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1

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

2

Preliminary Safety Analysis for the IRIS Reactor  

SciTech Connect

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 transients for the whole primary and safety systems was investigated. Since the project was in a conceptual phase, the reported analyses must be considered preliminary. In fact, neither the reactor components, nor the safety systems and the reactor signal logics were completely defined at that time. Three 'conventional' design basis accidents have been preliminary evaluated: a Loss Of primary Flow Accident, a Loss Of Coolant Accident and a Loss Of Feed Water accident. The results show the effectiveness of the safety systems also in LOCA conditions; the core remains covered for the required grace period. This provides the basis to move forward to the preliminary design. (authors)

Ricotti, M.E.; Cammi, A.; Cioncolini, A.; Lombardi, C. [Politecnico di Milano, Piazza Leonardo da Vinci, 32, 20133 Milano (Italy); Cipollaro, A.; Orioto, F. [Universita di Pisa, Via Diotisalvi 2, 56126 Pisa (Italy); Conway, L.E. [Westinghouse Electric Company (United States); Barroso, A.C. [CNEN, Comissao Nacional de Energia Nuclear, Rua General Severiano 90, Rio de Janeiro, RJ-22-294-900 (Brazil)

2002-07-01

3

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

E-print Network

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

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

2003-01-01

4

Determination of a test section parameters for IRIS nuclear reactor pressurizer  

Microsoft Academic Search

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

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

2009-01-01

5

Fissioning Plasma Core Reactor  

NASA Technical Reports Server (NTRS)

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

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

2002-01-01

6

Gamma dose from activation of internal shields in IRIS reactor.  

PubMed

The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressuriser and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield. PMID:16381688

Agosteo, Stefano; Cammi, Antonio; Garlati, Luisella; Lombardi, Carlo; Padovani, Enrico

2005-01-01

7

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

8

Reactor core isolation cooling system  

DOEpatents

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

Cooke, F.E.

1992-12-08

9

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

10

Research on plasma core reactors  

NASA Technical Reports Server (NTRS)

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

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

1976-01-01

11

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

12

Lateral restraint assembly for reactor core  

DOEpatents

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

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

1986-01-01

13

Wire core reactor for NTP  

NASA Technical Reports Server (NTRS)

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

Harty, R. B.

1991-01-01

14

Pulsed Gas Core Reactor For Burst Power  

NASA Astrophysics Data System (ADS)

Studies are being performed on burst power mode gas core reactors that employ closed cycle disk MHD generators for energy conversion. The disk MHD generator is configured to be an integral part of the reactor. Consequently, significant fissioning occurs throughout the MHD duct and fission fragment induced ionization of the uranium bearing fuel gas/ working fluid is anticipated to yield the required nonequilibrium electrical conductivity (> 100 mho/m) despite the relatively low gas temperatures. Calculations performed to date have shown that the Burst Power Gas Core Reactor-Disk. MHD Generator system can achieve overall efficiencies of 25 percent effective radiator temperatures of 1200 K, reactor specific powers of 100 to 200 kWt/kg and system specific powers of 5 kWe/kg.

Dugan, Edward T.; Lear, William E.; Welch, Gerard E.

1988-04-01

15

Applications of plasma core reactors to terrestrial energy systems  

Microsoft Academic Search

Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrial applications

T. S. Lantham; F. R. Biancardi; R. J. Rodgers

1974-01-01

16

Gas core reactors for coal gasification  

NASA Technical Reports Server (NTRS)

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

Weinstein, H.

1976-01-01

17

Fabricating the Solid Core Heatpipe Reactor  

SciTech Connect

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

Ring, Peter J.; Sayre, Edwin D. [Advanced Methods and Materials, Inc., 1190 Mountain View-Alviso Road, Suite P, Sunnyvale, CA 94089 (United States); Houts, Mike [NASA Marshall Space Flight Center, Huntsville, Alabama 35812 (United States)

2006-01-20

18

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

19

Gas-core reactor power transient analysis  

NASA Technical Reports Server (NTRS)

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

Kascak, A. F.

1972-01-01

20

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

Microsoft Academic Search

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

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

2011-01-01

21

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

22

Annular core research reactor high flux neutron radiography facility  

Microsoft Academic Search

Sandia National Laboratories (SNL) has been performing neutron radiography since 1964. The radiography facilities have evolved from an aperture in a radiation exposure room in the now retired Sandia Engineering Reactor to a divergent collimator radiography facility adjacent to the core of the Annular Core Research Reactor (ACRR). The maximum thermal neutron flux achieved in these facilities has been limited

F. M. McCrory; J. G. Kelly; M. E. Vernon; D. A. Tichenor

1990-01-01

23

Applications of plasma core reactors to terrestrial energy systems  

NASA Technical Reports Server (NTRS)

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

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

1974-01-01

24

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

25

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

26

Hanging core support system for a nuclear reactor. [LMFBR  

DOEpatents

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

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

1984-04-26

27

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

28

State space modeling of reactor core in a pressurized water reactor  

NASA Astrophysics Data System (ADS)

The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

Ashaari, A.; Ahmad, T.; Shamsuddin, Mustaffa; M, Wan Munirah W.; Abdullah, M. Adib

2014-07-01

29

Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)  

NASA Technical Reports Server (NTRS)

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

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

1977-01-01

30

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

31

Nuclear waste disposal utilizing a gaseous core reactor  

NASA Technical Reports Server (NTRS)

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

Paternoster, R. R.

1975-01-01

32

DEVELOPMENT OF CORE ELEMENTS FOR THE ENRICO FERMI POWER REACTOR  

Microsoft Academic Search

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

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

1958-01-01

33

Local-spectrum-modified fast reactor cores with hydrides  

Microsoft Academic Search

the application of different hydride materials to core components except for the driver fuel assembly is studied to achieve high core performances for fast reactor by reducing neutron energy and increasing reaction rates locally. The mixtures of absorber materials such as Gd and Zr-hydride are employed to control materials to increase reactivity worth. An optimized composition and layout of the

Yokoyama Tsugio; Konashi Kenji; Iwasaki Tomohiko; Terai Takayuki; Yamawaki Michio

2006-01-01

34

Shield Design for a Space Based Vapor Core Reactor  

SciTech Connect

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

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

2002-07-01

35

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

36

Hanging core support system for a nuclear reactor  

DOEpatents

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

Burelbach, James P. (Glen Ellyn, IL); Kann, William J. (Park Ridge, IL); Pan, Yen-Cheng (Naperville, IL); Saiveau, James G. (Hickory Hills, IL); Seidensticker, Ralph W. (Wheaton, IL)

1987-01-01

37

Multi-megawatt pin core space reactor  

Microsoft Academic Search

The design of multimegawatt nuclear power systems that use the hydrogen needed for platform cooling as the working fluid in an open thermodynamic cycle is discussed. The hydrogen is heated by a pin-fuel, fast-spectrum reactor and generates power through a pair of counterrotating turbines that drive four wound rotor alternatives. An overview is given of the system, concentrating on features

R. J. Hornung; E. Normand; A. Stevens; K. R. Teare

1989-01-01

38

Structural materials for breeder reactor cores and coolant circuits  

SciTech Connect

The structural components of principal interest in LMFBR cores and cooling circuits include the reactor vessel, primary and secondary piping, intermediate heat exchanger (IHX), and steam generator. Load-bearing components inside the vessel, among these the fuel cladding and duct, are also included. The operating conditions present in a fast-breeder nuclear reactor impose a number of requirements on the mechanical, physical, and neutronic properties of the materials used to construct these components.

Diercks, D.R.

1984-02-01

39

Thermal barrier and support for nuclear reactor fuel core  

DOEpatents

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

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

1987-01-01

40

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

41

Gas core reactors for actinide transmutation and breeder applications  

NASA Technical Reports Server (NTRS)

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

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

1978-01-01

42

Feasibility study of full-reactor gas core demonstration test  

NASA Technical Reports Server (NTRS)

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

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

1973-01-01

43

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

44

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

45

Gamma thermometer based reactor core liquid level detector  

DOEpatents

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, Thomas J. (Knoxville, TN)

1983-01-01

46

NEUTRON RADIOGRAPHY (NRAD) REACTOR 64-ELEMENT CORE UPGRADE  

SciTech Connect

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

John D. Bess

2014-03-01

47

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

48

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

49

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

50

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

51

Development and Assessment of Advanced Reactor Core Protection System  

NASA Astrophysics Data System (ADS)

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

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

52

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

Microsoft Academic Search

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

T. A. Doyle; A. L. Hess

1962-01-01

53

System Study: Reactor Core Isolation Cooling 1998–2012  

SciTech Connect

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

T. E. Wierman

2013-10-01

54

Dynamic analysis of gas-core reactor system  

NASA Technical Reports Server (NTRS)

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

Turner, K. H., Jr.

1973-01-01

55

Gas core reactor concepts and technology - Issues and baseline strategy  

NASA Technical Reports Server (NTRS)

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

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

1991-01-01

56

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

57

Local-spectrum-modified fast reactor cores with hydrides  

SciTech Connect

the application of different hydride materials to core components except for the driver fuel assembly is studied to achieve high core performances for fast reactor by reducing neutron energy and increasing reaction rates locally. The mixtures of absorber materials such as Gd and Zr-hydride are employed to control materials to increase reactivity worth. An optimized composition and layout of the control materials has shown the feature of burnable poison even in the fast reactor where the reaction rate of absorber nuclides is increased enough to annihilate themselves at the end of cycles. The radial blanket made of the mixture of oxide uranium and Zr hydride is examined to decrease the required thickness as well as to achieve a non-proliferation feature in plutonium isotope compositions in discharged fuels. The shielding performance of radial shield made of Zr-hydride is evaluated to decrease the whole core diameter. Special fuel assemblies mixed with minor actinides and Zr-hydride located at the core peripherals are studied to transmute the minor actinides to fissionable materials effectively. The results has indicates that the application of the hydride materials will increase the core performances twice or triple in general. (authors)

Tsugio, Yokoyama [Aitel Corporation, 8 Shinsugita-cho, Isogo-ku, Yokohama, 235-8523 (Japan); Kenji, Konashi [Institute for Materials Research, Tohoku University, Oarai, Ibaraki-ken, 311-1313 (Japan); Tomohiko, Iwasaki [Department of Quantum Science and Energy Engineering, Tohoku Universiyt, Aoba 6-6-01-2, Aramaki, Aoba-ku, Sendai, 980-8579 (Japan); Takayuki, Terai [Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo, 113-8656 (Japan); Michio, Yamawaki [Department of Applied Science, Tokai University, 1117 Kitakaname, Hiratsuka, Kanagawa 259- 1292 (Japan)

2006-07-01

58

One pass core design of a super fast reactor  

SciTech Connect

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

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

2013-07-01

59

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

NASA Astrophysics Data System (ADS)

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

2010-12-01

60

Post impact behavior of mobile reactor core containment systems  

NASA Technical Reports Server (NTRS)

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

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

1972-01-01

61

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

62

Gas core reactors for actinide transmutation. [uranium hexafluoride  

NASA Technical Reports Server (NTRS)

The preliminary design of a uranium hexafluoride actinide transmutation reactor to convert long-lived actinide wastes to shorter-lived fission product wastes was analyzed. It is shown that externally moderated gas core reactors are ideal radiators. They provide an abundant supply of thermal neutrons and are insensitive to composition changes in the blanket. For the present reactor, an initial load of 6 metric tons of actinides is loaded. This is equivalent to the quantity produced by 300 LWR-years of operation. At the beginning, the core produces 2000 MWt while the blanket generates only 239 MWt. After four years of irradiation, the actinide mass is reduced to 3.9 metric tonnes. During this time, the blanket is becoming more fissile and its power rapidly approaches 1600 MWt. At the end of four years, continuous refueling of actinides is carried out and the actinide mass is held constant. Equilibrium is essentially achieved at the end of eight years. At equilibrium, the core is producing 1400 MWt and the blanket 1600 MWt. At this power level, the actinide destruction rate is equal to the production rate from 32 LWRs.

Clement, J. D.; Rust, J. H.; Wan, P. T.; Chow, S.

1979-01-01

63

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

64

MCNP/MCNPX model of the annular core research reactor.  

SciTech Connect

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

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

2006-10-01

65

A Solid Core Heatpipe Reactor with Cylindrical Thermoelectric Converter Modules  

NASA Astrophysics Data System (ADS)

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

Sayre, Edwin D.; Vaidyanathan, Sam

2006-01-01

66

Aerosol formation and growth in a laminar core reactor  

SciTech Connect

A novel aerosol flow reactor is described in which the core of a lamina-flow of premixed reactants is irradiated to produce particles along the axis of the flow. The reactor was studied experimentally by irradiating an NH/sub 3//C/sub 3/H/sub 6/NO/sub 2//air mixture to produce NH/sub 4/NO/sub 3/ aerosol. A theory developed for particle formation and growth in the core reactor, accounting for radial diffusion of the condensing species and the parabolic velocity profile, explained the behavior of the system. The theory was formulated in terms of moments of the particle size distribution. The formation of NH/sub 4/NO/sub 3/ particles from HNO/sub 3/ and NH/sub 3/ vapors followed classical nucleation theory with HNO/sub 3/ considered as the monomer. Values of the surface tension of the solid NH/sub 4/ NO/sub 3/ and the rate of HNO/sub 3/ formation were determined by comparing theory and experiment.

Kodas, T.T.; Pratsinis, S.E.; Friedlander, S.K.

1985-01-01

67

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

68

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

E-print Network

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

Paris-Sud XI, Université de

69

Criticality analysis of pulse core and laser module coupled small reactor with low enriched uranium  

Microsoft Academic Search

Nuclear-pumped laser can directly convert nuclear energy to optical energy. A coupled reactor which consists of two pulse cores with highly enriched metallic uranium and a subcritical thermal laser module with highly enriched metallic uranium is one of the reactors for nuclear-pumped laser. In this paper, criticality analysis of a coupled reactor which consists of pulse cores with 20% enriched

Hiroki Takezawa; Toru Obara; Andrey Gulevich; Oleg Kukharchuk

2008-01-01

70

Design and analysis of a nuclear reactor core for innovative small light water reactors  

NASA Astrophysics Data System (ADS)

In order to address the energy needs of developing countries and remote communities, Oregon State University has proposed the Multi-Application Small Light Water Reactor (MASLWR) design. In order to achieve five years of operation without refueling, use of 8% enriched fuel is necessary. This dissertation is focused on core design issues related with increased fuel enrichment (8.0%) and specific MASLWR operational conditions (such as lower operational pressure and temperature, and increased leakage due to small core). Neutron physics calculations are performed with the commercial nuclear industry tools CASMO-4 and SIMULATE-3, developed by Studsvik Scandpower Inc. The first set of results are generated from infinite lattice level calculations with CASMO-4, and focus on evaluation of the principal differences between standard PWR fuel and MASLWR fuel. Chapter 4-1 covers aspects of fuel isotopic composition changes with burnup, evaluation of kinetic parameters and reactivity coefficients. Chapter 4-2 discusses gadolinium self-shielding and shadowing effects, and subsequent impacts on power generation peaking and Reactor Control System shadowing. The second aspect of the research is dedicated to core design issues, such as reflector design (chapter 4-3), burnable absorber distribution and programmed fuel burnup and fuel use strategy (chapter 4-4). This section also includes discussion of the parameters important for safety and evaluation of Reactor Control System options for the proposed core design. An evaluation of the sensitivity of the proposed design to uncertainty in calculated parameters is presented in chapter 4-5. The results presented in this dissertation cover a new area of reactor design and operational parameters, and may be applicable to other small and large pressurized water reactor designs.

Soldatov, Alexey I.

71

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

72

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

73

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

74

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

75

EVALUATION OF REACTOR CORE MATERIALS FOR A GAS-COOLED REACTOR EXPERIMENT  

Microsoft Academic Search

An evaluation of core materials for a gas-cooled reactor is being made. ; Work on the ZrH\\/sub n\\/ moderator has been confined to the high-hydrogen or delta-; phase material. Methods for preparing sound hydride bodies of the highhydrogen ; composition have been developed. Both solid hydride and hydride powder compacts ; are being clad by a pressure-bonding technique. The hot

1957-01-01

76

Depletion analysis of the UMLRR reactor core using MCNP6  

NASA Astrophysics Data System (ADS)

Accurate knowledge of the neutron flux and temporal nuclide inventory in reactor physics calculations is necessary for a variety of application in nuclear engineering such as criticality safety, safeguards, and spent fuel storage. The Monte Carlo N- Particle (MCNP6) code with integrated buildup depletion code (CINDER90) provides a high-fidelity tool that can be used to perform 3D, full core simulation to evaluate fissile material utilization, and nuclide inventory calculations as a function of burnup. The University of Massachusetts Lowell Research Reactor (UMLRR) reactor has been modeled with the deterministic based code, VENTURE and with an older version of MCNP (MCNP5). The MIT developed MCODE (MCNP ORIGEN DEPLETION CODE) was used previously to perform some limited depletion calculations. This work chronicles the use of MCNP6, released in June 2013, to perform coupled neutronics and depletion calculation. The results are compared to previously benchmarked results. Furthermore, the code is used to determine the ratio of fission products 134Cs and 137Cs (burnup indicators), and the resultant ratio is compared to the burnup of the UMLRR.

Odera, Dim Udochukwu

77

Prosthetic iris devices.  

PubMed

Congenital iris defects may usually present either as subtotal aniridia or colobomatous iris defects. Acquired iris defects are secondary to penetrating iris injury, iatrogenic after surgical excision of iris tumours, collateral trauma after anterior segment surgery, or can be postinflammatory in nature. These iris defects can cause severe visual disability in the form of glare, loss of contrast sensitivity, and loss of best corrected visual acuity. The structural loss of iris can be reconstructed with iris suturing, use of prosthetic iris implants, or by a combination of these, depending on the relative amount of residual iris stromal tissue and health of the underlying pigment epithelium. Since the first implant of a black iris diaphragm posterior chamber intraocular lens in 1994, advances in material and design technology over the last decade have led to advances in the prosthetic material, surgical technique, and instrumentation in the field of prosthetic iris implants. In this article, we review the classification of iris defects, types of iris prosthetic devices, implantation techniques, and complications. PMID:24513351

Srinivasan, Sathish; Ting, Darren S J; Snyder, Michael E; Prasad, Somdutt; Koch, Hans-Reinhard

2014-02-01

78

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

NASA Astrophysics Data System (ADS)

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

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

2014-06-01

79

Data mining reactor fuel grab load trace data to support nuclear core condition monitoring  

Microsoft Academic Search

A critical component of an advanced-gas cooled reactor (AGR) station is the graphite core. As a station ages, the graphite bricks that comprise the core can distort and may eventually crack. As the core cannot be replaced the core integrity ultimately determines the station life. Monitoring these distortions is usually restricted to the routine outages, which occur every few years,

Graeme M. West; Gordon J. Jahn; S. D. J. McArthur; James R. McDonald; Jim Reed

2006-01-01

80

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

NASA Astrophysics Data System (ADS)

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

Determan, William R.; Otting, William D.

1992-01-01

81

Post impact behavior of mobile reactor core containment systems.  

NASA Technical Reports Server (NTRS)

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

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

1972-01-01

82

Plasma core reactor simulations using RF uranium seeded argon discharges  

NASA Technical Reports Server (NTRS)

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

Roman, W. C.

1976-01-01

83

Plasma core reactor simulations using RF uranium seeded argon discharges  

NASA Technical Reports Server (NTRS)

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

Roman, W. C.

1975-01-01

84

Core reactivity estimation in space reactors using recurrent dynamic networks  

NASA Technical Reports Server (NTRS)

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

Parlos, Alexander G.; Tsai, Wei K.

1991-01-01

85

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

86

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

E-print Network

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

Paris-Sud XI, Université de

87

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

DOEpatents

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

Tokarz, R.D.

1981-10-27

88

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

Microsoft Academic Search

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

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

1984-01-01

89

Preliminary neutronics calculations for conversion of the Tehran research reactor core from HEU to LEU fuel  

Microsoft Academic Search

The 5-MW highly enriched uranium (HEU)-fueled Tehran Research Reactor is considered for conversion to high-density, low-enriched uranium (LEU) fuel. A preliminary neutronics calculation is performed as part of the conversion goal. In this study, two cores are considered: the HEU reference core and a proposed LEU core similar to the reference core, and a proposed LEU core similar to the

Nejat; S. M. R

1993-01-01

90

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

NASA Technical Reports Server (NTRS)

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

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

1991-01-01

91

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

SciTech Connect

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

Dobranich, D.

1987-05-01

92

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

SciTech Connect

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

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

2013-07-01

93

Fault current limiter-predominantly resistive behavior of a BSCCO shielded-core reactor  

SciTech Connect

Tests were conducted to determine the electrical and magnetic characteristics of a superconductor shielded core reactor (SSCR). The results show that a closed-core SSCR is predominantly a resistive device and an open-core SSCR is a hybrid resistive/inductive device. The open-core SSCR appears to dissipate less than the closed-core SSCR. However, the impedance of the open-core SSCR is less than that of the closed-core SSCR. Magnetic and thermal diffusion are believed to be the mechanism that facilitates the penetration of the superconductor tube under fault conditions.

Ennis, M. G.; Tobin, T. J.; Cha, Y. S.; Hull, J. R.

2000-06-30

94

Modelling Pressurized Water Reactor cores in terms of porous media  

NASA Astrophysics Data System (ADS)

The aim of this study is to develop a tractable model of a nuclear reactor core taking the complexity of the structure (including its nonlinear behaviour) and fluid flow coupling into account. The mechanical behaviour modelling includes the dynamics of both the fuel assemblies and the fluid. Each rod bundle is modelled in the form of a deformable porous medium; then, the velocity field of the fluid and the displacement field of the structure are defined over the whole domain. The fluid and the structure are first modelled separately, before being linked together. The equations of motion for the structure are obtained using a Lagrangian approach and, to be able to link up the fluid and the structure, the equations of motion for the fluid are obtained using an arbitrary Lagragian Eulerian approach. The finite element method is applied to spatially discretize the equations. Simulations are performed to analyse the effects of the characteristics of the fluid and of the structure. Finally, the model is validated with a test involving two fuel assemblies, showing good agreement with the experimental data.

Ricciardi, G.; Bellizzi, S.; Collard, B.; Cochelin, B.

2009-01-01

95

Criticality safety analysis on fissile materials in Fukushima reactor cores  

SciTech Connect

The present study focuses on the criticality analysis for geological disposal of damaged fuels from Fukushima reactor cores. Starting from the basic understanding of behaviors of plutonium and uranium, a scenario sequence for criticality event is considered. Due to the different mobility of plutonium and uranium in geological formations, the criticality safety is considered in two parts: (1) near-field plutonium system and (2) far-field low enriched uranium (LEU) system. For the near-field plutonium system, a mathematical analysis for pure-solute transport was given, assuming a particular buffer material and waste form configuration. With the transport and decay of plutonium accounted, the critical mass of plutonium was compared with the initial load of a single canister. Our calculation leads us to the conclusion that our system with the initial loading being the average mass of plutonium in an assembly just before the accident is very unlikely to become critical over time. For the far-field LEU system, due to the uncertainties in the geological and geochemical conditions, calculations were made in a parametric space that covers the variation of material compositions and different geometries. Results show that the LEU system could not remain sub-critical within the entire parameter space assumed, although in the iron-rich rock, the neutron multiplicity is significantly reduced.

Liu, Xudong; Lemaitre-Xavier, E.; Ahn, Joonhong [Department of Nuclear Engineering, University of California, Berkeley, Berkeley, CA 94720 (United States); Hirano, Fumio [Japan Atomic Energy Agency, Geological Isolation Research and Development Directorate, Tokai-mura, Ibaraki 319-1194 (Japan)

2013-07-01

96

McCARD for Neutronics Design and Analysis of Research Reactor Cores  

NASA Astrophysics Data System (ADS)

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

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

2014-06-01

97

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

NASA Technical Reports Server (NTRS)

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

Weinstein, H.; Lavan, Z.

1975-01-01

98

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

SciTech Connect

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

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

2013-02-01

99

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

100

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

E-print Network

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

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

2008-01-01

101

IRI in Windows environment  

NASA Astrophysics Data System (ADS)

The International Reference Ionosphere (IRI) is a widely used model recommended for international use by URSI and COSPAR. An interface to Windows has been developed and the stand-alone IRI program is executable on PC systems using Windows 95/98 or Windows NT as the platform. This IRI-Windows version allows computation and plotting of any of the IRI parameters as function of any one or two variables: height, latitude, longitude, UT or LT time, and solar activity. The user can conveniently specify the variables and various options of the model. This paper gives a brief description of the program features and the operating procedure.

Huang, X.; Reinisch, B. W.; Bilitza, D.

102

IRIS Agenda and Literature Searches  

EPA Science Inventory

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

103

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

E-print Network

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

Kersting, Alyssa (Alyssa Rae)

2011-01-01

104

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

E-print Network

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

Aldrich, David C.

1979-01-01

105

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

106

Adaptability Studies with Bearded Iris in Texas.  

E-print Network

iris gratis: Kenwood Iris Gardens, Cincinnati, O., Longfield Iri5 Farm, Bluff- ton, Ind., Otwell Iris Fields, Carlinville, Ill., Earl Salbach, Berkeley, Calif., Schreiner's Iris Gardens, St. Paul, Minn., Carl Starker, Jennings Lodge, Ore., Treholme...

Yarnell, S. H. (Sidney Howe)

1942-01-01

107

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

108

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

109

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

Microsoft Academic Search

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

A. Middleton; G. C. Law

1974-01-01

110

Gas Core Reactor-MHD Power System with Cascading Power Cycle  

Microsoft Academic Search

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

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

2002-01-01

111

Nuclear design of the burst power ultrahigh temperature UF4 vapor core reactor system  

Microsoft Academic Search

Static and dynamic neutronic analyses are being performed, as part of an integrated series of studies, on an innovative burst power UF4 Ultrahigh Temperature Vapor Core Reactor (UTVR)\\/Disk Magnetohydrodynamic (MHD) generator for space nuclear power applications. This novel reactor concept operates on a direct, closed Rankine cycle in the burst power mode (hundreds of MWe for thousands of seconds). The

Samer D. Kahook; Edward T. Dugan

1991-01-01

112

Nuclear design of the burst power ultrahigh temperature UF sub 4 vapor core reactor system  

Microsoft Academic Search

Static and dynamic neutronic analyses are being performed, as part of an integrated series of studies, on an innovative burst power UFâ Ultrahigh Temperature Vapor Core Reactor (UTVR)\\/Disk Magnetohydrodynamic (MHD) generator for space nuclear power applications. This novel reactor concept operates on a direct, closed Rankine cycle in the burst power mode (hundreds of MW{sub e} for thousands of seconds).

S. D. Kahook; E. T. Dugan

1991-01-01

113

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

NASA Astrophysics Data System (ADS)

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

Qvist, Staffan Alexander

114

Simulation of reactivity transients in a miniature neutron source reactor core  

Microsoft Academic Search

Computer simulation was carried out for reactivity induced transients in a HEU core of a tank-in-pool reactor, a miniature neutron source reactor (MNSR). The reactivity transients without scram at initial power of 3 W were studied. From the low power level, the power steadily increased with time and then rose sharply to higher peak values followed by a gradual decrease

E. H. K. Akaho; B. T. Maakuu

2002-01-01

115

Design Study of Small Lead-Cooled Fast Reactor Cores Using SiC Cladding and Structure  

Microsoft Academic Search

Neutronics of a reactor core with SiC cladding and structure was compared with that with steel cladding and structure analytically for small lead-cooled fast reactors. Uranium nitride fuel was used for this reactor. U235 enrichment was 11% in inner core and 13% in outer core for relatively flat neutron flux distributions and power density distribution. The core design was optimized

Abu Khalid Rivai; Minoru Takahashi

2007-01-01

116

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

PubMed

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

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

2013-12-01

117

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

118

Application of advanced core process monitoring procedures in German power reactors  

Microsoft Academic Search

The nuclear reactor core design and the nuclear fuel management have been changed remarkable during the last few years. This development was initiated by increasing costs for the fuel recycling and nuclear waste storage. The fuel material, the fuel pellet fabrication, the fuel assembly structure and the core composition have been varied to get an effective fuel exploitation. Based on

J. Pohlus

2003-01-01

119

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

Microsoft Academic Search

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

J. K. Wheeler; A. Sesonske

1986-01-01

120

An evolutionary approach for a compact-split-core reactor.  

NASA Technical Reports Server (NTRS)

An economical approach for advanced reactor power development is presented, and systems that result from the several stages of this plan are described. The development starts with a highly modularized heat-pipe, radioisotopic design and evolves into a low-specific-weight, high-performance reactor system.

Breitwieser, R.; Lantz, E.

1973-01-01

121

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

122

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

123

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

Microsoft Academic Search

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

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

2011-01-01

124

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

NASA Astrophysics Data System (ADS)

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 objectives; coupled with MHD conversion, system efficiencies of 40 percent for open cycle operation and heat rejection temperatures of 1500 K or higher for closed cycle operation are anticipated. The design of the gas core reactor/MHD generator configuration to directly produce pulsed electrical power, thereby alleviating external power conditioning requirements, is also a research objective.

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

125

a Dosimetry Assessment for the Core Restraint of AN Advanced Gas Cooled Reactor  

NASA Astrophysics Data System (ADS)

This paper describes calculations of neutron damage rates within the core restraint structures of Advanced Gas Cooled Reactors (AGRs). Using advanced features of the Monte Carlo radiation transport code MCBEND, and neutron source data from core follow calculations performed with the reactor physics code PANTHER, a detailed model of the reactor cores of two of British Energy's AGR power plants has been developed for this purpose. Because there are no relevant neutron fluence measurements directly supporting this assessment, results of benchmark comparisons and successful validation of MCBEND for Magnox reactors have been used to estimate systematic and random uncertainties on the predictions. In particular, it has been necessary to address the known under-prediction of lower energy fast neutron responses associated with the penetration of large thicknesses of graphite.

Thornton, D. A.; Allen, D. A.; Tyrrell, R. J.; Meese, T. C.; Huggon, A. P.; Whiley, G. S.; Mossop, J. R.

2009-08-01

126

Some Nuclear Calculations of U-235-D2O Gaseous-Core Cavity Reactors  

NASA Technical Reports Server (NTRS)

The results of a multigroup, diffusion theory study of spherical gaseous-core cavity reactors are presented in this report. The reactor cavity of gaseous U235 is enclosed by a region of hydrogen gas and is separated from an external D2O moderator-reflector by a zirconium structural shell. Some cylindrical reactors are also investigated. A parametric study of spherical reactors indicates that, for the range of variables studied, critical mass increases as: (1) Fuel region is compressed within the reactor cavity, (2) moderator thickness is decreased, (3) structural shell thickness is increased, and (4) moderator temperature is increased. A buckling analogy is used to estimate the critical mass of fully reflected cylindrical reactors from spherical results without fuel compression. For a reactor cavity of a 120-centimeter radius uniformly filled with fuel, no structural shell, a moderator temperature of 70 F, and a moderator thickness of 100 centimeters, the critical mass of a spherical reactor is 3.1 kilograms while that of a cylinder with a length-to-diameter ratio of 1.0 (L/D = 1) is approximately 3.8 kilograms and, with L/D = 2, 5.9 kilograms. For the range of variables considered for U235-D2O gaseous-core cavity reactors, the systems are characterized by 95 to 99 percent thermal absorptions, with the flux reaching a maximum in the moderator about 10 to 15 centimeters from the reactor cavity.

Ragsdale, Robert G.; Hyland, Robert E.

1961-01-01

127

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

128

Nuclear reactor with low-level core coolant intake  

DOEpatents

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

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

1993-01-01

129

Preparations to ship the damaged TMI2 reactor core  

Microsoft Academic Search

The March 1979 accident at the Three Mile Island Nuclear Power Station (TMI) which damaged the core of Unit 2 resulted in numerous scientific and technical challenges. Some of those challenges include 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 the preparations for

R. C. Schmitt; G. J. Quinn

1985-01-01

130

Heat exchanger for reactor core and the like  

DOEpatents

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

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

1986-01-01

131

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

SciTech Connect

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

B. Boer; A. M. Ougouag

2010-09-01

132

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

133

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

NASA Astrophysics Data System (ADS)

The solid-core, gas-cooled, Submersion-Subcritical Safe Space (S?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?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, respectively.

King, Jeffrey C.; El-Genk, Mohamed S.

2006-01-01

134

Gas Core Reactor with Magnetohydrodynamic Power System and Cascading Power Cycle  

Microsoft Academic Search

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

Blair M. Smith; Samim Anghaie

2004-01-01

135

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

136

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

PubMed

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

Lashkari, A; Khalafi, H; Kazeminejad, H

2013-05-01

137

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

PubMed Central

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

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

2013-01-01

138

Prospects for poisoning reactor cores of the future  

Microsoft Academic Search

In this article, the most interesting rare earth nuclides for neutronics (gadolinium, samarium, erbium, europium and dysprosium) are studied and compared from the point of view of their possibilities for the control of potential core reactivity in PWRs to increase the cycle length. By modifying the absorption of the fuel network, burnable absorbers, modify the spectrum to a lesser or

Marielle Asou; Jacques Porta

1997-01-01

139

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

NASA Astrophysics Data System (ADS)

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

Damian, F.; Brun, E.

2014-06-01

140

Iris Recognition for Human Identification  

NASA Astrophysics Data System (ADS)

Iris recognition system is the biometric identification system. Iris has an intricate structure, uniqueness, stability, and natural protection. Due to these features of the iris it can be used for biometric identification. This system gives better performance than other biometric identification systems. A novel eyelash removal method for preprocessing of human iris images in a human iris recognition system is presented.. Discrete cosine transform (DCT) method is used for feature extraction. For matching of two-iris code Hamming distance calculation is used. EER value must be less for the optimum performance of the system.

Alandkar, Lajari; Gengaje, Sachin

2010-11-01

141

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

NASA Technical Reports Server (NTRS)

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

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

1996-01-01

142

Cosmic Ray Radiography of the Damaged Cores of the Fukushima Reactors  

E-print Network

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

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

2012-09-13

143

Cosmic Ray Radiography of the Damaged Cores of the Fukushima Reactors  

E-print Network

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

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

2012-01-01

144

Cosmic Ray Radiography of the Damaged Cores of the Fukushima Reactors  

NASA Astrophysics Data System (ADS)

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

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

2012-10-01

145

Design Study of Small Lead-Cooled Fast Reactor Cores Using SiC Cladding and Structure  

NASA Astrophysics Data System (ADS)

Neutronics of a reactor core with SiC cladding and structure was compared with that with steel cladding and structure analytically for small lead-cooled fast reactors. Uranium nitride fuel was used for this reactor. U235 enrichment was 11% in inner core and 13% in outer core for relatively flat neutron flux distributions and power density distribution. The core design was optimized using natural uranium blanket and nitride fuel for long life-core with reshuffling interval of 15 years. The analytical result indicated that neutron energy spectrum was slightly softer in the core with the SiC cladding and structure than that with steel cladding and structure. The SiC type reactor was designed to have criticality at the beginning of cycle (BOC), although the steel type reactor could not have criticality with the same size and geometry. In other words, the SiC type core can be designed smaller than the steel type core. The peak power densities could remain constant over the reactor operation. The consumption capability of uranium was quite high, i.e. 10% for 125 MWt reactor and 18.4% for 375 MWt reactor at the end of cycle (EOC).

Rivai, Abu Khalid; Takahashi, Minoru

146

Application of gaseous core reactors for transmutation of nuclear waste  

NASA Technical Reports Server (NTRS)

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

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

1976-01-01

147

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

Microsoft Academic Search

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

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

1982-01-01

148

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

149

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

150

Characterization of the fast neutron irradiation facility of the Portuguese Research Reactor after core conversion.  

PubMed

The fast neutron irradiation facility of the Portuguese Research Reactor was characterized after the reduction in uranium enrichment and rearrangement of the core configuration. In this work we report on the determination of the hardness parameter and the 1MeV equivalent neutron flux along the facility, in the new irradiation conditions, following ASTM E722 standard. PMID:21071234

Marques, J G; Sousa, M; Santos, J P; Fernandes, A C

2011-08-01

151

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

Microsoft Academic Search

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

McClung

1962-01-01

152

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

153

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

NASA Technical Reports Server (NTRS)

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

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

2010-01-01

154

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

NASA Technical Reports Server (NTRS)

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

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

2009-01-01

155

Long-life neutron detector for instrumentation of a nuclear reactor core  

Microsoft Academic Search

The invention concerns a neutron detector adapted for use in the core ; instrumentation of nuclear reactors. It comprises an emitter electrode which, ; upon irradiation with neutrons, emits electrons by means of (n,$gamma$) ; processes, a collector electrode, and a dielectric disposed between the emitter ; and the collector electrode. Thulium 169 or terbium 159 is used as the

Klar

1975-01-01

156

Full Core Reactor Analysis: Running Denovo on Jaguar  

SciTech Connect

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

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

2012-01-01

157

Full Core Reactor Analysis: Running Denovo on Jaguar  

SciTech Connect

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

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

2013-01-01

158

Full core reactor analysis: Running Denovo on Jaguar  

SciTech Connect

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

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

2012-07-01

159

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

SciTech Connect

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

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

2013-07-01

160

A demonstration of a whole core neutron transport method in a gas cooled reactor  

SciTech Connect

This paper illustrates a capability of the whole core transport method COMET. Building on previous works which demonstrated the accuracy of the method, this work serves to emphasize the robust capability of the method while also accentuating its efficiency. A set of core configurations is presented based on an operating gas-cooled thermal reactor, Japan's HTTR, and COMET determines the eigenvalue and fission density profile throughout each core configuration. Results for core multiplication factors are compared to MCNP for accuracy and also to compare runtimes. In all cases, the values given by COMET differ by those given by MCNP by less than the uncertainty inherent in the stochastic solution procedure, however, COMET requires runtimes shorter on the order of a few hundred. Figures are provided illustrating the whole core fission density profile, with segments of pins explicitly modeled individually, so that pin-level neutron flux behavior can be seen without any approximation due to simplification strategies such as homogenization. (authors)

Connolly, K. J.; Rahnema, F. [Nuclear and Radiological Engineering and Medical Physics Programs, George W. Woodruff School, Georgia Institute of Technology, Atlanta, GA (United States)

2013-07-01

161

MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core  

SciTech Connect

In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO{sub 2} fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory.

Arslan, M.; Villele, E. de; Gauthier, J.C.; Marincic, A. [AREVA - Tour AREVA, 1 Place Jean Millier, 92084 Paris La Defense (France)

2013-07-01

162

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

163

Critical and power experiments on the low-enriched uranium core of the upgraded Pakistan Research Reactor1  

Microsoft Academic Search

The Pakistan Research Reactor was converted from 93% highly enriched uranium fuel to 20% low-enriched uranium fuel in October 1991. The reactor power was also upgraded from 5 to 9 MW. A series of critical and power experiments were performed on the new core for verification of design data and to determine the nuclear performance of the reactor. The characteristics

S. A. Ansari; M. Iqbal; L. Ali; N. M. Butt

1994-01-01

164

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

SciTech Connect

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

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

2013-07-01

165

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

166

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

Microsoft Academic Search

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

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

2010-01-01

167

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

168

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

169

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

SciTech Connect

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

PARMA JR.,EDWARD J.

2000-01-01

170

Mixed enrichment core design for the NC State University PULSTAR Reactor  

SciTech Connect

The North Carolina State University PULSTAR Reactor license was renewed for an additional 20 years of operation on April 30, 1997. The relicensing period added additional years to the facility operating time through the end of the second license period, increasing the excess reactivity needs as projected in 1988. In 1995, the Nuclear Reactor Program developed a strategic plan that addressed the future maintenance, development, and utilization of the facility. Goals resulting from this plan included increased academic utilization of the facility in accordance with its role as a university research facility, and increased industrial service use in accordance with the mission of a land grant university. The strategic plan was accepted, and it is the intent of the College of Engineering to operate the PULSTAR Reactor as a going concern through at least the end of the current license period. In order to reach the next relicensing review without prejudice due to low excess reactivity, it is desired to maintain sufficient excess reactivity so that, if relicensed again, the facility could continue to operate without affecting users until new fuel assistance was provided. During the NC State University license renewal, the operation of the PULSTAR Reactor at the State University of New York at Buffalo (SUNY Buffalo) was terminated. At that time, the SUNY Buffalo facility had about 240 unused PULSTAR Reactor fuel pins with 6% enrichment. The objective of the work reported here was to develop a mixed enrichment core design for the NC State University PULSTAR reactor which would: (1) demonstrate that 6% enriched SUNY buffalo fuel could be used in the NC State University PULSTAR Reactor within the existing technical specification safety limits for core physics parameters; (2) show that use of this fuel could permit operating the NC State University PULSTAR Reactor to 2017 with increased utilization; and (3) assure that the decision whether or not to relicense the facility would not be prejudiced by reduced operations due to low excess reactivity.

Mayo, C.W.; Verghese, K.; Huo, Y.G.

1997-12-01

171

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

172

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

SciTech Connect

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

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

2012-07-01

173

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

PubMed

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

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

2015-02-01

174

Nuclear design of the burst power ultrahigh temperature UF4 vapor core reactor system  

NASA Astrophysics Data System (ADS)

Static and dynamic neutronic analyses are being performed, as part of an integrated series of studies, on an innovative burst power UF4 Ultrahigh Temperature Vapor Core Reactor (UTVR)/Disk Magnetohydrodynamic (MHD) generator for space nuclear power applications. This novel reactor concept operates on a direct, closed Rankine cycle in the burst power mode (hundreds of MWe for thousands of seconds). The fuel/working fluid is a mixture of UF4 and metal fluoride. Preliminary calculations indicate high overall system efficiencies (?20%), small radiator size (?5 m2/MWe), and high specific power (?5 kWe/kg). Neutronic analysis has revealed a number of attractive features for this novel reactor concept. These include some unique and very effective inherent negative reactivity control mechanisms such as the vapor-fuel density power coefficient of reactivity, the direct neutronic coupling among the multiple fissioning core regions (the central vapor core and the surrounding boiler columns), and the mass flow coupling feedback between the fissioning cores.

Kahook, Samer D.; Dugan, Edward T.

1991-01-01

175

Advanced Core Design And Fuel Management For Pebble-Bed Reactors  

SciTech Connect

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

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

2004-10-01

176

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

NASA Astrophysics Data System (ADS)

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

Mironidis, Anastasios Pantelis

1998-12-01

177

IRI topside correction  

E-print Network

The topside segment of the International Reference Ionosphere (IRI) electron density model (and also of the Bent model) is based on the limited amount of topside data available at the time (approx 40,000 Alouette 1 profiles). Being established from such a small database it is therefore not surprising that these models have well-known shortcomings, for example, at high solar activities. Meanwhile a large data base of close to 200,000 topside profiles from Alouette 1, 2, and ISIS 1, 2 has become available online. A program of automated scaling and inversion of a large volume of digitized ionograms adds continuously to this data pool. We have used the currently available ISIS/Alouette topside profiles to evaluate the IRI topside model and to investigate ways of improving the model. The IRI model performs generally well at middle latitudes and shows discrepancies at low and high latitudes and these discrepancies are largest during high solar activity. In the upper topside IRI consistently overestimates the measur...

Bilitza, D

2002-01-01

178

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

179

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

NASA Technical Reports Server (NTRS)

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

Niederauer, G. F.

1973-01-01

180

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

181

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

182

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

E-print Network

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

Yarsky, Peter

2005-01-01

183

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

184

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

NASA Astrophysics Data System (ADS)

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

Barbet, M.; Guillery, M.

185

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

186

Ultrahigh temperature vapor core nuclear reactor\\/MHD generator Rankine cycle space power system  

Microsoft Academic Search

Studies are being conducted on an innovative space power system that combines a uranium tetrafluoride (UF4) ultrahigh-temperature vapor core reactor (UTVR) and a disk magnetohydrodynamic (MHD) generator to obtain closed-cycle burst-mode operation (hundreds of MWe power level for a few thousand seconds). Use of UF4 as the vapor fuel and metal fluorides as the working fluid in the UTVR\\/MHD generator

Edward T. Dugan; Gerard E. Welch; Samer Kahook

1989-01-01

187

RELAP5-3D Code Application for RBMK-1500 Reactor Core Analysis  

SciTech Connect

The paper presents an evaluation of RELAP5-3D code suitability to model specific transients that take place during RBMK-1500 reactor operation, where the neutronic response of the core is important. A successful best estimate RELAP5-3D model of the Ignalina NPP RBMK-1500 reactor has been developed and validated against real plant data. Certain RELAP5-3D transient calculation results were benchmarked against calculation results obtained using the Russian code STEPAN, specially designed for RBMK reactor analysis. Comparison of the results obtained, using the RELAP5-3D and STEPAN codes, showed quite good mutual coincidence of the calculation results and good agreement with real plant data. (authors)

Bubelis, Evaldas; Kaliatka, Algirdas; Uspuras, Eugenijus [Lithuanian Energy Institute, Breslaujos str. 3, LT-44403 Kaunas (Lithuania)

2002-07-01

188

Effects of Iris Surface Curvature on Iris Recognition  

SciTech Connect

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

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

2013-01-01

189

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

SciTech Connect

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

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

2013-07-01

190

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

SciTech Connect

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

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

2013-07-01

191

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

NASA Astrophysics Data System (ADS)

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.; Ring, Peter J.; Brown, Neil; Elsner, Norbert B.; Bass, John C.

2008-01-01

192

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

193

Detection rate evaluation of ex-core detectors in the subcritical OPR-1000 reactor  

SciTech Connect

The OPR-1000 is a PWR reactor developed in Korea. One-type ex-core detectors for monitoring of power distributions were installed in the OPR-1000 reactor to alternate the three-types of the ex-core detectors. For the verification of the detection performances, neutron transport calculation was performed by using MCNP5 code. The reaction rate in the ex-core detectors and the neutron flux were evaluated by using MCNP5 code as changing the boron concentration from 1800 ppm to 1122 ppm in the subcritical condition. The reaction rate results in fission chamber show that minimum and maximum values are 0.03577 and 3.33563 reactions/cm{sup 3}-sec, respectively. This study can be directly used for the verification and improvement of fission chamber performance in using one-type ex-core detector. Also, it can be utilized for the production of the reference data in determining neutron source strength. It is expected the proposed simulation method can be utilized to the improvement of the dose monitoring system. (authors)

Won, B. H. [Dept. of Nuclear Engineering, Hanyang Univ., 17 Haendang, Seongdong, Seoul 133-791 (Korea, Republic of); Shin, C. H. [Innovative Technology Center for Radiation Safety, Hanyang Univ., 17 Haengdang, Seongdong, Seoul, 133-791 (Korea, Republic of); Kim, S. H.; Kim, H. C. [Dept. of Nuclear Engineering, Hanyang Univ., 17 Haendang, Seongdong, Seoul 133-791 (Korea, Republic of); Park, J. J. [Nuclear Safety Systems Team, Doosan Heavy Industries Co., 39-3, Seongbok, Suji, Yongin, Gyeonggi, 448-795 (Korea, Republic of); Kim, J. K. [Dept. of Nuclear Engineering, Hanyang Univ., 17 Haendang, Seongdong, Seoul 133-791 (Korea, Republic of)

2012-07-01

194

IRIS Seismic Monitor  

NSDL National Science Digital Library

The IRIS Seismic Monitor allows users to monitor global earthquakes in near real time. Researchers can locate the geology, vault conditions, site description, station instrumentation, and additional information on stations throughout the world. Visitors can learn about the latest earthquake news, including special reports of earthquakes that significantly affected human populations or had scientific significance. Students and teachers can find images and descriptions of plate tectonics as well as links to outside educational resources.

195

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

E-print Network

Very High Temperature Rector (VHTR) had been designated as one of those promising reactors for the Next Generation (IV) Nuclear Plant (NGNP). For a prismatic core VHTR, one of the most crucial design considerations is the bypass flow and crossflow...

Wang, Huhu 1985-

2012-12-13

196

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

SciTech Connect

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

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

2013-07-01

197

A high power density radial-in-flow reactor split core design for space power systems  

NASA Astrophysics Data System (ADS)

Application of the Rankine cycle to space power systems is difficult because of the problems and complexities associated with two phase flow systems in microgravity. A direct cycle system which could provide super heated vapor to the turbine inlet would greatly enhance the development of Rankine cycle power systems for space applications. The split core radial-in-flow reactor design provides a safe reliable core design for space power systems. It makes direct Rankine cycle power systems a very competitive design, eliminating the boiler and additional pumps of an indirect cycle along with the liquid vapor separator. A continuous power Rankine cycle system using this core design would produce the least weight system of any having the same power output.

Coomes, Edmund P.

198

Experimental determination of residual stress by neutron diffraction in a boiling water reactor core shroud  

SciTech Connect

Residual strains in a 51 mm (2-inch) thick 304L stainless steel plate have been measured by neutron diffraction and interpreted in terms of residual stress. The plate, measuring (300 mm) in area, was removed from a 6m (20-ft.) diameter unirradiated boiling water reactor core shroud, and included a multiple-pass horizontal weld which joined two of the cylindrical shells which comprise the core shroud. Residual stress mapping was undertaken in the heat affected zone, concentrating on the outside half of the plate thickness. Variations in residual stresses with location appeared consistent with trends expected from finite element calculations, considering that a large fraction of the residual hoop stress was released upon removal of the plate from the core shroud cylinder.

Payzant, A.; Spooner, S.; Zhu, Xiaojing; Hubbard, C.R. [and others

1996-06-01

199

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

200

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

SciTech Connect

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

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

2009-07-12

201

Thermal neutron measurements of the Rhode Island Nuclear Science Center reactor after conversion to a compact low enriched uranium core  

NASA Astrophysics Data System (ADS)

The Rhode Island Nuclear Science Center reactor, a 2 MW light water research reactor, was converted from a highly enriched uranium core to a more compact low enriched uranium core in 1993. Thermal neutron flux measurements of the instrumented beamports indicate increased thermal flux and a large improvement in the thermal to epithermal ratio. Measurements of irradiation facilities indicate an increased thermal flux, and more potential irradiation positions are now available.

Crow, M. L.; Jeng, U.; Nunes, A. C.; Malik, S. S.; Lin, D.; Bai, S.; Tehan, T.; Jacob, N.; Johnson, D. G.; Simoneau, W. A.; DiMeglio, A. F.

1995-02-01

202

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

203

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

DOEpatents

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

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

1986-01-01

204

IRI: An international Rawer initiative  

NASA Technical Reports Server (NTRS)

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.

1995-01-01

205

[Unilateral tumor of the iris].  

PubMed

We report the case of a 69-year-old asymptomatic woman who presented with the incidental diagnosis of a prominent, vascular, reddish tumor of the iris. The tumor displayed significant regression within months but reappeared years later with a similar morphology. Fluorescence angiography of the iris revealed a hypofluorescent area without intrinsic vasculature or prominent feeder vessels. In ultrasound biomicroscopy a solid, hypoechogenic iris mass was demonstrated most probably caused by varices of the iris which are a rare finding and difficult to differentiate from a cavernous hemangioma clinically. PMID:20878162

Ammermann, A; Platzeck, H; Hoerauf, H

2011-02-01

206

Instrument failure detection and reconstruction methodology in space-time nuclear reactor dynamic systems with fixed in-core detectors  

Microsoft Academic Search

The detection and isolation of instrument failures in nuclear reactors equipped with fixed in-core detectors were studied in order to improve reliability and safety. This was done by representing the reactor as a linear stochastic distributed parameter system. A bank of detection observers based on the Kalman filter concept was constructed in order to isolate component failures via robust observation.

Deog Yeon Oh; Hee Cheon No; Si Hwan Kim

1991-01-01

207

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

208

DEVELOPMENT AND MANUFACTURE OF FUEL, BLANKET, AND THERMOCOUPLE RODS FOR THE EXPERIMENTAL BREEDER REACTOR I, CORE IV  

Microsoft Academic Search

A description is given of the development and manufacture of Core IV for ; the Experimental Breeder Reactor I (EBR-I). A total of 420 rods and 10 ; thermocouple fuel rods containing plutonium-1.25 wt% aluminum fuel slugs were ; made. In addition, 120 blanket rods and 5 thermocouple A description is given of ; the developrnent and manufacture of Core

W. R. Jr. Burt; A. G. Hins; R. M. Mayfield; A. B. Shuck

1963-01-01

209

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

210

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

211

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

NASA Astrophysics Data System (ADS)

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

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

1999-08-01

212

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

213

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

NASA Astrophysics Data System (ADS)

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

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

2001-02-01

214

Comparison of Irradiation Conditions of VVER-1000 Reactor Pressure Vessel and Surveillance Specimens for Various Core Loadings  

NASA Astrophysics Data System (ADS)

The comparative analysis of irradiation conditions of surveillance specimens and pressure vessel of VVER-1000 reactor has been carried out for various configurations of the core. It is proved the fluences onto specimens and a pressure vessel don't correlate with each other but only the spectral indexes do. It is revealed that in the case of the specimen reconstitution technique application the data on the assembly orientation to the reactor core is sufficient to complete four representative groups from the samples of any container assembly. It is shown that the standard surveillance program of VVER-1000 allows obtaining reliable information on the reactor pressure vessel state.

Bukanov, V. N.; Diemokhin, V. L.; Grytsenko, O. V.; Vasylieva, O. G.; Pugach, S. M.

2009-08-01

215

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

216

A survey of alternative once-through fast reactor core designs  

SciTech Connect

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

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

2012-07-01

217

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

SciTech Connect

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

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

2013-07-01

218

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

SciTech Connect

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

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

2013-07-01

219

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

NASA Astrophysics Data System (ADS)

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

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

2014-06-01

220

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

221

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

SciTech Connect

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

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

1981-06-01

222

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

NASA Technical Reports Server (NTRS)

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

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

1981-01-01

223

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

224

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

225

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

226

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

SciTech Connect

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

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

1990-12-01

227

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

NASA Technical Reports Server (NTRS)

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

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

2007-01-01

228

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

NASA Technical Reports Server (NTRS)

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

Roman, W. C.

1979-01-01

229

International Referecne Ionosphere (IRI) - 2006  

NASA Astrophysics Data System (ADS)

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

Bilitza, D.

230

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

NASA Technical Reports Server (NTRS)

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

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

1976-01-01

231

Acute endothelial failure after cosmetic iris implants (NewIris®).  

PubMed

We report a case of an acute endothelial failure after the implantation of a new cosmetic, colored, artificial iris diaphragm implant called NewIris(®). A 21-year-old woman came to us complaining of progressive loss of vision and pain after NewIris lenses had been implanted. Decreased visual acuity, corneal edema, and increased intraocular pressure in both eyes appeared only 3 weeks after the surgery. The lenses were removed as soon as possible but had already severely affected the endothelial cell count. NewIris implants are an alternative to cosmetic contact lenses, but they are not as safe as other phakic anterior chamber intraocular lenses, nor are they a good option for the patient. PMID:21691579

Garcia-Pous, Maria; Udaondo, Patricia; Garcia-Delpech, Salvador; Salom, David; Díaz-Llopis, Manuel

2011-01-01

232

Acute endothelial failure after cosmetic iris implants (NewIris®)  

PubMed Central

We report a case of an acute endothelial failure after the implantation of a new cosmetic, colored, artificial iris diaphragm implant called NewIris®. A 21-year-old woman came to us complaining of progressive loss of vision and pain after NewIris lenses had been implanted. Decreased visual acuity, corneal edema, and increased intraocular pressure in both eyes appeared only 3 weeks after the surgery. The lenses were removed as soon as possible but had already severely affected the endothelial cell count. NewIris implants are an alternative to cosmetic contact lenses, but they are not as safe as other phakic anterior chamber intraocular lenses, nor are they a good option for the patient. PMID:21691579

Garcia-Pous, Maria; Udaondo, Patricia; Garcia-Delpech, Salvador; Salom, David; Díaz-Llopis, Manuel

2011-01-01

233

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

234

Shape Adaptive, Robust Iris Feature Extraction from Noisy Iris Images  

PubMed Central

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

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

2013-01-01

235

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

NASA Technical Reports Server (NTRS)

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

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

1973-01-01

236

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

E-print Network

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

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

2005-09-15

237

Developing Fully Coupled Dynamical Reactor Core Isolation System Models in RELAP-7 for Extended Station Black-Out Analysis  

SciTech Connect

The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) provides makeup water to the reactor vessel for core cooling when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. It was one of the very few safety systems still available during the Fukushima Daiichi accidents after the tsunamis hit the plants and the system successfully delayed the core meltdown for a few days for unit 2 & 3. Therefore, detailed models for RCIC system components are indispensable to understand extended station black-out accidents (SBO) for BWRs. As part of the effort to develop the new generation reactor system safety analysis code RELAP-7, major components to simulate the RCIC system have been developed. This paper describes the models for those components such as turbine, pump, and wet well. Selected individual component test simulations and a simplified SBO simulation up to but before core damage is presented. The successful implementation of the simplified RCIC and wet well models paves the way to further improve the models for safety analysis by including more detailed physical processes in the near future.

Haihua Zhao; Ling Zou; Hongbin Zhang; David Andrs; Richard Martineau

2014-04-01

238

INTEGRATED RISK INFORMATION SYSTEM (IRIS)  

EPA Science Inventory

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

239

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

240

Gas Core Reactor Numerical Simulation Using a Coupled MHD-MCNP Model  

NASA Technical Reports Server (NTRS)

Analysis is provided in this report of using two head-on magnetohydrodynamic (MHD) shocks to achieve supercritical nuclear fission in an axially elongated cylinder filled with UF4 gas as an energy source for deep space missions. The motivation for each aspect of the design is explained and supported by theory and numerical simulations. A subsequent report will provide detail on relevant experimental work to validate the concept. Here the focus is on the theory of and simulations for the proposed gas core reactor conceptual design from the onset of shock generations to the supercritical state achieved when the shocks collide. The MHD model is coupled to a standard nuclear code (MCNP) to observe the neutron flux and fission power attributed to the supercritical state brought about by the shock collisions. Throughout the modeling, realistic parameters are used for the initial ambient gaseous state and currents to ensure a resulting supercritical state upon shock collisions.

Kazeminezhad, F.; Anghaie, S.

2008-01-01

241

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

NASA Technical Reports Server (NTRS)

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

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

1973-01-01

242

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

SciTech Connect

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

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

2012-07-01

243

Dynamic loads from reactor pressure vessel core melt-through under high primary systems pressure  

SciTech Connect

Estimates are presented of the thermal-hydraulic load acting on a pressurized water reactor pressure vessel and its support girder after lower head failure at high pressure (227 MPa). The estimates are based on one-dimensional calculations performed with the RELAP5/MOD3 transient analysis thermal-hydraulics code. The information obtained provides a force-function input for structural dynamic calculations of an increased containment. On the assumption of a global circumferential rupture of the vessel lower head, the computations show a load peak of 340 MN and a continuing load of 160 MN acting on the vessel support ring. The analysis is related to the containment concept of Eibl, Kessler, and Hennies, which is aimed at developing passive mechanisms that can safely confine core-melt consequences.

Jacobs, G. [Forschungszentrum Karlsruhe (Germany)

1995-09-01

244

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

SciTech Connect

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

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

2012-07-01

245

Non-Invasive Imaging of Reactor Cores Using Cosmic Ray Muons  

NASA Astrophysics Data System (ADS)

Cosmic ray muons penetrate deeply in material, with some passing completely through very thick objects. This penetrating quality is the basis of two distinct, but related imaging techniques. The first measures the number of cosmic ray muons transmitted through parts of an object. Relatively fewer muons are absorbed along paths in which they encounter less material, compared to higher density paths, so the relative density of material is measured. This technique is called muon transmission imaging, and has been used to infer the density and structure of a variety of large masses, including mine overburden, volcanoes, pyramids, and buildings. In a second, more recently developed technique, the angular deflection of muons is measured by trajectory-tracking detectors placed on two opposing sides of an object. Muons are deflected more strongly by heavy nuclei, since multiple Coulomb scattering angle is approximately proportional to the nuclear charge. Therefore, a map showing regions of large deflection will identify the location of uranium in contrast to lighter nuclei. This technique is termed muon scattering tomography (MST) and has been developed to screen shipping containers for the presence of concealed nuclear material. Both techniques are a good way of non-invasively inspecting objects. A previously unexplored topic was applying MST to imaging large objects. Here we demonstrate extending the MST technique to the task of identifying relatively thick objects inside very thick shielding. We measured cosmic ray muons passing through a physical arrangement of material similar to a nuclear reactor, with thick concrete shielding and a heavy metal core. Newly developed algorithms were used to reconstruct an image of the ``mock reactor core,'' with resolution of approximately 30 cm.

Milner, Edward

2011-10-01

246

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

247

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

248

Operating experience of natural circulation core cooling in boiling water reactors  

SciTech Connect

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 stability during startup, shutdown, and at power conditions. As a consequence, the NRC has directed personnel at several national labs to help investigate SBWR stability issues. This paper will focus on some of the preliminary findings made at the INEL. Because of the broad range of stability issues this paper will mainly focus on potential geysering instabilities during startup. The two NC designs examined in detail are the US Humboldt Bay Unit 3 BWR-1 plant and Dodewaard plant in the Netherlands. The objective of this paper will be to review operating experience of these two plants and evaluate their relevance to planned SBWR operational procedures. For completeness, experimental work with early natural circulation GE test facilities will also be briefly discussed.

Kullberg, C.; Jones, K.; Heath, C.

1993-08-01

249

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

250

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

E-print Network

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

Laughlin, Robert B.

251

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

SciTech Connect

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

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

2007-07-01

252

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/ ?/cm 2-s/cm), -5.2×10 -18 A/R/h/cm (-1.133×10 -23 A/ ?/cm 2-s/cm) and 34×10 -18 A/R/h/cm (7.14×10 -23 A/ ?/cm 2-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

253

Reactivity analysis model based on finite difference method for three-dimensional fast breeder reactor core deformation  

Microsoft Academic Search

An analysis model has been proposed to evaluate reactivity due to horizontal fast breeder reactor (FBR) core deformation in seismic events by direct three-dimensional eigenvalue calculations, which are impossible for current neutronic analysis programs. The model is based on a current-centered finite difference neutron diffusion calculation method. Macroscopic neutron reaction cross sections are defined, which take into account changes in

Azekura

1987-01-01

254

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

255

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

Microsoft Academic Search

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

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

2010-01-01

256

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

E-print Network

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

Helvenston, Edward M. (Edward March)

2006-01-01

257

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

NASA Astrophysics Data System (ADS)

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

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

2010-06-01

258

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

259

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

NASA Technical Reports Server (NTRS)

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

Anghaie, S.; Chen, G.

1996-01-01

260

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

261

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

2014-03-01

262

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

NASA Astrophysics Data System (ADS)

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

Espel, Federico Puente

263

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

264

RECENT DEVELOPMENTS OF THE IRIS PROJECT OF INTEREST FOR LATIN AMERICA  

Microsoft Academic Search

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

M. D. Carelli; B. Petrovic

2004-01-01

265

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

266

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

SciTech Connect

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

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

1990-11-01

267

Iris stromal imbrication oversewing for pigment epithelial defects.  

PubMed

We present a novel iris repair technique for the management of iris transillumination defects secondary to iris pigment epithelium (IPE) loss, which includes iris oversewing over the defect through partial iris stromal bites with 10-0 polypropylene. This technique provides a healthy layer of iris covering the transillumination defect without the creation of new defects on the contiguous IPE. PMID:24814963

Snyder, Michael E; Perez, Mauricio A

2015-01-01

268

Static and dynamic neutronic analysis of a burst-mode multiple-cavity gas core reactor, Rankine cycle space power system  

Microsoft Academic Search

Static and dynamic neutronic analyses have been performed on an innovative burst-mode Ultrahigh-Temperature Vapor Core Reactor (UTVR) space nuclear power system. This novel reactor concept employs multiple neutronically coupled fissioning cores and operates on a direct closed Rankine cycle using a disk magnetohydrodynamic generator for energy conversion. The UTVR includes two types of fissioning core regions: (a) the central Ultrahigh-Temperature

E. T. Dugan; S. D. Kahook

1993-01-01

269

IRIS Toxicological Review of Hexachloroethane (2011 Final)  

EPA Science Inventory

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

270

Exploring New Directions in Iris Recognition  

E-print Network

A new approach in iris recognition based on Circular Fuzzy Iris Segmentation (CFIS) and Gabor Analytic Iris Texture Binary Encoder (GAITBE) is proposed and tested here. CFIS procedure is designed to guarantee that similar iris segments will be obtained for similar eye images, despite the fact that the degree of occlusion may vary from one image to another. Its result is a circular iris ring (concentric with the pupil) which approximates the actual iris. GAITBE proves better encoding of statistical independence between the iris codes extracted from different irides using Hilbert Transform. Irides from University of Bath Iris Database are binary encoded on two different lengths (768 / 192 bytes) and tested in both single-enrollment and multi-enrollment identification scenarios. All cases illustrate the capacity of the newly proposed methodology to narrow down the distribution of inter-class matching scores, and consequently, to guarantee a steeper descent of the False Accept Rate.

Popescu-Bodorin, Nicolaie

2011-01-01

271

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

NASA Technical Reports Server (NTRS)

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

Chow, S.

1976-01-01

272

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

273

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

PubMed

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

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

2011-11-01

274

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

275

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

276

Size Distribution of NaK Droplets Released During Rorsat Reactor Core Ejection  

NASA Astrophysics Data System (ADS)

NaK droplets consist of eutectic sodium-potassium alloy and have been released during RORSAT reactor core ejections mostly on orbits close to 950 km altitude. They contributed to the space debris environment in the centimeter and millimeter size regime. NaK droplets have been modeled before in ESA's MASTER Debris and Meteoroid Environment Model. The approach is currently revised for the MASTER 2005 upgrade. The new NaK model gives estimations of the parameters of the size distribution function, which are based on physical relations only. This physical approach confirms NASA radar observations. The core ejection causes an opening of the primary coolant circuit. The liquid coolant is released into space forming droplets up to a diameter of 5.5 cm. The reactor design is investigated to understand the possible mechanisms that cause the droplets generation. It is likely that the droplet generation process can be both capillary jet breakup and atomization. This paper presents results of the estimation of droplets sizes. A droplet size distribution is introduced, which is scientifically justified. The physical process of atomization resp. liquid jet breakup is considered, to derive the parameters of the size distribution function. The introduction of an improved distribution function is important. So far the cumulative size distribution function was a combination of several fitting curve segments to agree with measured data. The definition of several functions results in a large number of parameters. This drawback is corrected. The droplet size can be defined as function of the orifice diameter. The droplets sizes are related to the parameters of the size distribution function. The size distribution function shall contain only two parameters, which can be derived from the orifice diameter and the atomization conditions. In this way scientifically based estimations of the parameters are introduced. An estimation of the maximum droplet diameter assuming capillary jet breakup is presented. The minimum droplet diameter is estimated to be one order of magnitude smaller as the orifice diameter, assuming effervescent atomization. The resulting size distribution agrees with measurement data. A bimodal size distribution is derived, which is based on the Rosin-Rammler equation. The Rosin-Rammler equation is an empirical volume distribution function. The number of parameters is limited to two. It is likely that the coolant system contains two types of orifice diameters. This makes it necessary to apply the Rosin-Rammler distribution twice, resulting in a bimodal size distribution with altogether four parameters. The comparison shows that the new NaK model is in good agreement with the NASA model, which is based on radar observations. Results of orbit propagation simulation runs are presented in terms of spatial density.

Wiedemann, C.; Oswald, M.; Stabroth, S.; Klinkrad, H.; Vörsmann, P.

277

TE10 resonant iris with angular alignment  

E-print Network

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

Bornemann, Jens

278

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

279

Iris: The VAO SED Application  

NASA Astrophysics Data System (ADS)

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

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

2012-09-01

280

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

281

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

SciTech Connect

Current work modeling high temperature compact heat exchangers may demonstrate the design feasibility of a Vapor Core Reactor (VCR) driven combined cycle power plant. For solid nuclear fuel designs, the cycle efficiency is typically limited by a metallurgical temperature limit which is dictated by fuel and structural melting points. In a vapor core, the gas/vapor phase nuclear fuel is uniformly mixed with the topping cycle working fluid. Heat is generated homogeneously throughout the working fluid thus extending the metallurgical temperature limit. Because of the high temperature, magnetohydrodynamic (MHD) generation is employed for topping cycle power extraction. MHD rejected heat is transported via compact heat exchanger to a conventional Brayton gas turbine bottoming cycle. High bottoming cycle mass flow rates are required to remove the waste heat because of low heat capacities for the bottoming cycle gas. High mass flow is also necessary to balance the high Uranium Tetrafluoride (UF{sub 4}) mass flow rate in the topping cycle. Heat exchanger design is critical due to the high temperatures and corrosive influence of fluoride compounds and fission products existing in VCR/MHD exhaust. Working fluid compositions for the topping cycle include variations of Uranium Tetrafluoride, Helium and various electrical conductivity seeds for the MHD. Bottoming cycle working fluid compositions include variations of Helium and Xenon. Some thought has been given to include liquid metal vapor in the bottoming cycle for a Cheng or evaporative cooled design enhancement. The NASA Glenn Lewis Research Center code Chemical Equilibrium with Applications (CEA) is utilized for evaluating chemical species existing in the gas stream. Work being conducted demonstrates the compact heat exchanger design, utilization of the CEA code, and assessment of different topping and bottoming working fluid compositions. (authors)

Bays, Samuel E.; Anghaie, Samim; Smith, Blair; Knight, Travis [Innovative Space Power and Propulsion Institute, University of Florida, 202 Nuclear Science Building, Gainesville, FL 32611 (United States)

2004-07-01

282

Size distribution of NaK droplets released during RORSAT reactor core ejection  

NASA Astrophysics Data System (ADS)

NaK droplets consist of eutectic sodium-potassium alloy and have been released during RORSAT reactor core ejections mostly on orbits close to 950 km altitude. They contributed to the space debris environment in the centimeter and millimeter regime. NaK droplets have been modeled before in ESAs MASTER Debris and Meteoroid Environment Model. The approach is currently revised for the MASTER 2005 upgrade. The new NaK model gives estimations of the parameters of the size distribution function, which are based on physical relations only. NASA radar observations confirm this physical approach. A bimodal size distribution is derived, which is based on the Rosin-Rammler equation. The Rosin-Rammler equation is an empirical volume distribution function. The number of parameters is limited to two. It is likely that the coolant system contains two types of orifice diameters. This makes it necessary to apply the Rosin-Rammler distribution twice, resulting in a bimodal size distribution with altogether four parameters. The comparison shows that the new NaK model is in good agreement with the NASA model, which is based on radar observations. Results of orbit propagation simulation runs are presented.

Wiedemann, C.; Oswald, M.; Stabroth, S.; Klinkrad, H.; Vörsmann, P.

283

FLICA-4 (version 1) a computer code for three dimensional thermal analysis of nuclear reactor cores  

SciTech Connect

FLICA-4 is a thermal-hydraulic computer code developed at the French Energy Atomic Commission (CEA) for three dimensional steady state or transient two phase flow for design and safety thermal analysis of nuclear reactor cores. The two phase flow model of FLICA-4 is based on four balance equations for the fluid which includes: three balance equations for the mixture and a mass balance equation for the less concentrated phase which permits the calculation of non-equilibrium flows as sub cooled boiling and superheated steam. A drift velocity model takes into account the velocity disequilibrium between phases. The thermal behaviour of fuel elements can be computed by a one dimensional heat conduction equation in plane, cylindrical or spherical geometries and coupled to the fluid flow calculation. Convection and diffusion of solution products which are transported either by the liquid or by the gas, can be evaluated by solving specific mass conservation equations. A one dimensional two phase flow model can also be used to compute 1-D flow in pipes, guide tubes, BWR assemblies or RBMK channels. The FLICA-4 computer code uses fast running time steam-water functions. Phasic and saturation physical properties are computed by using bi-cubic spline functions. Polynomial coefficients are tabulated from 0.1 to 22 MPa and 0 to 800{degrees}C. Specific modules can be utilised in order to generate the spline coefficients for any other fluid properties.

Raymond, P.; Allaire, G.; Boudsocq, G. [Commissariat a l`Energie Atomique, Gif sur Yvette (France)] [and others

1995-12-31

284

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

SciTech Connect

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

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

2013-07-01

285

A Fully Implicit, Second Order in Time, Simulation of a Nuclear Reactor Core  

SciTech Connect

This paper will present a high fidelity solution algorithm for a model of a nuclear reactor core barrel. This model consists of a system of nine nonlinearly coupled partial differential equations. The coolant is modeled with the 1-D six-equation two-phase flow model of RELAP5. Nonlinear heat conduction is modeled with a single 2-D equation. The fission power comes from two 2-D equations for neutron diffusion and precursor concentration. The solution algorithm presented will be the physics-based preconditioned Jacobian-free Newton-Krylov (JFNK) method. In this approach all nine equations are discretized and then solved in a single nonlinear system. Newton's method is used to iterate the nonlinear system to convergence. The Krylov linear solution method is used to solve the matrices in the linear steps of the Newton iterations. The physics-based pre-conditioner provides an approximation to the solution of the linear system that accelerates the Krylov iterations. Results will be presented for two algorithms. The first algorithm will be the traditional approach used by RELAP5. Here the two-phase flow equations are solved separately from the nonlinear conduction and neutron diffusion. Because of this splitting of the physics, and the linearizations employed this method is first order accurate in time. A second algorithm will be the JFNK method solved second order in time accurate. Results will be presented which compare these two algorithms in terms of accuracy and efficiency. (author)

Mousseau, Vincent A. [Los Alamos National Laboratory, P.O. Box 1663 Los Alamos, NM 87545 (United States)

2006-07-01

286

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

287

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

288

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 ...

289

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

E-print Network

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

Hill, Natasha J.; Parks, Geoffrey T.

2014-08-24

290

Modeling And Analysis Of The Diskund Generator Component Of A Gas Core Reactor\\/MHD Rankine Cycle Space Power System  

Microsoft Academic Search

A gas core nuclear reactor (GCR)\\/disk magnetohydrodynamic (MHD) generator direct closed Rankine space power system concept is described. The GCR\\/disk MHD generator marriage facilitates efficient high electric power density system performance at relatively high operating temperatures. The aystem concept promises high specific power levels, on the order of 1 kWe\\/kg. An overview of the disk MHD generator component magnetofluiddynamic and

Gerard E. Welch; Edward T. Dugan; W. E. Lear; J. G. Appelbaum

1990-01-01

291

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

Microsoft Academic Search

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

Hidemasa Yamano; Yoshiharu Tobita; Satoshi Fujita; Werner Maschek

2009-01-01

292

The severe accident mitigation concept and the design measures for core melt retention of the European Pressurized Reactor (EPR)  

Microsoft Academic Search

For the mitigation of severe accidents, the European Pressurized Water Reactor (EPR) has adopted and improved the defense-in-depth approaches of its predecessors, the French “N4” and the German “Konvoi” plants. Beyond the corresponding evolutionary changes, the EPR includes a new, 4th level of defense-in-depth that is aimed at limiting the consequences of a postulated severe accident with core melting. It

Manfred Fischer

2004-01-01

293

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 initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has been evaluated as an acceptable benchmark experiment. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. 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

2013-03-01

294

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

295

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

NASA Astrophysics Data System (ADS)

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

Mosher, Scott W.

296

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

SciTech Connect

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

Michael A. Pope

2011-10-01

297

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

298

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

299

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

E-print Network

removal During an accident, the reactivity feedback of the coreCore coolant temperatures stabilize at an equilibrium value as decay heat removalcore. The engineering, design and operation of systems for heat removal

Qvist, Staffan Alexander

2013-01-01

300

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

301

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

302

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

303

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

304

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

305

IRIS thermal balance test within ESTEC LSS  

NASA Technical Reports Server (NTRS)

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

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

1988-01-01

306

Proton beam therapy for iris melanomas  

Microsoft Academic Search

AimsTo describe the results in terms of local control, eye preservation and systemic evolution of iris melanomas treated by proton beam irradiation.MethodsRetrospective review of the charts of patients with iris melanoma treated by proton beam therapy between April 1998 and September 2002. Ciliary body melanomas with iris involvement or tumours with extrascleral invasion were excluded. Treatment consisted of 60 Gy

L Lumbroso-Le Rouic; S Delacroix; R Dendale; C Levy-Gabriel; L Feuvret; G Noel; C Plancher; C Nauraye; P Garcia; V Calugaru; B Asselain; L Desjardins

2006-01-01

307

Enhancement of Transmutation Characteristics of the Minor Actinide Burning Fast Reactor Core Concept Using Hydride Fuel Targets and Its Introduction Scenario  

Microsoft Academic Search

Transmutation characteristics of the minor actinide (MA) burning fast reactor core using hydride fuel targets are enhanced to reduce long-term radiotoxicity of nuclear waste. A scenario which introduces the concept is investigated. (1) The MA burner core with plutonium (Pu) multi-recycling can transmute a large amount of MAs; the amount is about that produced in 21 LWRs per year. The

Koji FUJIMURA; Toshio SANDA; Michio YAMAWAKI; Kenji KONASHI

2001-01-01

308

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

PubMed

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

Gugiu, Daniela; Dumitrache, Ion

2005-01-01

309

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

310

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

SciTech Connect

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

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

2014-01-15

311

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, Raphaële; Sharma, Govind Kumar; Jayakumar, Tammana; Benoist, Philippe; Raj, Baldev

2010-02-01

312

Measurements of natural circulation flow in a scale model PWR reactor system during postulated degraded core accidents using laser anemometry  

NASA Technical Reports Server (NTRS)

The natural circulation of a single-phase fluid in a scale-model pressurized water reactor system was studied during a postulated degraded core accident. A half section of a one-seventh scale model with a plexiglass adiabatic window was employed. Water and SF6 were used as the fluid. LDA was used to perform velocity measurements along the center plane of the model at five elevations. It was found that the recirculation flow patterns are nearly symmetric except near the hot legs and in the upper head and that the fluid in the upper plenum is well mixed.

Kadambi, J. R.; Schneider, S. J.; Stewart, W. A.

1987-01-01

313

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

314

Bounding core temperature transients for severe and rapid water ingress scenarios in modular high temperature gas-cooled reactors  

SciTech Connect

A rapid water ingress transient, resulting from steam generator tube or tube-sheet failures, could lead to a reactivity insertion and core heatup in the Modular High Temperature Gas-Cooled Reactors. This paper considers the effect of hypothetical rapid and severe water ingress scenarios of extremely low probability, and assesses the effect of such transients on potentially excessive fuel temperatures and subsequent fuel failures. The results indicate that for the worst postulated scenarios the conservatively set limiting fuel temperature of 1600{degree}C is indeed exceeded, but only for a few seconds, and then only in a small fraction of the core. Therefore, it appears that even the most severe and rapid water ingress transients would not lead to significant fuel failures. Parametric variations of the key variables indicate that the reactivity worth of water and the fuel thermal properties must be established with high confidence as the design progresses. 7 refs., 11 figs., 3 tabs.

Kroeger, P.G.

1990-01-01

315

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

E-print Network

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

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

2007-01-01

316

Secondary Ion Mass Spectrometry Analysis of Materials to Develop In-core Safeguards Reactor Monitoring Devices  

SciTech Connect

During reactor operations and fuel burn up, some isotopic abundances change due to nuclear reactions and provide sensitive indicators of neutron fluence and fuel burnup. Secondary ion mass spectrometry (SIMS) analysis has been used to directly measure isotope ratios of selected impurity elements in irradiated nuclear reactor materials. Direct in situ SIMS measurements were made in graphite and metal samples, following shaping and surface cleaning. Other elements such as Be must be chemically separated and purified prior to SIMS analyses. Elements such as pre-existing impurity U and Pu produced from the U, are in low abundance and must also be chemically separated and are measured by thermal ionization mass spectrometry (TIMS). Studies combining SIMS and TIMS analyses demonstrate the value of this approach in determining reactor fluence profiles, power production, and other parameters. Future work proceeding from this analytical work will include developing monitoring devices designed for relatively easy placement and retrieval in a reactor, and direct SIMS analyses after exposure.

Gerlach, David C.; Reid, Bruce D.; Gesh, Christopher J.; Mitchell, Mark R.; Szechenyi, Scott C.; Douglas, Matthew; McNamara, Bruce K.; Ellis, Tere A.; Ermi, Ruby M.

2010-08-11

317

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

SciTech Connect

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

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

1993-12-31

318

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

SciTech Connect

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

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

2006-11-01

319

Specialised sympathetic neuroeffector associations in immature rat iris arterioles  

PubMed Central

Sympathetic nerve-mediated vasoconstriction in iris arterioles of mature rats occurs via the activation of ?1B-adrenoceptors alone, while in immature rat iris arterioles, vasoconstriction occurs via activation of both ?1- and ?2-adrenoceptors. In mature rats the vast majority of sympathetic varicosities form close neuroeffector junctions. Serial section electron microscopy of 14 d iris arterioles has been used to determine whether restriction in physiological receptor types with age may result from the establishment of these close neuroeffector junctions. Ninety varicosities which lay within 4 ?m of arteriolar smooth muscle were followed for their entire length. Varicosities rarely contained dense cored vesicles even after treatment with 5-hydroxydopamine. 47% of varicosities formed close associations with muscle cells and 88% formed close associations with muscle cells or melanocytes. Varicosities in bundles were as likely as single varicosities to form close associations with vascular smooth muscle cells, although the distribution of synaptic vesicles in single varicosities did not show the asymmetric accumulation towards the smooth muscle cells seen in the varicosities in bundles which were frequently clustered together. We conclude that restriction of physiological receptor types during development does not appear to correlate with the establishment of close neuroeffector junctions, although changes in presynaptic structures may contribute to the refinement of postsynaptic responses. PMID:10529061

SANDOW, SHAUN L.; HILL, CARYL E.

1999-01-01

320

Daugman's Iris Scanning Algorithm W. A. Barrett  

E-print Network

engineering students have approached this problem, but much work remains. Outline of Daugman's Algorithm Daugman [1, 2, 3] concerning the recognition of persons through their highly distinctive iris patterns in this is to develop our own approach to iris scanning for academic research purposes. Certain of our undergraduate

Barrett, William A.

321

Iris Color and Macular Pigment Optical Density  

Microsoft Academic Search

The present study was designed to assess the relationship between iris color and macular pigment optical density. Both melanin and carotenoids (responsible for iris color and macular pigment composition, respectively) appear to protect the retina through similar mechanisms and higher concentrations may reduce the incidence of retinal degenerations. To evaluate this relationship, 95 subjects were examined and the following variables

BILLY R. HAMMOND JR.; KENNETH FULD; MAX D. SNODDERLY

1996-01-01

322

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

323

Iris atrophy with hypoperfusion and microneovascularisation.  

PubMed Central

A series of 17 patients with stromal atrophy, hypoperfusion, and microneovascularisation of the iris investigated in the Glaucoma Investigation and Research Unit are described, and their iris angiograms were compared with those of normal irides of patients in the same age group seen in general clinics. In all but one of the 17 cases this iris atrophy was associated with glaucoma or ocular hypertension, which appeared to be secondary to the iris changes. The condition was bilateral and presented a typical slit-lamp appearance, with subtle evidence of microneovascularisation. There was neither history nor clinical evidence of previous trauma, heterochromia, or intraocular inflammation. The commonest form of iris atrophy affected the inner third of the iris stroma in a patchy manner, often with sparing above. However, diffuse atrophy occurred in two cases, and there were two cases of 'senile tears' of the iris. Some accompanying atrophy of the pigment epithelium was usual but less prominent. The changes on fluorescein angiography of the iris included the late appearance of dye with a long arteriovenous circulation time, fewer arteries than normal with sectorial hypoperfusion, leakage of dye from the pupil margin and peripupillary neovascularisation, stromal tufts, and sometimes more complex stromal microneovascularisation. An expanded prominent lesser vascular circle was a common feature of the condition. The condition is bilateral and distinct from other forms of iris atrophy. In all cases the iris changes appeared to be secondary to the vascular hypoperfusion and were not consistently associated with evidence of gross vascular disease. All patients had grey (blue) irides, and this may be an aetiological factor. The condition appears common enough to form a significant group of glaucoma patients and to be a separate clinical entity. Images PMID:2444247

Brooks, A. M.; Gillies, W. E.

1987-01-01

324

Design analysis of the molten core confinement within the reactor vessel in the case of severe accidents at nuclear power plants equipped with a reactor of the VVER type  

NASA Astrophysics Data System (ADS)

The present paper reports the results of the preliminary design estimate of the behavior of the core melt in vessels of reactors of the VVER-600 and VVER-1300 types (a standard optimized and informative nuclear power unit based on VVER technology—VVER TOI) in the case of beyond-design-basis severe accidents. The basic processes determining the state of the core melt in the reactor vessel are analyzed. The concept of molten core confinement within the vessel based on the idea of outside cooling is discussed. Basic assumptions and models, as well as the results of calculation of the interaction between molten materials of the core and the wall of the reactor vessel performed by means of the SOCRAT severe accident code, are presented and discussed. On the basis of the data obtained, the requirements on the operation of the safety systems are determined, upon the fulfillment of which there will appear potential prerequisites for implementing the concept of the confinement of the core melt within the reactor in cases of severe accidents at nuclear power plants equipped with VVER reactors.

Zvonaryov, Yu. A.; Budaev, M. A.; Volchek, A. M.; Gorbaev, V. A.; Zagryazkin, V. N.; Kiselyov, N. P.; Kobzar', V. L.; Konobeev, A. V.; Tsurikov, D. F.

2013-12-01

325

Reactor Physics Methods and Preconceptual Core Design Analyses for Conversion of the Advanced Test Reactor to Low-Enriched Uranium Fuel Annual Report for Fiscal Year 2012  

SciTech Connect

Under the current long-term DOE policy and planning scenario, both the ATR and the ATRC will be reconfigured at an appropriate time within the next several years to operate with low-enriched uranium (LEU) fuel. This will be accomplished under the auspices of the Reduced Enrichment Research and Test Reactor (RERTR) Program, administered by the DOE National Nuclear Security Administration (NNSA). At a minimum, the internal design and composition of the fuel element plates and support structure will change, to accommodate the need for low enrichment in a manner that maintains total core excess reactivity at a suitable level for anticipated operational needs throughout each cycle while respecting all control and shutdown margin requirements and power distribution limits. The complete engineering design and optimization of LEU cores for the ATR and the ATRC will require significant multi-year efforts in the areas of fuel design, development and testing, as well as a complete re-analysis of the relevant reactor physics parameters for a core composed of LEU fuel, with possible control system modifications. Ultimately, revalidation of the computational physics parameters per applicable national and international standards against data from experimental measurements for prototypes of the new ATR and ATRC core designs will also be required for Safety Analysis Report (SAR) changes to support routine operations with LEU. This report is focused on reactor physics analyses conducted during Fiscal Year (FY) 2012 to support the initial development of several potential preconceptual fuel element designs that are suitable candidates for further study and refinement during FY-2013 and beyond. In a separate, but related, effort in the general area of computational support for ATR operations, the Idaho National Laboratory (INL) is conducting a focused multiyear effort to introduce modern high-fidelity computational reactor physics software and associated validation protocols to replace several obsolete components of the current analytical tool set used for ATR neutronics support. 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). It will also greatly facilitate the LEU conversion effort, since the upgraded computational capabilities are now at a stage where they can be, and in fact have been, used for the required physics analysis from the beginning. In this context, extensive scoping neutronics analyses were completed for six preconceptual candidate LEU fuel element designs for the ATR (and for its companion critical facility, ATRC). Of these, four exhibited neutronics performance in what is believed to be an acceptable range. However, there are currently some concerns with regard to fabricability and mechanical performance that have emerged for one of the four latter concepts. Thus three concepts have been selected for more comprehensive conceptual design analysis during the upcoming fiscal year.

David W. Nigg; Sean R. Morrell

2012-09-01

326

New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions.  

PubMed

Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 × 10(20) n?cm(2). A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement. PMID:21456734

Geslot, B; Vermeeren, L; Filliatre, P; Lopez, A Legrand; Barbot, L; Jammes, C; Bréaud, S; Oriol, L; Villard, J-F

2011-03-01

327

New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions  

NASA Astrophysics Data System (ADS)

Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 × 1020 n/cm2. A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

Geslot, B.; Vermeeren, L.; Filliatre, P.; Lopez, A. Legrand; Barbot, L.; Jammes, C.; Bréaud, S.; Oriol, L.; Villard, J.-F.

2011-03-01

328

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

329

IRI, an International Standard for the Ionosphere  

NASA Astrophysics Data System (ADS)

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

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

2003-04-01

330

Design & development of soft-core processor based remote terminal units for nuclear reactors  

Microsoft Academic Search

Remote Terminal Units (RTUs) are single board, real time remote data acquisition & control systems that are used in Fast Breeder Reactors to acquire analog\\/digital signals [like voltage, signal inputs from surface thermocouple, leak detector & limit switches], sends digitized data packets over Ethernet to the nearest Local Control Centre (LCC) and generate control outputs in the form of potential

Aditya Gour; A. Santhana Raj; R. P. Behera; N. Murali; S. A. V Satya Murty

2011-01-01

331

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

Microsoft Academic Search

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

1976-01-01

332

The Long-Life Gas Turbine Fast Reactor Matrix Core Concept  

Microsoft Academic Search

A fast reactor version of the modular HTGR-GT is proposed which has good potential to satisfy the Gen IV goals of competitive economy, enhanced nuclear safety, plus reduced proliferation risk and nuclear waste. Good economy is pursued through a modular design of 300 MWe rating, a direct cycle and a simple balance of plant using a supercritical CO Brayton cycle.

P. Hejzlar; M. J. Driscoll; N. E. Todreas

2002-01-01

333

Startup of “Candle” burnup in fast reactor from enriched uranium core  

Microsoft Academic Search

A new reactor burnup strategy CANDLE was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move to an axial direction. Here important points are that the solid fuel is fixed at each position and that any movable burnup reactivity control mechanisms such as control rods are not required.This burnup strategy can derive many

Hiroshi Sekimoto; Seiichi Miyashita

2006-01-01

334

Creating geometry and mesh models for nuclear reactor core geometries using a lattice hierarchy-based approach.  

SciTech Connect

Nuclear reactor cores are constructed as rectangular or hexagonal lattices of assemblies, where each assembly is itself a lattice of fuel, control, and instrumentation pins, surrounded by water or other material that moderates neutron energy and carries away fission heat. We describe a system for generating geometry and mesh for these systems. The method takes advantage of information about repeated structures in both assembly and core lattices to simplify the overall process. The system allows targeted user intervention midway through the process, enabling modification and manipulation of models for meshing or other purposes. Starting from text files describing assemblies and core, the tool can generate geometry and mesh for these models automatically as well. Simple and complex examples of tool operation are given, with the latter demonstrating generation of meshes with 12 million hexahedral elements in less than 30 minutes on a desktop workstation, using about 4 GB of memory. The tool is released as open source software as part of the MeshKit mesh generation library.

Tautges, T. J.; Jain, R.; Mathematics and Computer Science

2010-01-01

335

Representation of the Auroral and Polar Ionosphere in the International Reference Ionosphere (IRI)  

NASA Technical Reports Server (NTRS)

This issue of Advances in Space Research presents a selection of papers that document the progress in developing and improving the International Reference Ionosphere (IRI), a widely used standard for the parameters that describe the Earths ionosphere. The core set of papers was presented during the 2010 General Assembly of the Committee on Space Research in Bremen, Germany in a session that focused on the representation of the auroral and polar ionosphere in the IRI model. In addition, papers were solicited and submitted from the scientific community in a general call for appropriate papers.

Bilitza, Dieter; Reinisch, Bodo

2013-01-01

336

Estimation of core-damage frequency to evolutionary ALWR (advanced light water reactor) due to seismic initiating events: Task 4. 3. 3  

Microsoft Academic Search

The Electric Power Research Institute (EPRI) is presently developing a requirements document for the design of advanced light water reactors (ALWRs). One of the basic goals of the EPRI ALWR Requirements Document is that the core-damage frequency for an ALWR shall be less than 1.0E-5. To aid in this effort, the Department of Energy's Advanced Reactor Severe Accident Program (ARSAP)

R. D. Brooks; D. G. Harrison; R. L. Summitt

1990-01-01

337

Proposed Design and Operation of a Heat Pipe Reactor using the Sandia National Laboratories Annular Core Test Facility and Existing UZrH Fuel Pins  

NASA Astrophysics Data System (ADS)

Heat Pipe Reactors (HPR) for space power conversion systems offer a number of advantages not easily provided by other systems. They require no pumping, their design easily deals with freezing and thawing of the liquid metal, and they can provide substantial levels of redundancy. Nevertheless, no reactor has ever been operated and cooled with heat pipes, and the startup and other operational characteristics of these systems remain largely unknown. Signification deviations from normal reactor heat removal mechanisms exist, because the heat pipes have fundamental heat removal limits due to sonic flow issues at low temperatures. This paper proposes an early prototypic test of a Heat Pipe Reactor (using existing 20% enriched nuclear fuel pins) to determine the operational characteristics of the HPR. The proposed design is similar in design to the HOMER and SAFE-300 HPR designs (Elliot, Lipinski, and Poston, 2003; Houts, et. al, 2003). However, this reactor uses existing UZrH fuel pins that are coupled to potassium heat pipes modules. The prototype reactor would be located in the Sandia Annular Core Research Reactor Facility where the fuel pins currently reside. The proposed reactor would use the heat pipes to transport the heat from the UZrH fuel pins to a water pool above the core, and the heat transport to the water pool would be controlled by adjusting the pressure and gas type within a small annulus around each heat pipe. The reactor would operate as a self-critical assembly at power levels up to 200 kWth. Because the nuclear heated HPR test uses existing fuel and because it would be performed in an existing facility with the appropriate safety authorization basis, the test could be performed rapidly and inexpensively. This approach makes it possible to validate the operation of a HPR and also measure the feedback mechanisms for a typical HPR design. A test of this nature would be the world's first operating Heat Pipe Reactor. This reactor is therefore called "HPR-1".

Wright, Steven A.; Lipinski, Ronald J.; Pandya, Tara; Peters, Curtis

2005-02-01

338

The voices of Iris: Cinematic representations of the aged woman and Alzheimer's disease in Iris (2001).  

PubMed

Audiences must be critical of film representations of the aged woman living with Alzheimer's disease and of dangerous reinscriptions of stereotypical equations about ageing as deterioration. This paper analyses the representation and decline of the aged woman through the different voices of Iris Murdoch in Richard Eyre's film Iris (2001). Key vocal scenes are considered: On-screen encounters between young and aged Iris, vocal representations of dementia symptoms and silencing Iris as her disease progresses. Further, Iris' recurrent unaccompanied song, "The Lark in the Clear Air," compels audiences to "see" Iris with their ears more than with their eyes, exemplifying the representational power of sound in film. This paper is an appeal for increased debate about sonic representations of aged women, ageing and Alzheimer's disease and dementia in film. The significance of audiences' critical awareness and understanding about the social implications of these representations is discussed. PMID:25370076

Graham, Megan E

2014-11-01

339

Unmasking Immune Reconstitution Inflammatory Syndrome (IRIS)  

PubMed Central

Immune reconstitution inflammatory syndromes (IRIS) in patients with acquired immune deficiency syndrome (AIDS) are characterised by atypical manifestations of opportunistic pathogens. These occur in patients experiencing improvement in CD4 cell counts following receipt of highly active anti-retroviral therapy (HAART). Although well established as a syndrome, IRIS still presents challenges in diagnosis and management. We report five cases of IRIS with diverse clinical presentations and due to different infectious aetiologies. A review of the published literature on this syndrome is also included. PMID:21509214

Balkhair, Abdullah; Ahamed, Sudheer; Sankhla, Dilip

2011-01-01

340

SUMER-IRIS Observations of AR11875  

NASA Astrophysics Data System (ADS)

We present results of the first joint observing campaign of IRIS and SOHO/SUMER. While the IRIS datasets provide information on the chromosphere and transition region, SUMER provides complementary diagnostics on the corona. On 2013-10-24, we observed an active region, AR11875, and the surrounding plage for approximately 4 hours using rapid-cadence observing programs. These datasets include spectra from a small C -class flare which occurs in conjunction with an Ellerman-bomb type event. Our analysis focusses on how the high spatial resolution and slit jaw imaging capabilities of IRIS shed light on the unresolved structure of transient events in the SUMER catalog.

Schmit, Donald; Innes, Davina

2014-05-01

341

Phytochemical investigations on Iris germanica.  

PubMed

Phytochemical investigations on the methanol extract of Iris germanica resulted in the isolation of a new benzene derivative, 2'-methyl-6'-hydroxy cyclohexenyl-3-methyl-1-acetophenone ether (1). Further, another known benzene derivative, isopenol (2), also afforded two known isoflavones, irisolone (3) and irisolidone (4). The structure of the new compound was determined on the basis of spectroscopic data, including 2D-NMR experiments, while the known compounds were identified on the basis of their spectral data and existing literature evidence. The comparison of the spectral data of the irisolidone (3) with that reported for the molecule led us to revise some of the reported 1H-NMR chemical shift assignments. PMID:20077306

Asghar, Syeda Farina; Habib-ur-Rehman; Atta-ur-Rahman; Choudhary, M Iqbal

2010-01-01

342

Coarse-grained parallel genetic algorithm applied to a nuclear reactor core design optimization problem  

Microsoft Academic Search

This work extends the research related to genetic algorithms (GA) in core design optimization problems, which basic investigations were presented in previous work. Here we explore the use of the Island Genetic Algorithm (IGA), a coarse-grained parallel GA model, comparing its performance to that obtained by the application of a traditional non-parallel GA. The optimization problem consists on adjusting several

Cláudio M. N. A. Pereira; Celso M. F. Lapa

2003-01-01

343

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

SciTech Connect

Early fuel removal from the active core of a liquid-metal-cooled fast breeder reactor undergoing a core-disruptive accident may reduce the potential for large energetics resulting from recriticalities. This paper presents a review of analyses with the SIMMER-II computer program of the effectiveness of possible fuel escape paths. Where possible, how SIMMER-II compares with or is validated against experiments that simulated the escape paths also is discussed.

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

1985-01-01

344

Analysis of proposed gamma-ray detection system for the monitoring of core water inventory in a pressurized water reactor  

SciTech Connect

An initial study has been performed of the feasibility of employing an axial array of gamma detectors located outside the pressure vessel to monitor the coolant in a PWR. A one-dimensional transport analysis model is developed for the LOFT research reactor and for a mock-PWR geometry. The gamma detector response to coolant voiding in the core and downcomer has been determined for both geometries. The effects of various conditions (for example, time after shutdown, materials in the transport path, and the relative void fraction in different water regions) on the detector response are studied. The calculational results have been validated by a favorable comparison with LOFT experimental data. Within the limitations and approximations considered in the analysis, the results indicate that the gamma-ray detection scheme is able to unambiguously respond to changes in the coolant inventory within any vessel water region.

Markoff, D.M.

1987-12-01

345

IRIS Launch Animation - Duration: 1:48.  

NASA Video Gallery

This animation demonstrates the launch and deployment of NASA's Interface Region Imaging Spectrograph (IRIS) mission satellite via a Pegasus rocket. The launch is scheduled for June 26, 2013 from V...

346

Accident source terms for boiling water reactors with high burnup cores.  

SciTech Connect

The primary objective of this report is to provide the technical basis for development of recommendations for updates to the NUREG-1465 Source Term for BWRs that will extend its applicability to accidents involving high burnup (HBU) cores. However, a secondary objective is to re-examine the fundamental characteristics of the prescription for fission product release to containment described by NUREG-1465. This secondary objective is motivated by an interest to understand the extent to which research into the release and behaviors of radionuclides under accident conditions has altered best-estimate calculations of the integral response of BWRs to severe core damage sequences and the resulting radiological source terms to containment. This report, therefore, documents specific results of fission product source term analyses that will form the basis for the HBU supplement to NUREG-1465. However, commentary is also provided on observed differences between the composite results of the source term calculations performed here and those reflected NUREG-1465 itself.

Gauntt, Randall O.; Powers, Dana Auburn; Leonard, Mark Thomas

2007-11-01

347

Nonlinear seismic analysis of a reactor structure impact between core components  

NASA Technical Reports Server (NTRS)

The seismic analysis of the FFTF-PIOTA (Fast Flux Test Facility-Postirradiation Open Test Assembly), subjected to a horizontal DBE (Design Base Earthquake) is presented. The PIOTA is the first in a set of open test assemblies to be designed for the FFTF. Employing the direct method of transient analysis, the governing differential equations describing the motion of the system are set up directly and are implicitly integrated numerically in time. A simple lumped-nass beam model of the FFTF which includes small clearances between core components is used as a "driver" for a fine mesh model of the PIOTA. The nonlinear forces due to the impact of the core components and their effect on the PIOTA are computed.

Hill, R. G.

1975-01-01

348

Examination of temperature dependent subgroup formulations in direct whole core transport calculation for power reactors  

SciTech Connect

The traditional subgroup method which has been applied for lattice transport calculations has an inherent limitation for non-uniform temperature distributions. As a measure to incorporate temperature dependence into the subgroup formulation, the subgroup level and number density adjustment method have been proposed. In this paper, the temperature dependent subgroup formulations employed for reflecting the non-uniform temperature effects on the resonance spatial self-shielding are examined for the whole core transport calculation with the thermal feedback. For 2D pin-cell problem with non-uniform temperature profiles, the inherent limitation of conventional subgroup method is confirmed. And the improvement in terms of reactivity is observed with the proposed adjustment scheme. For the real PWR core calculation with thermal feedback in the hot-full-power condition, the noticeable correction for the fuel temperature coefficient by about 10% more negative is obtained with the correction schemes. (authors)

Jung, Y. S.; Lee, U. C.; Joo, H. G. [Dept. of Nuclear Engineering, Seoul National Univ., 599 Gwanak-ro, Gwanak-gu, Seoul (Korea, Republic of)

2012-07-01

349

Two-level algorithm for efficient space-time reactor core calculations  

Microsoft Academic Search

In order to perform more sophisticated transient analyses, Siemens has coupled the nodal core simulator PANBOX2 with the plant analysis code RELAP5\\/MOD2. The coupling replaces the point-kinetics approximation, which is used in the RELAP5 model with the transient three-dimensional neutron diffusion calculations of PANBOX2. This coupling produces more accurate results, but calculation times become very long due to the complexity

C. Jackson; H. Finnemann; D. Cacuci; R. Boeer

1994-01-01

350

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

SciTech Connect

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

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

2012-07-01

351

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

NASA Astrophysics Data System (ADS)

Biometrics is a method of identifying individuals by their physiological or behavioral characteristics. Among other biometric identifiers, iris recognition has been widely used for various applications that require a high level of security. When a conventional iris recognition camera is used, the size and position of the iris region in a captured image vary according to the X, Y positions of a user's eye and the Z distance between a user and the camera. Therefore, the searching area of the iris detection algorithm is increased, which can inevitably decrease both the detection speed and accuracy. To solve these problems, we propose a new method of iris localization that uses wide field of view (WFOV) and narrow field of view (NFOV) cameras. Our study is new as compared to previous studies in the following four ways. First, the device used in our research acquires three images, one each of the face and both irises, using one WFOV and two NFOV cameras simultaneously. The relation between the WFOV and NFOV cameras is determined by simple geometric transformation without complex calibration. Second, the Z distance (between a user's eye and the iris camera) is estimated based on the iris size in the WFOV image and anthropometric data of the size of the human iris. Third, the accuracy of the geometric transformation between the WFOV and NFOV cameras is enhanced by using multiple matrices of the transformation according to the Z distance. Fourth, the searching region for iris localization in the NFOV image is significantly reduced based on the detected iris region in the WFOV image and the matrix of geometric transformation corresponding to the estimated Z distance. Experimental results showed that the performance of the proposed iris localization method is better than that of conventional methods in terms of accuracy and processing time.

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

2013-10-01

352

Simulation of in-core neutron noise measurements for axial void profile reconstruction in boiling water reactors  

SciTech Connect

A possibility to reconstruct the axial void profile from the simulated in-core neutron noise which is caused by density fluctuations in a Boiling Water Reactor (BWR) heated channel is considered. For this purpose, a self-contained model of the two-phase flow regime is constructed which has quantitatively and qualitatively similar properties to those observed in real BWRs. The model is subsequently used to simulate the signals of neutron detectors induced by the corresponding perturbations in the flow density. The bubbles are generated randomly in both space and time using Monte-Carlo techniques. The axial distribution of the bubble production is chosen such that the mean axial void fraction and void velocity follow the actual values of BWRs. The induced neutron noise signals are calculated and then processed by the standard signal analysis methods such as Auto-Power Spectral Density (APSD) and Cross-Power Spectral Density (CPSD). Two methods for axial void and velocity profiles reconstruction are discussed: the first one is based on the change of the break frequency of the neutron auto-power spectrum with axial core elevation, while the second refers to the estimation of transit times of propagating steam fluctuations between different axial detector positions. This paper summarizes the principles of the model and presents a numerical testing of the qualitative applicability to estimate the required parameters for the reconstruction of the void fraction profile from the neutron noise measurements. (authors)

Dykin, V.; Pazsit, I. [Chalmers Univ. of Technology, Div. of Nuclear Engineering, Dept. of Applied Physics, SE-412 96 Gothenburg (Sweden)

2012-07-01

353

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

SciTech Connect

This fuel summary report contains background and summary information for the Peach Bottom Unit 1, High-Temperature, Gas-Cooled Reactor Cores 1 and 2. This report contains detailed information about the fuel in the two cores, the Peach Bottom Unit 1 operating history, nuclear parameters, physical and chemical characteristics, and shipping and storage canister related data. The data in this document have been compiled from a large number of sources and are not qualified beyond the qualification of the source documents. This report is intended to provide an overview of the existing data pertaining to spent fuel management and point to pertinent reference source documents. For design applications, the original source documentation must be used. While all referenced sources are available as records or controlled documents at the Idaho National Engineering and Environmental Laboratory (INEEL), some of the sources were marked as informal or draft reports. This is noted where applicable. In some instances, source documents are not consistent. Where they are known, this document identifies those instances and provides clarification where possible. However, as stated above, this document has not been independently qualified and such clarifications are only included for information purposes. Some of the information in this summary is available in multiple source documents. An effort has been made to clearly identify at least one record document as the source for the information included in this report.

Karel I. Kingrey

2003-04-01

354

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

Microsoft Academic Search

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 the Oak Ridge National Laboratory (ORNL). Damage propagation is postulated to occur from thermal conduction between damaged and undamaged plates due to direct thermal contact. Such direct thermal contact may occur

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

1995-01-01

355

Large Core Code Evaluation Working Group Benchmark Problem Four: neutronics and burnup analysis of a large heterogeneous fast reactor. Part 1. Analysis of benchmark results. [LMFBR  

Microsoft Academic Search

The Large Core Code Evaluation Working Group Benchmark Problem Four was specified to provide a stringent test of the current methods which are used in the nuclear design and analyses process. The benchmark specifications provided a base for performing detailed burnup calculations over the first two irradiation cycles for a large heterogeneous fast reactor. Particular emphasis was placed on the

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

1984-01-01

356

A comparative neutronic feasibility study for a hydrogen, deuterium and helium cold neutron sources situated in the center of a nuclear reactor core  

Microsoft Academic Search

A tool was developed to calculate the average cold neutron flux that could be generated for a spherically shaped cold neutron source situated in the center of a nuclear reactor core. The tool also estimates the subsequent nuclear heating of the cold source. The results were compared for three different cold source mediums; hydrogen, deuterium and helium. The tool utilizes

Malek Chatila

2003-01-01

357

Results of recent reactor-material tests on dispersal of oxide fuel from a disrupted core  

SciTech Connect

The results of experimental investigations and related analyses are reported addressing the dispersal of molten oxide fuel from a disrupted core via various available pathways for the CRBR system. These investigations included the GAPFLOW tests in which pressure-driven and gravity drainage tests were performed using dispersal pathways mocking up the intersubassembly gaps, the CAMEL C6 and C7 tests in which molten fuel entered sodium-filled control assembly ducts under prototypic thermal-hydraulic conditions, and the Lower Internals Drainage (LID) tests in which molten fuel drained downward through simulated below-core structure (orifice plate stacks) as the bottom of control assembly ducts. The results of SHOTGUN tests addressing basic freezing of molten UO/sub 2/ and UO/sub 2//metal mixtures flowing through circular tubes are also reported. Test results have invariably shown the existance of stable UO/sub 2/ crusts on the inside surfaces of the flow paths. Appreciable removal of fuel was indicated prior to freezing-induced immobilization. Application of heat transfer models based upon the presence of stable, insulating fuel crusts tends to overpredict the removal process.

Spencer, B.W.; Wilson, R.J.; Vetter, D.L.; Erickson, E.G.; Dewey, G.

1985-01-01

358

The New Iris Data: Modular Data Generators Iris Ad Michael R. Berthold  

E-print Network

The New Iris Data: Modular Data Generators Iris Adä Michael R. Berthold Nycomed- source environment for data generation. Using an existing graphical data flow tool, the user can combine various types of modules for numeric and categorical data generators. Ad- ditional functionality is added

Berthold, Michael R.

359

Sequences associated with human iris pigmentation.  

PubMed Central

To determine whether and how common polymorphisms are associated with natural distributions of iris colors, we surveyed 851 individuals of mainly European descent at 335 SNP loci in 13 pigmentation genes and 419 other SNPs distributed throughout the genome and known or thought to be informative for certain elements of population structure. We identified numerous SNPs, haplotypes, and diplotypes (diploid pairs of haplotypes) within the OCA2, MYO5A, TYRP1, AIM, DCT, and TYR genes and the CYP1A2-15q22-ter, CYP1B1-2p21, CYP2C8-10q23, CYP2C9-10q24, and MAOA-Xp11.4 regions as significantly associated with iris colors. Half of the associated SNPs were located on chromosome 15, which corresponds with results that others have previously obtained from linkage analysis. We identified 5 additional genes (ASIP, MC1R, POMC, and SILV) and one additional region (GSTT2-22q11.23) with haplotype and/or diplotypes, but not individual SNP alleles associated with iris colors. For most of the genes, multilocus gene-wise genotype sequences were more strongly associated with iris colors than were haplotypes or SNP alleles. Diplotypes for these genes explain 15% of iris color variation. Apart from representing the first comprehensive candidate gene study for variable iris pigmentation and constituting a first step toward developing a classification model for the inference of iris color from DNA, our results suggest that cryptic population structure might serve as a leverage tool for complex trait gene mapping if genomes are screened with the appropriate ancestry informative markers. PMID:14704187

Frudakis, Tony; Thomas, Matthew; Gaskin, Zach; Venkateswarlu, K; Chandra, K Suresh; Ginjupalli, Siva; Gunturi, Sitaram; Natrajan, Sivamani; Ponnuswamy, Viswanathan K; Ponnuswamy, K N

2003-01-01

360

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

SciTech Connect

Gas-cooled fast reactor (GFR) designs are being developed to meet Gen IV goals of sustainability, economics, safety and reliability, and proliferation resistance and physical protection as part of an International Generation IV Nuclear Energy System Research Initiative effort. Different organizations are involved in the development of a variety of GFR design concepts. The current analysis has focused on the evaluation of low-pressure drop, pin-core designs with favorable passive cooling properties. Initial evaluation of the passive cooling safety case for the GFR during depressurized decay heat removal accidents with concurrent loss of electric power have resulted in requirements for a reduction of core power density to the 100 w/cc level and a low core pressure drop of 0.5 bars. Additional design constraints and the implementation of their constraints are evaluated in this study to enhance and passive cooling properties of the reactor. Passive cooling is made easier by a flat radial distribution of the decay heat. One goal of this study was to evaluate the radial power distribution and determine to what extent it can be flattened, since the decay heat is nearly proportional to the fission power at shutdown. In line with this investigation of the radial power profile, an assessment was also made of the control rod configuration. The layout provided a large number of control rod locations with a fixed area provided for control rods. The number of control rods was consistent with other fast reactor designs. The adequacy of the available control rod locations was evaluated. Future studies will be needed to optimize the control rod designs and evaluate the shutdown system. The case for low pressure drop core can be improved by the minimization of pressure drop sources such as the number of required fuel spacers in the subassembly design and by the details of the fuel pin design. The fuel pin design is determined by a number of neutronic, thermal-hydraulic (gas dynamics) and fuel performance considerations. For the purposes of this study, the starting point is the fuel pin design established by the CEA-ANL/US I-NERI collaboration project for the selected 2400 MWt large rector option. Structural mechanics factors are now included in the design assessment. In particular, thermal bowing establishes a bound on the minimum of fuel pin spacers required in each fuel subassembly to prevent the local flow channel restrictions and pin-to-pin mechanical interaction. There are also fabrication limitations on the maximum length of SiC fuel pin cladding which can be manufactured. This geometric limitation effects the minimum ceramic clad thickness which can be produced. This ties into the fuel pin heat transfer and temperature thresholds. All these additional design factors were included in the current iteration on the subassembly design to produce a lower core pressure drop. A more detailed definition of the fuel pin/subassembly design is proposed here to meet these limitations. This subassembly design was then evaluated under low pressure natural convection conditions to assess its acceptability for the decay heat removal accidents. A number of integrated decay heat removal (DHR) loop plus core calculations were performed to scope the thermal-hydraulic response of the subassembly design to the accidents of interest. It is evident that there is a large sensitivity to the guard containment back pressure for these designs. The implication of this conclusion and possible design modifications to reduce this sensitivity will be explored under the auspices of the International GENIV GFR collaborative R&D plan. Chapter 2 describes the core reference design for the 2,400 MWt GFR being evaluated. The methodology, modeling, and codes used in the analysis of the fuel pin structural behavior are described in Chapter 3. Chapter 4 provides the result of the thermal-hydraulic study of the assembly design for the accidents of interest. An evaluation of the performance and control rod reactivity control is also presented in Chapter 2.

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

2006-07-31

361

ORNL Biometric Eye Model for Iris Recognition  

SciTech Connect

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

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

2012-01-01

362

Experimental and Numerical Observations of Hydrate Reformation during Depressurization in a Core-Scale Reactor  

SciTech Connect

Gas hydrate has been predicted to reform around a wellbore during depressurization-based gas production from gas hydrate-bearing reservoirs. This process has an adverse effect on gas production rates and it requires time and sometimes special measures to resume gas flow to producing wells. Due to lack of applicable field data, laboratory scale experiments remain a valuable source of information to study hydrate reformation. In this work, we report laboratory experiments and complementary numerical simulations executed to investigate the hydrate reformation phenomenon. Gas production from a pressure vessel filled with hydrate-bearing sand was induced by depressurization with and without heat flux through the boundaries. Hydrate decomposition was monitored with a medical X-ray CT scanner and pressure and temperature measurements. CT images of the hydrate-bearing sample were processed to provide 3-dimensional data of heterogeneous porosity and phase saturations suitable for numerical simulations. In the experiments, gas hydrate reformation was observed only in the case of no-heat supply from surroundings, a finding consistent with numerical simulation. By allowing gas production on either side of the core, numerical simulations showed that initial hydrate distribution patterns affect gas distribution and flow inside the sample. This is a direct consequence of the heterogeneous pore network resulting in varying hydraulic properties of the hydrate-bearing sediment.

Seol, Yongkoo; Myshakin, Evgeniy

2011-01-01

363

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

SciTech Connect

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

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

2006-10-13

364

On a Quest to Improve the Solar Forcing in IRI  

NASA Astrophysics Data System (ADS)

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

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

2013-12-01

365

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

SciTech Connect

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

Farzad Rahnema

2009-11-12

366

Cataract surgery combined with implantation of an artificial iris.  

PubMed

We describe 6 patients who presented with cataract or aphakia and absent or nonfunctional irides. The etiologies included congenital aniridia, traumatic iris loss, and chronic mydriasis secondary to recurrent herpetic uveitis. In 5 eyes, a prosthetic iris was successfully implanted in combination with small incision cataract surgery. In 2 eyes, a single-piece iris diaphragm and optical lens was implanted. Artificial irides offer a safe alternative for patients who previously had no viable options for iris reconstruction. PMID:10569173

Osher, R H; Burk, S E

1999-11-01

367

Artificial iris diaphragm and silicone oil surgery.  

PubMed

In order to avoid contact between silicone oil and the cornea and subsequent painful corneal dystrophy in aniridial eyes, an artificial iris diaphragm was constructed. It consists of polymethylacrylat (PMMA) and simulates the situation of the iris with a central pupillary opening and inferior iridectomy. To date, these diaphragms have been implanted in 11 cases of the severest ocular trauma with accompanying aniridia and proliferative vitreoretinopathy. In the presence of sufficient residual secretion of the ciliary body (9 cases), the diaphragm assumes the function of normal iris and prevents the silicone oil from coming into contact with the corneal endothelium. The transparent diaphragm ensures a view through to the fundus. In the early postoperative period, there was, as anticipated, a fibrinous reaction in the area of the anterior segment. PMID:1455092

Heimann, K; Konen, W

1992-01-01

368

Jets and Bombs: Characterizing IRIS Spectra  

NASA Astrophysics Data System (ADS)

For almost two decades, SUMER has provided an unique perspective on explosive events in the lower solar atmosphere. One of the hallmark observations during this tenure is the identification of quiet sun bi-directional jets in the lower transition region. We investigate these events through two distinct avenues of study: a MHD model for reconnection and the new datasets of the Interface Region Imaging Spectrograph (IRIS). Based on forward modeling optically thin spectral profiles, we find the spectral signatures of reconnection can vary dramatically based on viewing angle and altitude. We look to the IRIS data to provide a more complete context of the chromospheric and coronal environment during these dynamic events. During a joint IRIS-SUMER observing campaign, we observed spectra of multiple jets, a small C flare, and an Ellerman bomb event. We discuss the questions that arise from the inspection of these new data.

Schmit, Donald; Innes, Davina

2014-06-01

369

Iris-Corinna Schwarz (PhD, MAPS) Department of Linguistics iris@ling.su.se Phonetics Lab/Babylab  

E-print Network

Iris-Corinna Schwarz (PhD, MAPS) Department of Linguistics iris@ling.su.se Phonetics Lab/Babylab w current Assistant professor at the Phonetics lab/Babylab, Department of Linguistics, Stockholm University

370

Iris Biometic Processor Enhanced Module FPGA-Based Design  

Microsoft Academic Search

Iris Identification is nowadays one of the most promising techniques in Authentication. Most modern iris recognition systems are currently deployed on traditional sequential digital systems, such as a simple DSPs or MIPS processor. However, in this method, we can only match each data one by one, which will waste much time. In this study, iris matching, a repeatedly executed portion

Zhou Hu-lin; Xie Mei

2010-01-01

371

IRIS TOXICOLOGICAL REVIEW AND SUMMARY DOCUMENTS FOR BERYLLIUM AND COMPOUNDS  

EPA Science Inventory

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

372

Iris recognition based on score level fusion by using SVM  

Microsoft Academic Search

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

Hyun-ae Park; Kang Ryoung Park

2007-01-01

373

[Bilateral acute depigmentation of the iris syndrome].  

PubMed

Bilateral acute depigmentation of the iris syndrome (BADI syndrome) is a new clinical entity. Young females from 20 to 45 years of age are most commonly affected. It is characterized by bilateral nontransilluminating depigmentation of the iris stroma. During the acute phase, this clinical entity also combines with red painful eye, pigmentation of the trabecular meshwork, anterior chamber flare, circulating pigment, and pigmented deposit on the endothelium cornea. At the acute stage, the symptoms are controlled with topical corticosteroid treatment. The prognosis is good. We report a 41-year-old woman presenting with BADI syndrome. PMID:21531477

Portmann, A; Gueudry, J; Siahmed, K; Muraine, M

2011-05-01

374

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

E-print Network

The MIT Research Reactor (MITR) is evaluating the conversion from highly enriched uranium (HEU) to low enrichment uranium (LEU) fuel. In addition to the fuel element re-design from 15 to 18 plates per element, a reactor ...

Chiang, Keng-Yen

2012-01-01

375

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

E-print Network

Worldwide, there is limited test reactor capacity to perform the required irradiation experiments on advanced fast reactor materials and fuel designs. This is particularly true in the U.S., which no longer has an operating ...

Ellis, Tyler Shawn

2009-01-01

376

Results of analyses performed on concrete cores removed from floors and D-ring walls of the TMI2 reactor building  

Microsoft Academic Search

The March 28, 1979 loss-of-coolant accident at Three Mile Island Unit 2 (TMI-2) exposed about 7200 m² of concrete surfaces within the Reactor building to liquid and vapor-phase contaminants. The majority of those surfaces are protected by coatings of epoxy-based, nuclear grade paints. during September 1983, seventeen high quality cores were extracted from the concrete floors and D-ring walls ar

C. V. McIsaac; C. M. Davis; J. T. Horan; D. G. Keefer

1984-01-01

377

Results of a Neutronic Simulation of HTR-Proteus Core 4.2 using PEBBED and other INL Reactor Physics Tools: FY09 Report  

Microsoft Academic Search

The Idaho National Laboratorys deterministic neutronics analysis codes and methods were applied to the computation of the core multiplication factor of the HTR-Proteus pebble bed reactor critical facility. A combination of unit cell calculations (COMBINE-PEBDAN), 1-D discrete ordinates transport (SCAMP), and nodal diffusion calculations (PEBBED) were employed to yield keff and flux profiles. Preliminary results indicate that these tools, as

Hans D. Gougar

2009-01-01

378

[Intraoperative floppy iris syndrome versus lens-iris diaphragm retropulsion syndrome].  

PubMed

Intraoperative floppy iris syndrome (IFIS) as well as lens-iris diaphragm retropulsion syndrome (LIDRS) may negatively influence the course of the modern cataract procedure. In case of simultaneous presence of both syndromes, the LIDRS may paradoxically ease the surgery by means of partial or transient elimination of the IFIS. It causes the dilatation of the pupil, pushes the iris back, and deepens the anterior chamber and this way reduces the fluttering of the iris and makes possible to perform important phases of the surgery, during that we are working in the proximity of the pupil. In case of coincidence with the IFIS, after the spontaneous interruption of the LIDRS, we can repeatedly induce it by means of switching the irrigation off and after shallowing of the anterior chamber by turning the irrigation on again. PMID:19110963

Mazal, Z

2008-11-01

379

Pressure Vessel and Internals of the International Reactor Innovative and Secure  

SciTech Connect

IRIS (International Reactor Innovative and Secure) is a modular, integral light water cooled, low-to-medium power reactor, which addresses the requirements defined by the US DOE for Generation IV reactors. Its integrated layout features a pressure vessel containing all the main primary circuit components: the internals and the biological shield, here described together with the pressure vessel, plus the steam generators, the pressurizer, and the main coolant pumps described in companion papers. For this reason the pressure vessel is a crucial component of the plant, which deserves the most demanding design effort. The wide inner annulus around the core is exploited to insert steel plates, in order to improve the inner shielding capability up to the elimination of the external biological shielding and to simplify decommissioning activities by having all the irradiated components inside the vessel. (authors)

Lombardi, C.V.; Padovani, E.; Cammi, A. [Politecnico di Milano, Piazza Leonardo da Vinci, 32, 20133 Milano (Italy); Collado, J.M. [Equipos Nucleares S.A. (Spain); Santoro, R.T.; Barnes, J.M. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States)

2002-07-01

380

A preliminary feasibility study of passive in-core thermionic reactors for highly compact space nuclear power systems  

Microsoft Academic Search

Results of a preliminary feasibility study on a new concept for a highly compact space reactor power systems are presented. Notwithstanding the preliminary nature of the present study, the results which include a new space reactor configuration and its associated technologies indicate promising avenues for the devleopment of highly compact space reactors. The calculations reported in this study include a

A. G. Parlos; E. U. Khan; R. Frymire; S. Negron; J. K. Thomas; K. L. Peddicord

1991-01-01

381

A preliminary feasibility study of passive in-core thermionic reactors for highly compact space nuclear power systems  

Microsoft Academic Search

Results of a preliminary feasibility study on a new concept for a highly compact space reactor power systems are presented. Notwithstanding the preliminary nature of the present study, the results which include a new space reactor configuration and its associated technologies indicate promising avenues for the development of highly compact space reactors. The calculations reported in this study include a

A. G. Parlos; E. U. Khan; R. Frymire; S. Negron; J. K. Thomas; K. L. Peddicord

1991-01-01

382

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

DOEpatents

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

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

1987-01-01

383

Evaluation of a Method for Remote Detection of Fuel Relocation Outside the Original Core Volumes of Fukushima Reactor Units 1-3  

SciTech Connect

This paper presents the results of a study to evaluate the feasibility of remotely detecting and quantifying fuel relocation from the core to the lower head, and to regions outside the reactor vessel primary containment of the Fukushima 1-3 reactors. The goals of this study were to determine measurement conditions and requirements, and to perform initial radiation transport sensitivity analyses for several potential measurement locations inside the reactor building. The radiation transport sensitivity analyses were performed based on reactor design information for boiling water reactors (BWRs) similar to the Fukushima reactors, ORIGEN2 analyses of 3-cycle BWR fuel inventories, and data on previously molten fuel characteristics from TMI- 2. A 100 kg mass of previously molten fuel material located on the lower head of the reactor vessel was chosen as a fuel interrogation sensitivity target. Two measurement locations were chosen for the transport analyses, one inside the drywell and one outside the concrete biological shield surrounding the drywell. Results of these initial radiation transport analyses indicate that the 100 kg of previously molten fuel material may be detectable at the measurement location inside the drywell, but that it is highly unlikely that any amount of fuel material inside the RPV will be detectable from a location outside the concrete biological shield surrounding the drywell. Three additional fuel relocation scenarios were also analyzed to assess detection sensitivity for varying amount of relocated material in the lower head of the reactor vessel, in the control rods perpendicular to the detector system, and on the lower head of the drywell. Results of these analyses along with an assessment of background radiation effects and a discussion of measurement issues, such as the detector/collimator design, are included in the paper.

Douglas W. Akers; Edwin A. Harvego

2012-08-01

384

IRIS Update Batch 1, Group 1  

EPA Science Inventory

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

385

INDUSTRIAL RESEARCH AND DEVELOPMENT INFORMATION SYSTEM (IRIS)  

EPA Science Inventory

The National Science Foundation's (NSF) Industrial Research and Development Information System (IRIS) links an online interface to a historical database with more than 2,500 statistical tables containing all industrial research and development (R&D) data published by NSF since 19...

386

Hurricane Iris from TRMM: October 9, 2001  

NSDL National Science Digital Library

TRMM views hurricane Iris as it strikes Honduras, October 9, 2001. Time is about 09:00 UT, Orbit T03. Isosurfaces are: Yellow=0.5 inches-hour, Green=1.0 inches-hour, Red=2.0 inches-hour on rainfall rates.

Bridgman, Tom; Adler, Robert

2001-10-09

387

IRI/LDEO Introduction to Climate Data  

NSDL National Science Digital Library

A collection of the datasets from the IRI/LDEO Climate Data Library, which contain important information about our planet Earth. Datasets include: Topography, ENSO (El Nino-Southern Oscillation) Monitor, Historical Temperature and Precipitation, and Ocean Climatology, including ocean temperature, salinity, and nutrients, including dissolved oxygen, nitrate, phosphate, and silicate. Figures are very easy to manipulate and the parameters are explained in detail.

2010-11-03

388

The IRIS Mission: A Colorful EPO Program  

NASA Astrophysics Data System (ADS)

We will overview NASA’s IRIS mission EPO program, which includes a nationwide spectroscopy contest, K-12 resources, a summer program for undergraduates, informal outreach elements, and a dynamic social media program based on the highly successful Camilla/Little SDO program for NASA’s SDO mission.

Scherrer, Deborah K.

2012-05-01

389

NASA HyspIRI Workshop Report  

Technology Transfer Automated Retrieval System (TEKTRAN)

On October 21-23rd 2008 NASA held a three-day workshop to consider the Hyperspectral and Infrared Imager (HyspIRI) mission recommended for implementation by the 2007 National Research Council Earth Science Decadal Survey. The open workshop provided a forum to present the initial observational requir...

390

Iridals from Iris tectorum and Belamcanda chinensis  

Microsoft Academic Search

Three iridals, iridotectorals A and B, and iridobelamal A, were isolated from rhizomes of Iris tectorum and Belamcanda chinensis, respectively, along with five known iridals. Their structures were elucidated on the basis of spectral evidence. The human promyelocytic leukemia (HL-60) cell-adhesion activity of the eight iridals is also discussed.

Kunihiko Takahashfi; Yoji Hoshino; Sumiko Suzuki; Yoshio Hano; Taro Nomura

2000-01-01

391

Checking the new IRI model The bottomside B parameters  

E-print Network

Electron density profiles obtained at Pruhonice (50.0, 15.0), El Arenosillo (37.1, 353.2) and Havana (23, 278) were used to check the bottom-side B parameters BO (thickness parameter) and B1 (shape parameter) predicted by the new IRI - 2000 version. The electron density profiles were derived from ionograms using the ARP technique. The data base includes daytime and nighttime ionograms recorded under different seasonal and solar activity conditions. Comparisons with IRI predictions were also done. The analysis shows that: a) The parameter B1 given by IRI 2000 reproduces better the observed ARP values than the IRI-90 version and b) The observed BO values are in general well reproduced by both IRI versions: IRI-90 and IRI-2000.

Mosert, M; Ezquer, R; Lazo, B; Miro, G

2002-01-01

392

The results of the CCI-3 reactor material experiment investigating 2-D core-concrete interaction and debris coolability with a silliceous concrete crucible.  

SciTech Connect

The OECD-sponsored Melt Coolability and Concrete Interaction (MCCI) program is conducting reactor material experiments and associated analysis with the objectives of resolving the ex-vessel debris coolability issue, and to 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 for future plants. Despite years of international research, there are remaining 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 result, there are differences in the 2-D cavity erosion predicted by codes such as MELCOR, WECHSL, and COSACO. In the continuing effort to bridge this data gap, the third in a series of large scale Core-Concrete Interaction experiments (CCI-3) has been conducted as part of the MCCI program. This test involved the interaction of a 375 kg core-oxide melt within a two-dimensional siliceous concrete crucible. The initial phase of the test was conducted under dry conditions. After a predetermined ablation depth was reached, 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 summary description of the test facility and an overview of test results.

Farmer, M. T.; Basu, S.; Nuclear Engineering Division; NRC

2006-01-01

393

Vapor core propulsion reactors  

NASA Technical Reports Server (NTRS)

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

Diaz, Nils J.

1991-01-01

394

SM2 REACTOR CORE AND VESSEL REVIEW REPORT FOR PERIOD DECEMBER 15, 1959 TO MARCH 18, 1960  

Microsoft Academic Search

The SM-2 core lifetime was calculated and stuck rod criteria formulated. ; The control system could shutdown the core under the most adverse conditions ; with any single rod stuck out. The one-half-inch EuâOâ flux ; supressors adequately supressed the power spike at the bottom of the core. The ; nuclear effects of increasing the dimensione of the fuel matrix

Hoover

1960-01-01

395

Risk-informed design guidance for a Generation-IV gas-cooled fast reactor emergency core cooling system  

E-print Network

Fundamental objectives of sustainability, economics, safety and reliability, and proliferation resistance, physical protection and stakeholder relations must be considered during the design of an advanced reactor. However, ...

Delaney, Michael J. (Michael James), 1979-

2004-01-01

396

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

Microsoft Academic Search

Summary  A routine method is described for the determination of total IRI (imraunoreactive insulin) in insulintreated diabetics. The method involves an easy acid ethanol extraction, whereby antibody-bound IRI is dissociated and separated, together with the free IRI from the serum proteins and the antibodies. The recovery of IRI in this procedure is about 80%. After the separation, the isolated total IRI

Lise G. Heding

1972-01-01

397

[Cultivation of Iris ensata Thunb. callus tissue].  

PubMed

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

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

2004-01-01

398

Experimental detailed power distribution in a fast spectrum thermionic reactor fuel element at the core/BeO reflector interface region  

NASA Technical Reports Server (NTRS)

A zero-power critical assembly was designed, constructed, and operated for the prupose of conducting a series of benchmark experiments dealing with the physics characteristics of a UN-fueled, Li-7-cooled, Mo-reflected, drum-controlled compact fast reactor for use with a space-power conversion system. The critical assembly was modified to simulate a fast spectrum advanced thermionics reactor by: (1) using BeO as a reflector in place of some of the existing molybdenum, (2) substituting Nb-1Zr tubing for some of the existing Ta tubing, and (3) inserting four full-scale mockups of thermionic type fuel elements near the core and BeO reflector boundary. These mockups were surrounded with a buffer zone having the equivalent thermionic core composition. In addition to measuring the critical mass of this thermionic configuration, a detailed power distribution in one of the thermionic element stages in the mixed spectrum region was measured. A power peak to average ratio of two was observed for this fuel stage at the midplane of the core and adjacent to the reflector. Also, the power on the outer surface adjacent to the BeO was slightly more than a factor of two larger than the power on the inside surface of a 5.08 cm (2.0 in.) high annular fuel segment with a 2.52 cm (0.993 in. ) o.d. and a 1.86 cm (0.731 in.) i.d.

Klann, P. G.; Lantz, E.

1973-01-01

399

Atypical Mycobacterial Panophthalmitis Seen With Iris Nodules  

Microsoft Academic Search

typical mycobacterial infections are frequent, late complications of human immunode- ficiency virus infections and may have a variety of clinical manifestations. We describe a patient with end-stage acquired immune deficiency syndrome and disseminated atypi- cal mycobacteriosis caused by Mycobacterium avium-intracellulare complex, with promi- nent iris nodules as the initial manifestation of a unilateral localized panophthalmitis. Acid-fast bacilli were identified cytologically

Pearl S. Rosenbaum; Joyce N. Mbekeani; Yvonne Kress

400

IRIS: Industrial Research and Development Information System  

NSDL National Science Digital Library

The National Science Foundation's Industrial Research and Development Information System (IRIS) houses a database of all of the statistics produced and published from the 1953-1998 cycles of the annual Survey of Industrial Research and Development (R&D). The statistics would be useful to workers in economics or anyone interested in learning about how funds are allocated among research areas. NSF states that the results of the survey are used by government agencies, corporations, and research organizations to determine productivity factors, formulate tax policy, and to investigate company performance. The statistics available in IRIS describe national estimates of the total expenditures on R&D performed within the United States by industrial firms, given in dollar amounts. Tabulations from the survey contain R&D statistics by industry, size of company, source of funds, character of R&D, R&D as a percentage of net sales, and R&D contracted to outside organizations and performed outside of the United States. They also contain estimates of the sales and total employment of R&D-performing companies, employment of R&D scientists and engineers, and statistics by state. Users have a variety of options for searching and browsing the Excel tables and Word documentation in IRIS -- by year, topic, or by measure -- and the resulting tables can display all years combined or just selected years.

2001-01-01

401

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

NASA Astrophysics Data System (ADS)

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

Ucar, Dundar

402

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

NASA Astrophysics Data System (ADS)

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

Setyawan, Daddy; Rohman, Budi

2014-09-01

403

Simulation of (16)O (n, p) (16)N reaction rate and nitrogen-16 inventory in a high performance light water reactor with one pass core.  

PubMed

The rate of activation of the isotope (16)O to (16)N in a typical HPLWR one pass concept was calculated using MCNP code. A mathematical model was used to track the inventory of the radioisotope (16)N in a unit mass of coolant traversing the system. The water leaving the moderator channels has the highest activity in the circuit, but due to interaction with fresh coolant at the lower plenum, the activity is downscaled. The calculated core exit activity is higher than values reported in literature for commercial boiling water reactors. PMID:25084129

Kebwaro, Jeremiah Monari; Zhao, Yaolin; He, Chaohui

2014-12-01

404

Procedure of recovery of pin-by-pin fields of energy release in the core of VVER-type reactor for the BIPR-8 code  

NASA Astrophysics Data System (ADS)

The procedure of recovery of pin-by-pin energy-release fields for the BIPR-8 code and the algorithm of the BIPR-8 code which is used in nodal computation of the reactor core and on which the recovery of pin-by-pin fields of energy release is based are briefly described. The description and results of the verification using the module of recovery of pin-by-pin energy-release fields and the TVS-M program are given.

Gordienko, P. V.; Kotsarev, A. V.; Lizorkin, M. P.

2014-12-01

405

An iris recognition algorithm based on DCT and GLCM  

NASA Astrophysics Data System (ADS)

With the enlargement of mankind's activity range, the significance for person's status identity is becoming more and more important. So many different techniques for person's status identity were proposed for this practical usage. Conventional person's status identity methods like password and identification card are not always reliable. A wide variety of biometrics has been developed for this challenge. Among those biologic characteristics, iris pattern gains increasing attention for its stability, reliability, uniqueness, noninvasiveness and difficult to counterfeit. The distinct merits of the iris lead to its high reliability for personal identification. So the iris identification technique had become hot research point in the past several years. This paper presents an efficient algorithm for iris recognition using gray-level co-occurrence matrix(GLCM) and Discrete Cosine transform(DCT). To obtain more representative iris features, features from space and DCT transformation domain are extracted. Both GLCM and DCT are applied on the iris image to form the feature sequence in this paper. The combination of GLCM and DCT makes the iris feature more distinct. Upon GLCM and DCT the eigenvector of iris extracted, which reflects features of spatial transformation and frequency transformation. Experimental results show that the algorithm is effective and feasible with iris recognition.

Feng, G.; Wu, Ye-qing

2008-04-01

406

The Importance of Being Random: Statistical Principles of Iris Recognition  

NSDL National Science Digital Library

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

Daugman, John.

2001-01-01

407

New algorithm for iris recognition based on video sequences  

NASA Astrophysics Data System (ADS)

Among existing biometrics, iris recognition systems are among the most accurate personal biometric identification systems. However, the acquisition of a workable iris image requires strict cooperation of the user; otherwise, the image will be rejected by a verification module because of its poor quality, inducing a high false reject rate (FRR). The FRR may also increase when iris localization fails or when the pupil is too dilated. To improve the existing methods, we propose to use video sequences acquired in real time by a camera. In order to keep the same computational load to identify the iris, we propose a new method to estimate the iris characteristics. First, we propose a new iris texture characterization based on Fourier-Mellin transform, which is less sensitive to pupil dilatations than previous methods. Then, we develop a new iris localization algorithm that is robust to variations of quality (partial occlusions due to eyelids and eyelashes, light reflects, etc.), and finally, we introduce a fast and new criterion of suitable image selection from an iris video sequence for an accurate recognition. The accuracy of each step of the algorithm in the whole proposed recognition process is tested and evaluated using our own iris video database and several public image databases, such as CASIA, UBIRIS, and BATH.

Bourennane, Salah; Fossati, Caroline; Ketchantang, William

2010-07-01

408

Limbus Impact on Off-angle Iris Degradation  

SciTech Connect

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

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

2013-01-01

409

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

E-print Network

The accurate analysis of nuclear reactor transients frequently requires that neutronics, thermal-hydraulics and feedback be included. A number of coupled neutronics/thermal-hydraulics codes have been developed for this ...

Griggs, D. P.

1984-01-01

410

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

E-print Network

The MIT Research Reactor (MITR) is in the process of conducting a design study to convert from High Enrichment Uranium (HEU) fuel to Low Enrichment Uranium (LEU) fuel. The currently selected LEU fuel design contains 18 ...

Wang, Yunzhi (Yunzhi Diana)

2009-01-01

411

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

Code of Federal Regulations, 2011 CFR

...in an acceptable evaluation model or in the application of such a model that affects the temperature...at a rate in excess of the capability of the reactor coolant makeup...system. (2) An evaluation model is the calculational...

2011-01-01

412

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

Code of Federal Regulations, 2010 CFR

...in an acceptable evaluation model or in the application of such a model that affects the temperature...at a rate in excess of the capability of the reactor coolant makeup...system. (2) An evaluation model is the calculational...

2010-01-01

413

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

E-print Network

Oak Ridge National Laboratory OTOC Once-Through-and-Out Cycle P&T Partitioning and Transmutation PWR Pressurized Water Reactor SNF Spent Nuclear Fuel THTR Thorium High Temperature Reactor TRISO Tri-structural Isotropic TRU Transuranium Nuclide... content of the fuel. In discharged spent nuclear fuels (SNF), the reduction of radiotoxicity is mainly driven by the decay of the radionuclides. The time required to attain a tolerable level of toxicity is governed by the half lives of the radioactive...

Alajo, Ayodeji Babatunde

2011-08-08

414

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

SciTech Connect

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

Croucher, D.W.

1981-05-01

415

The Results of the CCI-3 Reactor Material Experiment Investigating 2-D Core-Concrete Interaction and Debris Coolability with a Siliceous Concrete Crucible  

SciTech Connect

The OECD-sponsored Melt Coolability and Concrete Interaction (MCCI) program conducted reactor materials experiments and associated analysis to achieve the following two objectives: 1) resolve 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 are remaining 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 result, there are differences in the 2-D cavity erosion profiles predicted by codes such as WECHSL, COSACO, TOLBIAC, MEDICIS, and MELCOR. In the continuing effort to bridge this data gap, the third in a series of large scale Core-Concrete Interaction experiments (CCI-3) has been conducted as part of the MCCI program. This test investigated the long-term interaction of a 375 kg core-oxide melt within a two-dimensional siliceous concrete crucible. The initial phase of the test was conducted under dry conditions. After a predetermined time interval, the cavity was flooded with water 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 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)

2006-07-01

416

Nuclear safety analyses and core design calculations to convert the Texas A & M University Nuclear Science Center reactor to low enrichment uranium fuel. Final report  

SciTech Connect

This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclear safety analysis capabilities established and maintained at the NSCR.

Parish, T.A.

1995-03-02

417

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

SciTech Connect

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

Hans D. Gougar

2009-08-01

418

Updating of ASME Nuclear Code Case N-201 to Accommodate the Needs of Metallic Core Support Structures for High Temperature Gas Cooled Reactors Currently in Development  

SciTech Connect

On September 29, 2005, ASME Standards Technology, LLC (ASME ST-LLC) executed a multi-year, cooperative agreement with the United States DOE for the Generation IV Reactor Materials project. The project's objective is to update and expand appropriate materials, construction, and design codes for application in future Generation IV nuclear reactor systems that operate at elevated temperatures. Task 4 was embarked upon in recognition of the large quantity of ongoing reactor designs utilizing high temperature technology. Since Code Case N-201 had not seen a significant revision (except for a minor revision in September, 2006 to change the SA-336 forging reference for 304SS and 316SS to SA-965 in Tables 1.2(a) and 1.2(b), and some minor editorial changes) since December 1994, identifying recommended updates to support the current high temperature Core Support Structure (CSS) designs and potential new designs was important. As anticipated, the Task 4 effort identified a number of Code Case N-201 issues. Items requiring further consideration range from addressing apparent inconsistencies in definitions and certain material properties between CC-N-201 and Subsection NH, to inclusion of additional materials to provide the designer more flexibility of design. Task 4 developed a design parameter survey that requested input from the CSS designers of ongoing high temperature gas cooled reactor metallic core support designs. The responses to the survey provided Task 4 valuable input to identify the design operating parameters and future needs of the CSS designers. Types of materials, metal temperature, time of exposure, design pressure, design life, and fluence levels were included in the Task 4 survey responses. The results of the survey are included in this report. This research proves that additional work must be done to update Code Case N-201. Task 4 activities provide the framework for the Code Case N-201 update and future work to provide input on materials. Candidate materials such as Type 321 and Type 347 austenitic stainless steels, Modified 9Cr-1Mo steel for core support structure construction, and Alloy 718 for Threaded Structural Fasteners were among the recommended materials for inclusion in the Code Case. This Task 4 Report identifies the need to address design life beyond 3 x 105 hours, especially in consideration of 60-year design life. A proposed update to the latest Code Case N-201 revision (i.e., Code Case N-201-5) including the items resolved in this report is included as Appendix A.

Mit Basol; John F. Kielb; John F. MuHooly; Kobus Smit

2007-05-02

419

A three-dimensional neutronics-thermohydraulics simulation of core disruptive accident in sodium-cooled fast reactor  

Microsoft Academic Search

The SIMMER-III code is a two-dimensional, multi-velocity-field, multi-phase, multi-component, Eulerian, fluid-dynamics code coupled with a fuel-pin model and a space- and energy-dependent neutron transport kinetics model. Since the three-dimensional representation of the core enables realistic distribution of the materials constituting the core, including control rods, SIMMER-IV has been developed as a direct extension of SIMMER-III to three dimensions with retaining

Hidemasa Yamano; Yoshiharu Tobita; Satoshi Fujita

2009-01-01

420

Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2008  

SciTech Connect

This report documents progress made during FY 2008 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Scoping experiments with various manufacturing methods for forming the LEU alloy profile are presented.

Primm, Trent [ORNL; Chandler, David [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Sease, John D [ORNL; Jolly, Brian C [ORNL

2009-03-01

421

Isolated Bilateral Congenital Iris Sphincter Agenesis  

PubMed Central

Purpose. We herein report a patient with bilateral congenital total iris sphincter agenesis with no other abnormality detected on systemic examination. Methods. A 24-year-old laborer presented to us for a routine checkup with complaint of photophobia and inability to work under sunlight. Examination revealed bilateral absence of sphincter and 6.5?mm pupil in both eyes in the undilated state. Results. Accommodation was poor in both eyes. Systemic examination was within normal limits. He was prescribed bifocal photochromic glasses for constant wear. Conclusions. Congenital sphincter agenesis can occur in an isolated form without systemic abnormalities which can be managed conservatively. PMID:22606462

Rao, Aparna

2011-01-01

422

FACE, EXPRESSION, AND IRIS RECOGNITION USING LEARNING-BASED APPROACHES  

E-print Network

FACE, EXPRESSION, AND IRIS RECOGNITION USING LEARNING-BASED APPROACHES by Guodong Guo (Computer Sciences) at the UNIVERSITY OF WISCONSIN­MADISON 2006 #12;FACE, EXPRESSION, AND IRIS RECOGNITION, the key idea is to use learning- based methods whenever possible. For face recognition, we propose a face

Dyer, Charles R.

423

Description of Day-to-Day Variability in IRI  

NASA Astrophysics Data System (ADS)

The International Reference Ionosphere (IRI) describes the monthly average behavior of Earth's ionosphere based on most of the accessible and reliable ground and space observations of ionospheric parameters. IRI is doing an excellent job in accurately representing these average conditions as countless comparisons with additional data have shown and as acknowledged by the fact that international organizations (COSPAR, URSI, ISO, ECSS) have accepted IRI as their ionosphere standard. However, with our ever-increasing dependence on space technology it has become important to go beyond the monthly averages and to provide a description of the day-to-day variability of the ionosphere. We will review past and ongoing efforts to provide IRI users with a quantitative description of ionospheric variability depending on altitude, time of day, time of year, latitude and solar and magnetic activity. We will present new results from an analysis of ISIS and Alouette topside sounder data. The IRI team is also pursuing the development of an IRI Real-Time (IRI-RT) that uses assimilative algorithms or updating procedures to combine IRI with real-time data for a more accurate picture of current ionospheric conditions. We will review the status of these activities and report on latest results.

Bilitza, Dieter; Liu, Boding; Rodriguez, Joseph E.

2013-04-01

424

IRIS Toxicological Review of Ammonia (Revised External Review Draft)  

EPA Science Inventory

In August 2013, EPA submitted a revised draft IRIS assessment of ammonia to the agency's Science Advisory Board (SAB) and posted this draft on the IRIS website. EPA had previously released a draft of the assessment for public comment, held a public meeting about the draft, and ...

425

ASSOCIATIONS AND PRESENCE IN THE SEED CAPSULES OF IRIS HEXAGONA  

Microsoft Academic Search

Adults and larvae of the loberine erotylid beetle Loberus impressus (LeConte) were found associated with fungi growing on corolla and seed capsule tissue of the blue flag iris, Iris hexagona. We examined adult beetle specimens using light and scanning electron microscopy to determine if specialized structures (mycangia) may function in transporting fungi. Two pairs of deep pits on the ventral

PETER A. VAN ZANDT; VICTOR R. TOWNSEND; CHRISTOPHER E. CARLTON; MEREDITH BLACKWELL

426

A Novel Cryptographic Algorithm Based on Iris Feature  

Microsoft Academic Search

Biometric cryptography is a technique using biometric features to encrypt the data, which can improve the security of the encrypted data and overcome the shortcomings of the traditional cryptography. This paper proposes a novel biometric cryptographic algorithm based on the most accurate biometric feature -- iris. In this algorithm, a 256-dimension textural feature vector is extracted from the preprocessed iris

Xiukun Li; Xiangqian Wu; Ning Qi; Kuanquan Wang

2008-01-01

427

Erratum: Iris Color and Macula Pigment Optical Density  

Microsoft Academic Search

The present study was designed to assess the relationship between iris color and macular pigment optical density. Both melanin and carotenoids (responsible for iris color and macular pigment composition, respectively) appear to protect the retina through similar mechanisms and higher concentrations may reduce the incidence of retimal degenerations. The evaluate this relationship, 95 subjects were examined and the following variables

KENNETH FULD; MAX D. SNODDERLY

1996-01-01

428

IRIS Toxicological Review of Trimethylbenzenes (Revised External Review Draft)  

EPA Science Inventory

In August 2013, EPA submitted a revised draft IRIS assessment of trimethylbenzenes to the agency's Science Advisory Board (SAB) and posted this draft on the IRIS website. EPA had previously released a draft of the assessment for public comment, held a public meeting about the dr...

429

Formation of a nuclear reactor's molten core bath in a crucible-type corium catcher for a nuclear power station equipped with VVER reactors  

NASA Astrophysics Data System (ADS)

Results from a calculation study on analyzing the formation of a melt bath in a crucible-type catcher for the conditions of a severe accident at a nuclear power station equipped with VVER-1000 reactors are presented. It is shown that the heat loads exerted on the water-cooled walls of the corium catcher shell are limited to a permissible level at which the necessary margins to nucleate boiling crisis and to destruction are ensured under the conditions of thermal and mechanical loading of the shell. An important role of sacrificial material in the efficient operation of the corium catcher is pointed out.

Beshta, S. V.; Vitol', S. A.; Granovskii, V. S.; Kalyago, E. K.; Kovtunova, S. V.; Krushinov, E. V.; Sulatskaya, M. B.; Sulatskii, A. A.; Khabenskii, V. B.; Al'Myashev, V. I.; Gusarov, V. V.

2011-05-01

430

Comparison of Two IRI Plasmasphere Extensions with GPS-TEC Observations  

NASA Technical Reports Server (NTRS)

Two plasmasphere extensions of the International Reference Ionosphere are made available for the users. It is aimed to estimate the effect of charged particles on technical devices in the Earth's environment and to define the ionosphere-plasmasphere operational conditions compatible with existing and future systems of radio communication, radio navigation and other relevant radio technologies in the ranges of medium and higher frequencies. The Global Core Plasma Model (GCPM-2000) of Gallagher et al. (2000) is an empirical description of thermal plasma densities in the plasmasphere, plasmapause, magnetospheric trough, and polar cap. GCPM-2000 uses the Kp index and is coupled to IRI in the transition region 500-600 km. The IZMIRAN plasmasphere model (Chasovitin et al., 1998; Gulyaeva et al., 2002) is an empirical model based on whistler and satellite observations. It presents global vertical analytical profiles of electron density smoothly fitted to IRI electron density profile at 1000 km altitude and extended towards the plasmapause (up to 36,000 km). For the smooth fitting of the two models, the shape of the IRI topside electron density profile is improved using ISIS 1, ISIS 2, and IK19 satellite inputs (Gulyaeva, 2003). The plasmasphere model depends on solar activity and magnetic activity (kp-index). The two IRI plasmasphere extensions are compared in the present study with the total electron content derived from records of Global Positioning Satellites (GPS-TEC) observations for different latitudinal, solar activity, magnetic activity, diurnal and seasonal conditions. The differences of model TEC with observed TEC in the topside ionosphere and plasmasphere are discussed.

Gulyacva, Tamara; Gallagher, Dennis

2005-01-01

431

Neutronics and Thermal Hydraulics Study for Using a Low-Enriched Uranium Core in the Advanced Test Reactor -- 2008 Final Report  

SciTech Connect

The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuel cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis was performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff versus effective full power days (EFPDs) between the HEU and the LEU cores. The MCNP ATR 1/8th core model was used to optimize the U 235 loading in the LEU core, such that the differences in K-eff and heat flux profiles between the HEU and LEU cores were minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the ATR reference HEU case study. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm (20 mil). In this work, the proposed LEU (U-10Mo) core conversion case with nominal fuel meat thickness of 0.330 mm (13 mil) and U-235 enrichment of 19.7 wt% is used to optimize the radial heat flux profile by varying the fuel meat thickness from 0.191 mm (7.0 mil) to 0.330 mm (13.0 mil) at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). A 0.8g of Boron-10, a burnable absorber, was added in the inner and outer plates to reduce the initial excess reactivity, and the peak to average ratio of the inner/outer heat flux more effectively. Because the B-10 (n,a) reaction will produce Helium-4 (He-4), which might degrade the LEU foil type fuel performance, an alternative absorber option is proposed. The proposed LEU case study will have 6.918 g of Cadmium (Cd) mixed with the LEU at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19) as a burnable absorber to achieve peak to average ratios similar to those for the ATR reference HEU case study.

G. S. Chang; M. A. Lillo; R. G. Ambrosek

2008-06-01

432

Analysis of BDBA in RBMK-1500 reactor with long-term loss of heat removal from the core  

Microsoft Academic Search

The Ignalina nuclear power plant (NPP) is a twin-unit with two RBMK-1500, graphite moderated, boiling water, multichannel reactors. The accident management guidelines for beyond design basis accidents (BDBAs) are in a stage of preparation at Ignalina NPP. The most challenging event from BDBAs is the unavailability of water sources for heat removal from fuel channels (FCs). Due to specific design

A. Kaliatka; E. Ušpuras; M. Vaišnoras

2008-01-01

433

Facilitate, Collaborate, Educate: the Role of the IRIS Consortium in Supporting National and International Research in Seismology (Invited)  

NASA Astrophysics Data System (ADS)

Over the twenty-five years since its founding in 1984, the IRIS Consortium has contributed in fundamental ways to change the practice and culture of research in seismology in the US and worldwide. From an original founding group of twenty-two U.S. academic institutions, IRIS membership has now grown to 114 U.S. Member Institutions, 20 Educational Affiliates and 103 Foreign Affiliates. With strong support from the National Science Foundation, additional resources provided by other federal agencies, close collaboration with the U.S. Geological Survey and many international partners, the technical resources of the core IRIS programs - the Global Seismographic Network (GSN), the Program for Array Seismic Studies of the Continental Lithosphere (PASSCAL), the Data Management System (DMS) and Education and Outreach - have grown to become a major national and international source of experimental data for research on earthquakes and Earth structure, and a resource to support education and outreach to the public. While the primary operational focus of the Consortium is to develop and maintain facilities for the collection of seismological data for basic research, IRIS has become much more than an instrument facility. It has become a stimulus for collaboration between academic seismological programs and a focus for their interactions with national and international partners. It has helped establish the academic community as a significant contributor to the collection of data and an active participant in global research and monitoring. As a consortium of virtually all of the Earth science research institutions in the US, IRIS has helped coordinate the academic community in the development of new initiatives, such as EarthScope, to strengthen the support for science and argue for the relevance of seismology and its use in hazard mitigation. The early IRIS pioneers had the foresight to carefully define program goals and technical standards for the IRIS facilities that have stood the test of time. Many of these technical standards for equipment and data exchange have extended to become de-facto global standards and influenced instrument development and network practices throughout the world. A governance structure for the Consortium was created that continues to encourage strong community guidance in the operation and evolution of IRIS programs. From the outset, there was a commitment to maintaining a complete archive of all IRIS data, with significant emphasis on metadata and quality control. Building on long-standing traditions of collaboration in seismological research, an explicit IRIS policy of free and open data exchange has expanded the culture of data sharing among it members, which, through example, encouragement and support, has extended to influence the data policies of numerous other organizations in seismology and the Earth sciences. The technical revolution in sensors, communications and data collection that accompanied the early development of IRIS is now available to the entire world. The challenge for IRIS and the seismology community in the decades ahead will be to encourage the implementation of these technologies, along with appropriate training and resources, to further the research community’s endeavors to understand the structure and evolution of our planet and apply that knowledge to the mitigation of earthquake hazards.

Simpson, D. W.; Beck, S. L.

2009-12-01

434

Intraocular lens implantation for patients with coloboma of the iris  

PubMed Central

The aim of this study was to analyze the techniques for intraocular lens (IOL) implantation in patients with coloboma of the iris. A retrospective cohort study was used to analyze the degree of iris coloboma and the characteristics of the crystalline lens in 56 patients with iris coloboma. The patients with a lesser degree of coloboma of the iris and an intact lens capsule were treated by iris suture and IOL implantation into the posterior chamber. Patients with an iris coloboma confined to one quadrant, severe iris atrophy and significant lens capsule coloboma were treated with an annular suture at the edge of the pupil and IOL implantation into the anterior chamber. Patients with a greater degree of iris coloboma and an intact lens capsule were treated with an artificial iris and IOL implantation. The patients were followed up for between five months and five years after surgery. Data relating to vision, photophobia, IOL location, postoperative complications and treatment were also obtained at follow-up. The vision of the patients was improved to varying degrees following the surgery, with the exception of those with amblyopia or serious corneal scars. The photophobia of the patients had also improved. The patients’ levels of satisfaction and comfort were deemed to be satisfactory. Early postoperative complications included hyphema, increased intraocular pressure and uveitis. However, serious complications such as corneal decompensation and IOL dislocation were not observed. Various techniques for IOL implantation were selected based on the degree of iris and lens capsule coloboma; these techniques were capable of improving the vision and photophobia of the patients. PMID:24926350

LI, JUANJUAN; LI, YAN; HU, ZHULIN; KONG, LEI

2014-01-01

435

Comet whole-core solution to a stylized 3-dimensional pressurized water reactor benchmark problem with UO{sub 2}and MOX fuel  

SciTech Connect

A stylized pressurized water reactor (PWR) benchmark problem with UO{sub 2} and MOX fuel was used to test the accuracy and efficiency of the coarse mesh radiation transport (COMET) code. The benchmark problem contains 125 fuel assemblies and 44,000 fuel pins. The COMET code was used to compute the core eigenvalue and assembly and pin power distributions for three core configurations. In these calculations, a set of tensor products of orthogonal polynomials were used to expand the neutron angular phase space distribution on the interfaces between coarse meshes. The COMET calculations were compared with the Monte Carlo code MCNP reference solutions using a recently published an 8-group material cross section library. The comparison showed both the core eigenvalues and assembly and pin power distributions predicated by COMET agree very well with the MCNP reference solution if the orders of the angular flux expansion in the two spatial variables and the polar and azimuth angles on the mesh boundaries are 4, 4, 2 and 2. The mean and maximum differences in the pin fission density distribution ranged from 0.28%-0.44% and 3.0%-5.5%, all within 3-sigma uncertainty of the MCNP solution. These comparisons indicate that COMET can achieve accuracy comparable to Monte Carlo. It was also found that COMET's computational speed is 450 times faster than MCNP. (authors)

Zhang, D.; Rahnema, F. [Georgia Inst. of Technology, 770 State Street, Atlanta, GA 30332-0745 (United States)

2012-07-01

436

Pigment Melanin: Pattern for Iris Recognition  

E-print Network

Recognition of iris based on Visible Light (VL) imaging is a difficult problem because of the light reflection from the cornea. Nonetheless, pigment melanin provides a rich feature source in VL, unavailable in Near-Infrared (NIR) imaging. This is due to biological spectroscopy of eumelanin, a chemical not stimulated in NIR. In this case, a plausible solution to observe such patterns may be provided by an adaptive procedure using a variational technique on the image histogram. To describe the patterns, a shape analysis method is used to derive feature-code for each subject. An important question is how much the melanin patterns, extracted from VL, are independent of iris texture in NIR. With this question in mind, the present investigation proposes fusion of features extracted from NIR and VL to boost the recognition performance. We have collected our own database (UTIRIS) consisting of both NIR and VL images of 158 eyes of 79 individuals. This investigation demonstrates that the proposed algorithm is highly s...

Hosseini, Mahdi S; Soltanian-Zadeh, Hamid

2009-01-01

437

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

Microsoft Academic Search

The government of Mexico has expressed interest to utilize the Laguna Verde boiling water reactor (BWR) nuclear power plant for the disposition of reprocessed spent uranium oxide (UOX) fuel in the form of reactor-grade mixed-oxide (MOX) fuel. MOX fuel would replace spent UOX fuel as a fraction in the core from 18--30% depending on the fuel loading cycle. MOX fuel

Lisa Rene Vickers

2001-01-01

438

Online user-friendly slant total electron content computation from IRI-Plas: IRI-Plas-STEC  

NASA Astrophysics Data System (ADS)

Slant total electron content (STEC), the total number of free electrons on a ray path, is an important space weather observable. STEC is the main input for computerized ionospheric tomography (CIT). STEC can be estimated using the dual-frequency GPS receivers. GPS-STEC contains the space weather variability, yet the estimates are prone to measurement and instrument errors that are not related to the physical structure of the ionosphere. International Reference Ionosphere Extended to Plasmasphere (IRI-Plas) is the international standard climatic model of ionosphere and plasmasphere, providing vertical electron density profiles for a desired date, time, and location. IRI-Plas is used as the background model in CIT. Computation of STEC from IRI-Plas is a tedious task for researchers due to extensive geodetic calculations and IRI-Plas runs. In this study, IONOLAB group introduces a new space weather service to facilitate the computation of STEC from IRI-Plas (IRI-Plas-STEC) at www.ionolab.org. The IRI-Plas-STEC can be computed online for a desired location, date, hour, elevation, and azimuth angle. The user-friendly interface also provides means for computation of IRI-STEC for a desired location and date to indicate the variability in hour of the day, elevation, or azimuth angles. The desired location can be chosen as a GPS receiver in International GNSS Service (IGS) or EUREF Permanent Network (EPN). Also instead of specifying elevation and azimuth angles, the user can directly choose from the GPS satellites and obtain IRI-Plas-STEC for a desired date and/or hour. The computed IRI-Plas-STEC values are presented directly on the screen or via e-mail as both text and plots.

Tuna, Hakan; Arikan, Orhan; Arikan, Feza; Gulyaeva, Tamara L.; Sezen, Umut

2014-01-01

439

DESIGN STUDY FOR A LOW-ENRICHED URANIUM CORE FOR THE HIGH FLUX ISOTOPE REACTOR, ANNUAL REPORT FOR FY 2010  

SciTech Connect

This report documents progress made during FY 2010 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in