431

A fundamental approach to specify thermal and pressure loadings on containment buildings of sodium cooled fast reactors during a core disruptive accident  

Microsoft Academic Search

Reactor Containment Building (RCB) is the ultimate barrier to the environment against activity release in any nuclear power plant. It has to be designed to withstand both positive and negative pressures that are credible. Core Disruptive Accident (CDA) is an important event that specifies the design basis for RCB in sodium cooled fast reactors. In this paper, a fundamental approach

K. Velusamy; P. Chellapandi; K. Satpathy; Neeraj Verma; G. R. Raviprasan; M. Rajendrakumar; S. C. Chetal

2011-01-01

432

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

NASA Astrophysics Data System (ADS)

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

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

2005-05-01

433

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

SciTech Connect

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

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

2005-05-24

434

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

435

Reactor core physics design and operating data for Cycles 1 and 2 of the Zion Unit 2 PWR power plant. Final report  

Microsoft Academic Search

A set of design and operating data relevant to Cycles 1 and 2 of the Zion Station Unit 2 pressurized water reactor (PWR) is presented. In general, these data constitute a substantial enhancement of fundamental information potentially usable in the process of validating PWR core analysis methodology. The design data is limited to the nuclear aspects of the core; thermal,

A. J. Jr. Impink; B. A. Guthrie

1979-01-01

436

Ultrasonic NDE in a reactor core with an AlN transducer  

Microsoft Academic Search

In radiation environments ultrasonic nondestructive evaluation has great potential for improving reactor safety and furthering the understanding of radiation effects and materials. In both nuclear power plants and materials test reactors, elevated temperatures and high levels of radiation present challenges to ultrasonic NDE methodologies. The challenges are primarily due to the degradation of the ultrasonic sensors utilized. We present results

D. A. Parks; B. R. Tittmann

2012-01-01

437

Adoption of iris-based authentication  

Microsoft Academic Search

Even though iris-based systems have proven to be very promising in a world where security is crucial, surprisingly enough, this means of authentication has not been given a very warm welcome from the users. In order to appropriately confront this issue, critical success factors of the deployment of networked-based systems for iris authentication - namely technical, human, and implementation aspects,

S. Mohammadi; A. Kaldi

2008-01-01

438

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

439

Iris recognition: an emerging biometric technology  

Microsoft Academic Search

This paper examines automated iris recognition as a biometrically based technology for personal identification and verification. The motivation for this endeavor stems from the observation that the human iris provides a particularly interesting structure on which to base a technology for noninvasive biometric assessment. In particular the biomedical literature suggests that irises are as distinct as fingerprints or patterns of

RICHARD P. WILDES

1997-01-01

440

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

441

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

442

[Aniridia-IOL and Artificial Iris Reconstruction].  

PubMed

Aniridia is defined as missing iris tissue which can be partial, subtotal or total. Characteristic clinical symptoms include photophobia and decreased visual acuity due to an increased light perception. In addition, disturbing cosmetic problems are prevalent. Modern iris reconstruction implants offer visual and cosmetic rehabilitation. Amongst them are aniridia intraocular lenses (IOL), iris segment implants and the "artificial iris". Different overall and pupil diameters are available for total or partial implants. At the same time aphakia or cataract can be treated when using aniridia IOLs. Intra- and extracapsular fixation is possible. The "artificial iris" can be folded and implanted through small incisions. The aesthetic results are improved significantly due to customised colour selection providing increased patient satisfaction postoperatively. PMID:23757172

Thomas, B C; Rabsilber, T M; Auffarth, G U

2013-06-06

443

Two step procedure by using a 1-D slab spectral geometry for a pebble bed reactor core analysis  

SciTech Connect

In this paper, we demonstrate a two step procedure for a pebble bed reactor core analysis. In the first step, we generate equivalent cross sections from a 1-D slab spectral geometry model with the help of the equivalence theory. In the second step, we perform a diffusion calculation by using the equivalent cross sections generated in the first step. A simple 2-D benchmark problem derived from the PMBR-400 reactor was introduced to verify the two step procedure. We compared the two step solutions with the Monte Carlo solutions for the problem and found that the two step solutions agreed well with the Monte Carlo solutions within an acceptable error range. (authors)

Lee, H. C.; Kim, K. S.; Noh, J. M.; Joo, H. K. [Korea Atomic Energy Research Inst., 150 Deokjin-Dong, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of)

2006-07-01

444

Results of analyzing accidents with core meltdown in fast reactors with sodium as the coolant  

Microsoft Academic Search

cross section is completely blocked at the inlet. The sodium temperature reaches the saturation point first in the center of the core's height about 0.5 sec after the time of blocking and the zone of boiling extends upward and downward thereafter. Evaporation of the liquid sodium film in the center lasts 0.3 sec. Evaporation begins in the upper core portion

G. B. Usynin; G. N. Vlasichev; Yu. I. Anoshkin; M. A. Semenychev; S. V. Boldin

1992-01-01

445

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

NASA Astrophysics Data System (ADS)

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

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

2010-02-01

446

ALARA Analysis for Shippingport Pressurized Water Reactor Core 2 Fuel Storage in the Canister Storage Building (CSB)  

SciTech Connect

The addition of Shippingport Pressurized Water Reactor (PWR) Core 2 Blanket Fuel Assembly storage in the Canister Storage Building (CSB) will increase the total cumulative CSB personnel exposure from receipt and handling activities. The loaded Shippingport Spent Fuel Canisters (SSFCs) used for the Shippingport fuel have a higher external dose rate. Assuming an MCO handling rate of 170 per year (K East and K West concurrent operation), 24-hr CSB operation, and nominal SSFC loading, all work crew personnel will have a cumulative annual exposure of less than the 1,000 mrem limit.

LEWIS, M.E.

2000-04-06

447

Conceptual design analysis of an MHD power conversion system for droplet-vapor core reactors. Final report  

SciTech Connect

A nuclear driven magnetohydrodynamic (MHD) generator system is proposed for the space nuclear applications of few hundreds of megawatts. The MHD generator is coupled to a vapor-droplet core reactor that delivers partially ionized fissioning plasma at temperatures in range of 3,000 to 4,000 K. A detailed MHD model is developed to analyze the basic electrodynamics phenomena and to perform the design analysis of the nuclear driven MHD generator. An incompressible quasi one dimensional model is also developed to perform parametric analyses.

Anghaie, S.; Saraph, G.

1995-12-31

448

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

449

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

450

Safety analysis for core conversion (from HEU to LEU) of Pakistan research reactor-2 (PARR2)  

Microsoft Academic Search

PARR-2 (Pakistan Research Reactor-2), an MNSR (Miniature Neutron Source Reactor) is to be converted from HEU (High Enriched Uranium) to LEU (Low Enriched Uranium) fuel, along with all current MNSRs in various other countries. The purpose of conversion is to minimize the use of HEU for non-proliferation of high-grade nuclear fuel. The present report presents thermal hydraulic and safety analyses

Ishtiaq Hussain Bokhari; Showket Pervez

2010-01-01

451

Robust iris segmentation on uncalibrated noisy images using mathematical morphology  

Microsoft Academic Search

This paper proposes a new approach for fast iris segmentation that relies on the closed nested structures of iris anatomy (the sclera is brighter than the iris, and the iris is brighter than the pupil) and on its polar symmetry. The described method applies mathematical morphology for polar\\/radial-invariant image fil- tering and for circular segmentation using shortest paths from generalized

Miguel A. Luengo-oroz; Emmanuel Faure; Jesús Angulo

2010-01-01

452

Toward Noncooperative Iris Recognition: A Classification Approach Using Multiple Signatures  

Microsoft Academic Search

This paper focus 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 Proenca

453

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

454

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

455

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

456

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

457

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

458

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

459

Iris recognition: a biometric method after refractive surgery  

Microsoft Academic Search

Iris recognition, as a biometric method, outperforms others because of its high accuracy. Iris is the visible internal organ of human, so it is stable and very difficult to be altered. But if an eye surgery must be made to some individuals, it may be rejected by iris recognition system as imposters after the surgery, because the iris pattern was

YUAN Xiao-yan; ZHOU Hao; SHI Peng-fei

2007-01-01