These are representative sample records from Science.gov related to your search topic.
For comprehensive and current results, perform a real-time search at Science.gov.
1

The IRIS Spool-Type Reactor Coolant Pump  

Microsoft Academic Search

IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the major reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This

J. M. Kujawski; D. M. Kitch; L. E. Conway

2002-01-01

2

The IRIS General Plant Arrangement  

Microsoft Academic Search

IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the major reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This

J. Robertson; J. Love; R. Morgan; L. E. Conway

2002-01-01

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

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

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

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

9

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

SciTech Connect

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

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

2011-05-31

10

Steam Generator of the International Reactor Innovative and Secure  

Microsoft Academic Search

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

L. Cinotti; M. Bruzzone; N. Meda; G. Corsini; C. V. Lombardi; M. Ricotti; L. E. Conway

2002-01-01

11

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

12

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

13

The Economics of IRIS  

Microsoft Academic Search

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

K. Miller; D. Paramonov

2002-01-01

14

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

Microsoft Academic Search

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

Mitsuru KAMBE

2002-01-01

15

Gas-core reactor power transient analysis  

NASA Technical Reports Server (NTRS)

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

Kascak, A. F.

1972-01-01

16

Reactor pulse repeatability studies at the annular core research reactor  

SciTech Connect

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

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

2011-07-01

17

An efficient computational technique for light water reactor core dynamics  

Microsoft Academic Search

By combining a modified version of the so-called ''adiabatic'' method for reactor dynamic calculations with a simplified flow redistribution scheme, an efficient method for predicting three-dimensional core behavior has been developed for pressurized water reactor transients. Both the simplified core reactivity and the flow redistribution calculations are shown to yield close approximations of the results obtained by more rigorous approaches.

C. D. Wu; J. Weisman

1988-01-01

18

Support structure core assembly in a nuclear reactor  

Microsoft Academic Search

A nuclear reactor of the type including a reactor vessel and a core assembly to be maintained in a fixed position within the vessel is disclosed herein along with a structural arrangement also located in the vessel, for supporting the core assembly in its fixed position. The structural arrangement disclosed utilizes a plurality of components including a grillage of i-beams

C. F. Dupen; R. F. Raymond

1982-01-01

19

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

20

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

21

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

22

Advanced High Temperature Reactor Neutronic Core Design  

SciTech Connect

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

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

2012-01-01

23

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

24

State space modeling of reactor core in a pressurized water reactor  

SciTech Connect

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.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

2014-07-10

25

77 FR 30435 - In-core Thermocouples at Different Elevations and Radial Positions in Reactor Core  

Federal Register 2010, 2011, 2012, 2013, 2014

...Recommendations for Enhancing Reactor Safety in the 21st Century...Recommendations for Enhancing Reactor Safety in the 21st Century...typically cover accidents to the point of loss of core cooling and...Recommendations for Enhancing Reactor Safety in the 21st...

2012-05-23

26

Reactor core design of Gas Turbine High Temperature Reactor 300  

Microsoft Academic Search

Japan Atomic Energy Research Institute (JAERI) has been designing Japan’s original gas turbine high temperature reactor, Gas Turbine High Temperature Reactor 300 (GTHTR300). The greatly simplified design based on salient features of the High Temperature Gas-cooled Reactor (HTGR) with a closed helium gas turbine enables the GTHTR300 a highly efficient and economically competitive reactor to be deployed in early 2010s.

Kazuhiko Kunitomi; Shoji Katanishi; Shoji Takada; Xing Yan; Nobumasa Tsuji

2004-01-01

27

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

28

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

29

CORE HEAT TRANSFER IN LOCKHEED RADIATION EFFECTS REACTOR AND SHIELD DEVELOPMENT REACTOR  

Microsoft Academic Search

Solutions to the heat transfer problems are presented of the core ; components of the Radiation Effects Reactor. These components include the fuel ; elements, control rods, regulating rod, dummy fuel elements, beryllium reflector ; pieces, and start-up source. The results of the heat transfer in the fuel ; elements of the Shield Development Reactor are also presented. The core

R. O. Niemi; K. M. Horst

1957-01-01

30

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

31

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

32

Core design of the upgraded TREAT reactor  

SciTech Connect

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

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

1982-01-01

33

Unsteady Characteristics of Three-Core Molten Salt Reactor  

NASA Astrophysics Data System (ADS)

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

Yamamoto, Takahisa; Mitachi, Koshi; Nishio, Masatoshi

34

Core Physics and Kinetics Calculations for the Fissioning Plasma Core Reactor  

NASA Technical Reports Server (NTRS)

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

Butler, C.; Albright, D.

2007-01-01

35

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

36

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

37

The Reactor Core Neutronic model for the Pebble Bed Modular Reactor  

Microsoft Academic Search

This paper describes the technical aspects of the Reactor Core Neutronic model for the Pebble Bed Modular Reactor (PBMR). Included is a model design review with preliminary simulation results and model constraints. The PBMR Demonstration Power Plant is a First of a Kind Engineering plant which will be used for the production and generation of electricity in South Africa.The theory

Trevor Dudley; Piet de Villiers; Werner Bouwer; Oliver Tsaoi; Eben Mulder

2008-01-01

38

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

NASA Technical Reports Server (NTRS)

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

Bragg-Sitton, Shannon M.

2005-01-01

39

An economic optimization of pressurized light water reactor cores  

NASA Astrophysics Data System (ADS)

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

Pfeifer, Holger

40

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

41

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

42

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

43

Core design of supercritical-pressure light water reactor  

SciTech Connect

A large improvement in thermal efficiency can be achieved in a Light Water Reactor (LWR) by using the supercritical-pressure thermal cycle concept. No boiling or phase transition occurs above the critical pressure condition, which leads to a substantial simplification of the system by eliminating the steam separator, the dryers and the steam Generators in a direct cycle. In this feasibility study, the Supercritical-Pressure Light Water Reactor (SCLWR) was analyzed with reference to the experience of conventional LWRs. The main results of fuel, the core, the control rod, the core internals and the reactor vessel designs are presented in this report. The thermal hydraulic design which uses an accurate flow distribution to achieve a high thermal efficiency has been also analyzed. Finally, based on the present results, recommendations have been made for future work.

Tanaka, S.; Shirai, Y.; Mori, M. [Tokyo Electric Co., Yokohama (Japan). Nuclear Power R and D Center] [and others

1996-07-01

44

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

45

Gamma thermometer based reactor core liquid level detector  

SciTech Connect

A system is provided which employs a modified gamma thermometer for determining the liquid coolant level within a nuclear reactor core. The gamma thermometer which normally is employed to monitor local core heat generation rate (reactor power), is 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

Gamma thermometer based reactor core liquid level detector  

SciTech Connect

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

Burns, T.J.

1983-09-20

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

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

50

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

51

Specific Mass Estimates for A Vapor Core Reactor With MHD  

SciTech Connect

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 (UF{sub 4}) vapor as the fissioning fuel and alkali metals or their fluorides as working fluid in a closed Rankine cycle with MHD energy conversion. Gaseous and liquid-vapor core reactors can potentially provide the highest reactor and cycle temperature among all existing or proposed fission reactor designs. This unique feature makes this reactor concept a very natural and attractive candidate for very high power (10 to 1000 MWe) and low specific mass (0.4 to 5 kg/kWe) nuclear electric propulsion (NEP) applications since the MHD output could be coupled with minimal power conditioning to MPD thrusters or other types of thruster for producing thrust at very high specific impulse (I{sub sp} 1500 to 10,000 s). The exceptional specific mass performance of an optimized VCRMHD- NEP system could lead to a dramatic reduction in the cost and duration of manned or robotic interplanetary as well as interstellar missions. The VCR-MHD-NEP system could enable very efficient Mars cargo transfers or short (<8 month) Mars round trips with less initial mass in low Earth orbit (IMLEO). The system could also enable highly efficient lunar cargo transfer and rapid missions to other destinations throughout the solar system. (authors)

Knight, Travis; Smith, Blair; 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

52

Superconducting shielded core reactor with reduced AC losses  

DOEpatents

A superconducting shielded core reactor (SSCR) operates as a passive device for limiting excessive AC current in a circuit operating at a high power level under a fault condition such as shorting. The SSCR includes a ferromagnetic core which may be either closed or open (with an air gap) and extends into and through a superconducting tube or superconducting rings arranged in a stacked array. First and second series connected copper coils each disposed about a portion of the iron core are connected to the circuit to be protected and are respectively wound inside and outside of the superconducting tube or rings. A large impedance is inserted into the circuit by the core when the shielding capability of the superconducting arrangement is exceeded by the applied magnetic field generated by the two coils under a fault condition to limit the AC current in the circuit. The proposed SSCR also affords reduced AC loss compared to conventional SSCRs under continuous normal operation.

Cha, Yung S.; Hull, John R.

2006-04-04

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

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

55

Experimental Criticality Benchmarks for SNAP 10A/2 Reactor Cores  

SciTech Connect

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

Krass, A.W.

2005-12-19

56

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

57

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

SciTech Connect

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

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

2010-12-23

58

Piezoelectric material for use in a nuclear reactor core  

NASA Astrophysics Data System (ADS)

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

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

2012-05-01

59

Piezoelectric material for use in a nuclear reactor core  

SciTech Connect

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

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

2012-05-17

60

RMC - A Monte Carlo Code for Reactor Core Analysis  

NASA Astrophysics Data System (ADS)

A new Monte Carlo transport code RMC has been being developed by Department of Engineering Physics, Tsinghua University, Beijing as a tool for reactor core analysis on high-performance computing platforms. To meet the requirements of reactor analysis, RMC now has such functions as criticality calculation, fixed-source calculation, burnup calculation and kinetics simulations. Some techniques for geometry treatment, new burnup algorithm, source convergence acceleration, massive tally and parallel calculation, and temperature dependent cross sections processing are researched and implemented in RMC to improve the effciency. Validation results of criticality calculation, burnup calculation, source convergence acceleration, tallies performance and parallel performance shown in this paper prove the capabilities of RMC in dealing with reactor analysis problems with good performances.

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

2014-06-01

61

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

62

Coupled simulation of the reactor core using CUPID/MASTER  

SciTech Connect

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

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

2012-07-01

63

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

64

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

65

A solid reactor core thermal model for nuclear thermal rockets  

NASA Astrophysics Data System (ADS)

A Helium/Hydrogen Cooled Reactor Analysis (HERA) computer code has been developed. HERA has the ability to model arbitrary geometries in three dimensions, which allows the user to easily analyze reactor cores constructed of prismatic graphite elements. The code accounts for heat generation in the fuel, control rods, and other structures; conduction and radiation across gaps; convection to the coolant; and a variety of boundary conditions. The numerical solution scheme has been optimized for vector computers, making long transient analyses economical. Time integration is either explicit or implicit, which allows the use of the model to accurately calculate both short- or long-term transients with an efficient use of computer time. Both the basic spatial and temporal integration schemes have been benchmarked against analytical solutions.

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

1991-01-01

66

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

67

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

68

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

SciTech Connect

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

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

1988-06-01

69

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

70

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

71

SCDAP: a computer code for analyzing light-water-reactor severe core damage  

Microsoft Academic Search

The Severe Core Damage Analysis Package (SCDAP) computer code is being developed at the Idaho National Engineering Laboratory under the sponsorship of the US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research. SCDAP models the progression of light water reactor core damage including core heatup, core disruption, debris formation, debris heatup, and debris melting. SCDAP is being used to help

C. M. Allison; E. R. Carlson; R. H. Smith

1983-01-01

72

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

73

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

74

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

75

Core reactivity estimation in space reactors using recurrent dynamic networks  

SciTech Connect

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

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

1991-01-10

76

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

77

Three core concepts for producing uranium-233 in commercial pressurized light water reactors for possible use in water-cooled breeder reactors  

Microsoft Academic Search

Selected prebreeder core concepts are described which could be backfit into a reference light water reactor similar to current commercial reactors, and produce uranium-233 for use in water-cooled breeder reactors. The prebreeder concepts were selected on the basis of minimizing fuel system development and reactor changes required to permit a backfit. The fuel assemblies for the prebreeder core concepts discussed

G. H. Conley; G. K. Cowell; C. A. Detrick; J. Kusenko; E. G. Johnson; J. Dunyak; B. K. Flanery; M. S. Shinko; R. H. Giffen; D. S. Rampolla

1979-01-01

78

MEASUREMENT OF AIR BLAST EFFECTS FROM SIMULATED NUCLEAR REACTOR CORE EXCURSIONS  

Microsoft Academic Search

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

R. J. Larson; W. C. Olson

1957-01-01

79

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

80

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

81

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

NASA Technical Reports Server (NTRS)

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

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

1974-01-01

82

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

83

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

84

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

85

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

86

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

87

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

88

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

89

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

90

Accident analyses and passive measures reducing the consequences of a core-melt in CAPRA\\/CADRA reactor cores  

Microsoft Academic Search

As part of the Combustion Améliorée du Plutonium dans les Réacteurs Avancés\\/Consommation D'Actinides et de Déchets dans les Réacteurs Avancés (CAPRA\\/CADRA) program the feasibility of reactor systems with different neutron spectra and coolants is investigated to burn plutonium and also to destruct minor actinides and long lived fission products. In this paper, we deal with reactor cores with fast spectrum

W Maschek; D Struwe

2000-01-01

91

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

92

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

SciTech Connect

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

James W. Sterbentz

2008-05-01

93

MELCOR adaptation and validation for modeling of N Reactor core phenomena  

SciTech Connect

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

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

1990-01-01

94

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

95

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

E-print Network

CONFIGURATION ADJUSTMENT POTENTIAL OF THE VERY HIGH TEMPERATURE REACTOR PRISMATIC CORES WITH ADVANCED ACTINIDE FUELS A Thesis by DAVID E. AMES II Submitted to the Office of Graduate Studies of Texas A... REACTOR PRISMATIC CORES WITH ADVANCED ACTINIDE FUELS A Thesis by DAVID E. AMES II Submitted to the Office of Graduate Studies of Texas A&M University in partial fulfillment of the requirements for the degree of MASTER...

Ames, David E, II

2006-10-30

96

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

97

A Novel Iris Recognition System  

Microsoft Academic Search

Iris recognition as an emerging biometric recognition approach and itpsilas becoming a very active topic in both research and practical applications. The pattern of the human iris differs from person to person, even between monocular twins. This paper proposes a modified iris localization method and normalization method. In the iris localization after invert the iris image edges are detected by

V. C. Subbarayudu; Munaga V N K Prasad

2007-01-01

98

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

99

Operating experience of natural circulation core cooling in boiling water reactors  

Microsoft Academic Search

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

C. Kullberg; K. Jones; C. Heath

1993-01-01

100

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

Microsoft Academic Search

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

T. Schulenberg; J. Starflinger; J. Heinecke

2008-01-01

101

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

Microsoft Academic Search

As part of the update of the Safety analysis Report (SAR) for the Annular Core Research Reactor (ACRR), operational limiting events under the category of inadvertent withdrawal of an experiment while at power or during a power pulse were determined to be the most limiting event(s) for this reactor. This report provides a summary of the assumptions, modeling, and results

Restrepo

1992-01-01

102

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

SciTech Connect

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

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

2006-07-01

103

2006 IRIS 5-YEAR PROPOSAL COREMANTLE BOUNDARY Examining the Base of the Mantle Using IRIS FLED PASSCAL Data  

E-print Network

a region of the lowermost mantle well-modeled by a velocity discontinuity at the top of D''in the manner of Solomatov and Moresi. C.Region of the lowermost mantle that shows no evidence of a D''discontinuity179 2006 IRIS 5-YEAR PROPOSAL CORE­MANTLE BOUNDARY Examining the Base of the Mantle Using IRIS FLED

Wysession, Michael E.

104

Benchmarking of the FIBWR2 transient BWR (boiling water reactor) core hydraulics code  

Microsoft Academic Search

The FIBWR code has been widely used for boiling water reactor (BWR) steady-state core flow, pressure, and void distribution calculations. The code models the complex flow paths in a BWR core, including water tubes and leakage flow to the bypass. FIBWR results are used to calculate flow-weighted equivalent loss coefficients for geometrically more simplified codes such as SIMULATE and RETRAN.

B. J. Gitnick; D. A. Prelewicz

1990-01-01

105

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

Microsoft Academic Search

This is a lunar nuclear power plant with the advantages of minimum mass, with no moving parts, no pumped liquid coolant, a solid metal rugged core, with no single point of failure. The electrical output is 100 kilowatts with a 500 kilowatt thermal reactor. The thermoelectric converters surround the potassium heatpipes from the core and water heatpipes surround the converter

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

2008-01-01

106

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

107

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...

108

Fission product heating effects after shutdown of the Scottish Universities Research Reactor with simulated bottom blockage of the core tanks  

Microsoft Academic Search

The investigation described in this report was carried out using the Scottish Universities' Research Reactor at East Kilbride, Scotland. Reactor tests have been performed in order to investigate, after reactor shutdown, the fuel plate temperature behavior under a simulated bottom blockage of the core tanks. This simulated blockage was obtained by either returning water to partially fill the core tanks,

J. A. Riley; S. R. Donald

1972-01-01

109

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

110

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

111

Numerical simulation of a Hypothetical Core Disruptive Accident in a small-scale model of a nuclear reactor  

Microsoft Academic Search

In the case of a Hypothetical Core Disruptive Accident (HCDA) in a Liquid Metal Fast Breeder Reactor, it is assumed that the core of the nuclear reactor has melted partially and that the chemical interaction between molten fuel and liquid sodium has created a high-pressure gas bubble in the core. The violent expansion of this bubble loads and deforms the

M. F. Robbe; M. Lepareux; E. Treille; Y. Cariou

2003-01-01

112

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

NASA Technical Reports Server (NTRS)

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

Putre, H. A.

1973-01-01

113

Optimization of hydride fueled pressurized water reactor cores  

E-print Network

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

Shuffler, Carter Alexander

2004-01-01

114

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

115

IRIS Process (Pre-2004)  

EPA Science Inventory

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

116

IRIS Process (2008)  

EPA Science Inventory

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

117

IRIS Process (2009 Update)  

EPA Science Inventory

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

118

Fast reactor 3D core and burnup analysis using VESTA  

SciTech Connect

Burnup analyses using the VESTA code have been performed on a MOX-fuelled fast reactor model as specified by an IAEA computational benchmark. VESTA is a relatively new code that has been used for burnup credit calculations and thermal reactor models, but not typically for fast reactor applications. The detailed input and results of the IAEA benchmark provides an opportunity to gauge the use of VESTA in a fast reactor application. VESTA employs an ultra-fine multi-group binning approach that accelerates Monte Carlo burnup calculations. Using VESTA to compute the end of cycle (EOC) power fractions by enrichment zone showed agreement with the published values within 5%. When comparing the ultra-fine multi-group binning approach to the tally-based approach, EOC isotopic masses also agree within 5%. Using the ultra-fine multi-group binning approach, we obtain a wall-time speedup factor of 35 when compared to the tally-based approach for computing a k{sub eff} eigenvalue with burnup problem. The authors conclude the use of VESTA's ultra-fine multi-group binning approach with Monte Carlo transport performs accurate depletion calculations for this fast reactor benchmark. (authors)

Luciano, N.; Shamblin, J.; Maldonado, I. [Nuclear Engineering Dept., Univ. of Tennessee, Knoxville, TN 37996-2300 (United States)

2012-07-01

119

Mechanical design of core components for a high performance light water reactor with a three pass core  

SciTech Connect

Nuclear reactors using supercritical water as coolant can achieve more than 500 deg. C core outlet temperature, if the coolant is heated up in three steps with intermediate mixing to avoid hot streaks. This method reduces the peak cladding temperatures significantly compared with a single heat up. The paper presents an innovative mechanical design which has been developed recently for such a High Performance Light Water Reactor. The core is built with square assemblies of 40 fuel pins each, using wire wraps as grid spacers. Nine of these assemblies are combined to a cluster having a common head piece and a common foot piece. A downward flow of additional moderator water, separated from the coolant, is provided in gaps between the assemblies and in a water box inside each assembly. The cluster head and foot pieces and mixing chambers, which are key components for this design, are explained in detail. (authors)

Fischer, Kai [EnBW Kernkraft GmbH, Kernkraftwerk Philippsburg, D-76661 Philippsburg (Germany); Schneider, Tobias; Redon, Thomas [University of Karlsruhe, 76133 Karlsruhe (Germany); Schulenberg, Thomas; Starflinger, Joerg [Forschungszentrum Karlsruhe, Institute for Nuclear and Energy Technologies, 76021 Karlsruhe (Germany)

2007-07-01

120

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

NASA Technical Reports Server (NTRS)

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

Martin, James J.; Reid, Robert S.

2004-01-01

121

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

122

Optimization of the Materials Composition in External Core Catchers for Nuclear Reactors  

Microsoft Academic Search

Existing schemes of core melt retention apparatus for water-cooled water-moderated nuclear reactors are analyzed. In-shaft variants of melt catchers at nuclear power plants with VVÉR-1000 reactors are proposed. It is shown that TiO2- and Nd2O3-based materials increase the operational reliability of the retention apparatus by modifying the processes occurring in the melt and by preserving the integrity of refractory coatings

V. N. Mineev; F. A. Akopov; A. S. Vlasov; Yu. A. Zeigarnik; O. M. Traktuev

2002-01-01

123

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

Microsoft Academic Search

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

Evaldas Bubelis; Algirdas Kaliatka; Eugenijus Uspuras

2002-01-01

124

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

SciTech Connect

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

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

2006-01-20

125

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

126

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

127

Circular dielectric liquid iris.  

PubMed

We demonstrate a liquid iris diaphragm using dielectric force, enabling its aperture to vary from 4 mm at the resting state to 1.5 mm at 160 V(rms). The liquid iris is a packaged optical component comprised of transparent oil, opaque ink, and a set of driving electrodes on a glass substrate. The iris aperture shrinks with the dielectric force, which is exerted on the interface between the two nonconductive liquids. The transmittance was measured to exceed 85% with no antireflection coatings over the spectrum of visible light. The maximum electric power consumed is measured to be 5.7 mW. PMID:20634871

Tsai, C Gary; Yeh, J Andrew

2010-07-15

128

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

NASA Technical Reports Server (NTRS)

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

Rom, F. E.

1969-01-01

129

Development of small, fast reactor core designs using lead-based coolant.  

SciTech Connect

A variety of small (100 MWe) fast reactor core designs are developed, these include compact configurations, long-lived (15-year fuel lifetime) cores, and derated, natural circulation designs. Trade studies are described which identify key core design issues for lead-based coolant systems. Performance parameters and reactivity feedback coefficients are compared for lead-bismuth eutectic (LBE) and sodium-cooled cores of consistent design. The results of these studies indicate that the superior neutron reflection capability of lead alloys reduces the enrichment and burnup swing compared to conventional sodium-cooled systems; however, the discharge fluence is significantly increased. The size requirement for long-lived systems is constrained by reactivity loss considerations, not fuel burnup or fluence limits. The derated lead-alloy cooled natural circulation cores require a core volume roughly eight times greater than conventional compact systems. In general, reactivity coefficients important for passive safety performance are less favorable for the larger, derated configurations.

Cahalan, J. E.; Hill, R. N.; Khalil, H. S.; Wade, D. C.

1999-06-11

130

Nodal Diffusion Burnable Poison Treatment for Prismatic Reactor Cores  

SciTech Connect

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

A. M. Ougouag; R. M. Ferrer

2010-10-01

131

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

NASA Technical Reports Server (NTRS)

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

Whitmarsh, C. L., Jr.

1972-01-01

132

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

133

Internetwork chromospheric bright grains observed with IRIS  

E-print Network

The Interface Region Imaging Spectrograph (IRIS) reveals small-scale rapid brightenings in the form of bright grains all over coronal holes and the quiet sun. These bright grains are seen with the IRIS 1330 \\AA, 1400 \\AA\\ and 2796 \\AA\\ slit-jaw filters. We combine coordinated observations with IRIS and from the ground with the Swedish 1-m Solar Telescope (SST) which allows us to have chromospheric (Ca II 8542 \\AA, Ca II H 3968 \\AA, H\\alpha, and Mg II k 2796 \\AA), and transition region (C II 1334 \\AA, Si IV 1402) spectral imaging, and single-wavelength Stokes maps in Fe I 6302 \\AA at high spatial (0.33"), temporal and spectral resolution. We conclude that the IRIS slit-jaw grains are the counterpart of so-called acoustic grains, i.e., resulting from chromospheric acoustic waves in a non-magnetic environment. We compare slit-jaw images with spectra from the IRIS spectrograph. We conclude that the grain intensity in the 2796 \\AA\\ slit-jaw filter comes from both the Mg II k core and wings. The signal in the C II ...

Martínez-Sykora, Juan; Carlsson, Mats; De Pontieu, Bart; Pereira, Tiago M D; Boerner, Paul; Hurlburt, Neal; Kleint, Lucia; Lemen, James; Tarbell, Ted D; Title, Alan; Wuelser, Jean-Pierre; Hansteen, Viggo H; Golub, Leon; McKillop, Sean; Reeves, Kathy K; Saar, Steven; Testa, Paola; Tian, Hui; Jaeggli, Sarah; Kankelborg, Charles

2015-01-01

134

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

135

Gamma heating in reflector heat shield of gas core reactor  

NASA Technical Reports Server (NTRS)

Heating rate measurements made in a mock-up of a BeO heat shield for a gas core nuclear rocket engine yields results nominally a factor of two greater than calculated by two different methods. The disparity is thought to be caused by errors in neutron capture cross sections and gamma spectra from the low cross-section elements, D, O, and Be.

Lofthouse, J. H.; Kunze, J. F.; Young, T. E.; Young, R. C.

1972-01-01

136

Monitoring of Local Perturbations of the Multiplication Factor on Overloads in the Core of a Shut?Down Reactor  

Microsoft Academic Search

A method of revealing the fact of existence (absence) of a local region in the reactor core with an anomalous value of the multiplication factor, detection of this region, and on-line monitoring of its “movement” throughout the volume of the core of a shut-down reactor during the entire period of overloading is proposed.

V. V. Shidlovskii

2001-01-01

137

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

SciTech Connect

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

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

1994-02-01

138

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

139

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

140

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

141

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

142

Depressurized core heatup accident scenarios in advanced modular high temperature gas cooled reactors  

Microsoft Academic Search

The decay heat removal by a passive air colling system from a modular high tempeature gas cooled reactor during depressurized core heatup accident scenarios was analyzed. The effects of several design and operating parameters on the peak fuel and vessel temperatures were established. The results indicate that fuel and vessel temperatures remain well below failure levels and that significant safety

Kroeger

1988-01-01

143

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

144

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

145

Expert opinion and statistical evidence: an application to reactor core melt frequency  

Microsoft Academic Search

A model for the evaluation of probabilities of rare events by combining the available experience with expert opinion is developed, using the core melt frequency of nuclear power reactors as an example. A distribution for this frequency is assessed using the statistical evidence and including near misses. this distribution is subsequently modified, via Bayes' theorem, to include the estimate derived

G. Apostolakis; A. Mosleh

1979-01-01

146

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

147

A chemical equilibrium estimate of the aerosols produced in an overheated light water reactor core  

Microsoft Academic Search

The degree of vaporization of light water reactor core materials was estimated using a highly idealized procedure involving (a) specification of the phases that are present for both structural and fuel material, (b) estimation of the vapor pressures exerted by the individual components of each phase, and (c) assuming a degree of vaporization of each phase constituent, allowing equilibration between

R. P. Wichner; R. D. Spence

1985-01-01

148

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

Microsoft Academic Search

A new method of In-Core Monitor Housing replacement has been successfully applied to Tokai Unit 2 (BWR with 1100 MWe) in April of 2001. It was designed to replace a housing under dry condition of reactor pressure vessel (RPV): this enabled the elimination of water filled-up and drained processes during the replacement procedure resulting in the reduction of implementation schedule.

Tatsuo Ishida; Shoji Yamamoto; Fujitoshi Eguchi; Motomasa Fuse; Kouichi Kurosawa; Sadato Shimizu; Minoru Masuda; Shinya Fujii; Junji Tanaka; Bryce A. Jacobson

2002-01-01

149

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

SciTech Connect

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

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

2006-07-01

150

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

151

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

SciTech Connect

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

Burns, T.J.

1981-06-16

152

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

153

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

154

Shippingport operations with the Light Water Breeder Reactor core. (LWBR Development Program)  

SciTech Connect

This report describes the operation of the Shippingport Atomic Power Station during the LWBR (Light Water Breeder Reactor) Core lifetime. It also summarizes the plant-oriented operations during the period preceding LWBR startup, which include the defueling of The Pressurized Water Reactor Core 2 (PWR-2) and the installation of the LWBR Core, and the operations associated with the defueling of LWBR. The intent of this report is to examine LWBR experience in retrospect and present pertinent and significant aspects of LWBR operations that relate primarily to the nuclear portion of the Station. The nonnuclear portion of the Station is discussed only as it relates to overall plant operation or to unusual problems which result from the use of conventional equipment in radioactive environments. 30 refs., 69 figs., 27 tabs.

Budd, W.A. (ed.)

1986-03-01

155

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

156

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

157

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

158

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

159

Assessment of creep-fatigue damage at main vessel triple point of Prototype Fast Breeder Reactor after core disruptive accident  

Microsoft Academic Search

The main vessel (MV) carries all the major reactor components including core and liquid sodium filled to 12.4 m height. The\\u000a reactor core along with the inner vessel is placed on the grid plate, which is supported on the core support structure (CSS).\\u000a The CSS is supported on the MV through a support shell, welded to the dished end at

Bhuwan Chandra Sati; R. Srinivasan; P. Selvaraj; P. Chellapandi

2010-01-01

160

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

NASA Astrophysics Data System (ADS)

An Accelerator Driven System (ADS) consists in the coupling of an accelerator with a nuclear reactor. Such systems will need neutron detectors working in a wide energy range and insensitive to X- and ?-rays. Micromegas technology has been proposed to achieve this goal. The ability of Micromegas to detect neutrons over a wide energy range has already been demonstrated and this detector is, under certain conditions, insensitive to ?-rays. A new Micromegas neutron detector called Piccolo Micromegas has been designed to get integrated neutron fluxes on different energy domains. For the first time, Piccolo Micromegas has been placed in the core of a nuclear reactor at Casaccia in Italy. The configuration of the detector will be presented as well as its functioning and the reasons of its insensitivity to ?-rays. The results of the operation of the detector will also be shown for low reactor power to high reactor power and some improvements will be suggested.

Pancin, J.; Andriamonje, S.; Aune, S.; Giganon, A.; Giomataris, Y.; Lecolley, J. F.; Riallot, M.; Rosa, R.

2008-07-01

161

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

NASA Technical Reports Server (NTRS)

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

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

1971-01-01

162

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

163

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

164

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

165

Advanced core design and fuel management for pebble-bed reactors  

NASA Astrophysics Data System (ADS)

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.

Gougar, Hans David

166

Progress in the neutronic core conversion (HEU-LEU) analysis of Ghana research reactor-1.  

SciTech Connect

The Ghana Research Reactor-1 (GHARR-1) is a commercial version of the Miniature Neutron Source Reactor (MNSR) and has operated at different power levels since its commissioning in March 1995. As required for all nuclear reactors, neutronic and thermal hydraulic analysis are being performed for the HEU-LEU core conversion studies of the Ghana Research Reactor-1 (GHARR-1) facility, which is a commercial version of the Miniature Neutron Source Reactor (MNSR). Stochastic Monte Carlo particle transport methods and tools (MCNP4c/MCNP5) were used to fine-tune a previously developed 3-D MCNP model of the GHARR-1 facility and perform neutronic analysis of the 90.2% HEU reference and candidate LEU (UO{sub 2}, U{sub 3}Si{sub 2}, U-9Mo) fresh cores with varying enrichments from 12.6%-19.75%. In this paper, the results of the progress made in the Monte Carlo neutronic analysis of the HEU reference and candidate LEU fuels are presented. In particular, a comparative performance assessment of the LEU with respect to neutron flux variations in the fission chamber and experimental irradiation channels are highlighted.

Anim-Sampong, S.; Maakuu, B. T.; Akaho, E. H. K.; Andam, A.; Liaw, J. J. R.; Matos, J. E.; Nuclear Engineering Division; Ghana Atomic Energy Commission; Kwame Nkrumah Univ. of Science and Technology

2006-01-01

167

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

SciTech Connect

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

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

2006-07-01

168

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

SciTech Connect

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

Samim Anghaie

2002-08-13

169

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

using the multi-point reactor kinetics equations to accommodate the modular core configura- tion 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.

170

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

171

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

SciTech Connect

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

Lee, Bom Soon

1994-06-01

172

Lunar in-core thermionic nuclear reactor power system conceptual design  

NASA Technical Reports Server (NTRS)

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

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

1991-01-01

173

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

174

US neutron effects work under SUBWOG-6N. [Sandia pulse reactor and annular core pulse reactor  

Microsoft Academic Search

the scope of S-6N is to study the effects of neutrons on weapons materials and structures, particularly fissionable ones, but excluding electronic components. Included in the scope are: development of laboratory neutron sources and technology; development of reactor and experiment diagnostics; the study of materials and structural responses under dynamic neutron heating; and the study of the combined effects of

T. R. Schmidt; J. W. Bryson; T. F. Luera

1987-01-01

175

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

SciTech Connect

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

Richard Schultz

2012-04-01

176

Adaptive liquid crystal iris  

NASA Astrophysics Data System (ADS)

We report an adaptive iris using a twisted nematic liquid crystal (TNLC) and a hole-patterned electrode. When an external voltage is applied to the TNLC, the directors of the LC near the edge of the hole are unwound first. Increasing the voltage can continuously unwind the LC toward the center. When the TNLC is sandwiched between two polarizers, it exhibits an iris-like character. Either a normal mode or a reverse mode can be obtained depending on the orientations of the transmission axes of the two polarizers. In contrast to liquid irises, the aperture of the LC iris can be closed completely. Moreover, it has the advantages of large variability of the aperture diameter, good stability, and low power consumption. Applications of the device for controlling the laser energy and correcting optical aberration are foreseeable.

Zhou, Zuowei; Ren, Hongwen; Nah, Changwoon

2014-09-01

177

Method and apparatus for use to exchange o-ring interposed between in-core housing and in-core flange in nuclear reactor  

Microsoft Academic Search

This invention concerns a method and apparatus for use to exchange an o-ring interposed between an in-core housing disposed at the lower end of a guide tube for an in-core monitor for monitoring the operation of a nuclear reactor and an in-core flange attached to the housing. The apparatus comprises a drain system adapted to be connected to the lower

T. Tsuji; S. Watanabe

1981-01-01

178

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

NASA Technical Reports Server (NTRS)

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

Kazeminezhad, F.; Anghai, S.

2008-01-01

179

Size Distribution of NaK Droplets Released During Rorsat Reactor Core Ejection  

Microsoft Academic Search

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

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

2004-01-01

180

A Best-Estimate Reactor Core Monitor Using State Feedback Strategies to Reduce Uncertainties  

Microsoft Academic Search

The development and demonstration of a new algorithm to reduce modeling and state-estimation uncertainty in best-estimate simulation codes has been investigated. Demonstration is given by way of a prototype reactor core monitor. The architecture of this monitor integrates a control-theory-based, distributed-parameter estimation technique into a production-grade best-estimate simulation code. The Kalman Filter-Sequential Least-Squares (KFSLS) parameter estimation algorithm has been extended

Robert P. Martin; Robert M. Edwards

2000-01-01

181

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

182

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

SciTech Connect

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

Restrepo, L F

1992-08-01

183

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

184

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

SciTech Connect

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

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

2011-07-01

185

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

SciTech Connect

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

John D. Bess

2014-12-01

186

Using crypts as iris minutiae  

NASA Astrophysics Data System (ADS)

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

Shen, Feng; Flynn, Patrick J.

2013-05-01

187

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

188

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

189

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

190

Fisher Iris Data  

NSDL National Science Digital Library

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

Ethel Stanley (Beloit College; Biology)

2009-01-10

191

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

192

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

SciTech Connect

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

Hassan, Yassin

2013-10-22

193

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

194

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

195

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

SciTech Connect

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

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

2004-07-01

196

Fuel Design and Core Layout for a Gas-Cooled Fast Reactor  

SciTech Connect

The gas-cooled fast reactor (GCFR) is regarded as the primary candidate for a future sustainable nuclear power system. In this paper a general core layout is presented for a 2400-MW(thermal) GCFR. Two fuel elements are discussed: a TRISO-based coated particle and the innovative hollow sphere concept. Sustainability calls for recycling of all minor actinides (MAs) in the core and a breeding gain close to unity. A fuel cycle is designed allowing operation over a long period, requiring refueling with {sup 238}U only. The evolution of nuclides in the GCFR core is calculated using the SCALE system (one-dimensional and three-dimensional). Calculations were done over multiple irradiation cycles including reprocessing. The result is that it is possible to design a fuel and GCFR core with a breeding gain around unity, with recycling of all MAs from cycle to cycle. The burnup reactivity swing is small, improving safety. After several fuel batches an equilibrium core is reached. MA loading in the core remains limited, and the fuel temperature coefficient is always negative.

Rooijen, W.F.G. van; Kloosterman, J.L.; Hagen, T.H.J.J. van der; Dam, H. van [Delft University of Technology (Netherlands)

2005-09-15

197

IRIS Product Recommendations  

NASA Technical Reports Server (NTRS)

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

Short, David A.

2000-01-01

198

A preliminary assessment of a radiatively coupled in-core thermionic space reactor  

NASA Astrophysics Data System (ADS)

A radiatively-coupled in-core thermionic space reactor is proposed that may offer a number of economic and performance benefits. This design combines the advantages of fuel loading after conducting non-nuclear system tests (characteristic of a single-cell design) with the performance benefits of a multi-cell design. Permitting full system tests without nuclear fuel can significantly reduce testing costs while improving reliability of the flight system. In addition, the approach permits the entire system to be transported to the launch site without nuclear fuel. Consequently, program planners can avoid the expensive development of a large shipping cask, or the potential costly completion of system assembly at the launch site. The concept uses a fast reactor as the power source; therefore, the development of a moderator capable of long operational times and high temperature is unnecessary. A fast reactor also permits the use of refractory materials without a significant critical mass penalty from resonance capture of neutrons. The high operating temperature permitted by refractory materials and multi-cell performance improvements will increase system efficiency and reduce radiator surface area requirements. The combination of higher efficiency and reduced radiator area can reduce system size and mass, resulting in launch cost savings. A conceptual design of the reactor power system has been completed. The RSMASS-D model was used to estimate a mass optimized system configuration. System mass predictions for the proposed concept compare favorably to mass predictions for alternative space reactor power system approaches.

Marshall, Albert C.; King, Donald B.; Wilson, Volney C.; Houts, Michael G.

1997-01-01

199

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

200

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

201

Tutorial Notes Iris Recognition Tutorial @ BTAS 2013  

E-print Network

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

Bowyer, Kevin W.

202

Reliable Iris Localization Method With Application To Iris Recognition In Near Infrared Light  

Microsoft Academic Search

Iris recognition is accepted as one of the best biometric method. Implementing this method to the practical system requires the special preprocessing where the iris localization plays a crucial role. Iris localization consists of finding the iris boundaries as well as eyelids. In this paper a simple iris localization algorithm is proposed based on iris image segmentation with histogram analysis.

K. Grabowski; W. Sankowski; M. Zubert; M. Napieralska

2006-01-01

203

Parametric studies on heterogeneous cores for fast breeder reactors: The Pre-Racine and Racine experimental programs  

Microsoft Academic Search

The Pre-Racine and Racine experimental programs, which have been performed on the Masurca critical assembly at Cadarache since 1976, were designed for the study of the neutron physics characteristics of heterogeneous fast reactor cores. Geometrically simple configurations were chosen in which parameters, being typical for heterogeneous cores, were varied in a systematic manner while the basic fissile composition was kept

G. Humbert; F. Kappler; M. Martini; G. Norvez; G. Rimpault; B. Ruelle; W. Scholtyssek; A. Stanculescu

1984-01-01

204

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

205

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

206

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

207

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

208

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

209

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.

210

Recursive Functions in Iris  

Microsoft Academic Search

A complete and efficient implementation of linear, one-side recursive queries in Iris, an object-oriented database management system, is described. It is shown that recursion can be easily and efficiently added to a large class of existing database management systems. A B-tree type access path called the B++ tree that has been implemented to support the computation of recursive functions in

Philippe De Smedt; Stefano Ceri; Marie-anne Neimat; Ming-chien Shan; Rafi Ahmed

1993-01-01

211

Integrated risk information system (IRIS)  

SciTech Connect

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

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

1990-12-31

212

Physics-based multiscale coupling for full core nuclear reactor simulation  

SciTech Connect

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

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

2014-11-01

213

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

214

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

215

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

216

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

NASA Astrophysics Data System (ADS)

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

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

2014-02-01

217

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

SciTech Connect

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

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

1995-06-01

218

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

219

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

SciTech Connect

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

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

2011-07-01

220

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

221

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

222

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

SciTech Connect

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

David W. Nigg

2013-09-01

223

A chemical equilibrium estimate of the aerosols produced in an overheated light water reactor core  

SciTech Connect

The degree of vaporization of light water reactor core materials was estimated using a highly idealized procedure involving (a) specification of the phases that are present for both structural and fuel material, (b) estimation of the vapor pressures exerted by the individual components of each phase, and (c) assuming a degree of vaporization of each phase constituent, allowing equilibration between gaseous and condensed species within the assumed pressure vessel volume. Using this procedure, the aerosol was estimated to consist mainly of silver, indium oxide, cesium hydroxide, and cadmium for pressurized water reactors and cesium hydroxide, cesium iodide, and tellurium for boiling water reactors. If boron is included in the thermodynamic estimate, then boron will significantly alter or dominate the composition of the aerosol in the form of boron oxide and cesium borate. The structural materials make up < 9% of the aerosol at 36 to 57 kg, but this figure is in good agreement with estimates from severe accident sequence analysis studies (17 kg) and from Parker (10.7 kg). The SASCHA data are used in NUREG-0772 and give much higher estimates at 295 and 250 kg.

Wichner, R.P.; Spence, R.D.

1985-09-01

224

IRIS Process (2004-2008)  

EPA Science Inventory

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

225

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

226

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

NASA Astrophysics Data System (ADS)

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

Thomas, Justin W.

2006-12-01

227

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

Federal Register 2010, 2011, 2012, 2013, 2014

...ECCSs) of pressurized water reactors (PWRs...Advanced Pressurized-Water Reactor, U.S. Evolutionary Power Reactor, and AP1000...Criteria for Nuclear Power Plants and Fuel Reprocessing...Systems for Pressurized-Water Reactors.''...

2013-10-25

228

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

Microsoft Academic Search

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

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

2010-01-01

229

Heat Transfer Processes in Reactor Vessel Lower Plenum During Late Phase of In-Vessel Core Melt Progression  

Microsoft Academic Search

Late phase in-vessel core melt progression during the course of a hypothetical severe (melt-down) accident in a light water\\u000a reactor (LWR) is considered, with particular emphasis on thermal processes occurring in the lower plenum of the reactor pressure\\u000a vessel (RPV). The formation of a melt pool, from the initial state of a uniform composition, dried out, debris bed, is investigated.

B. R. Sehgal; V. A. Bui; T. N. Dinh; R. R. Nourgaliev

230

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

231

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

232

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

SciTech Connect

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

David W. Nigg; Devin A. Steuhm

2011-09-01

233

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

SciTech Connect

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

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

2012-09-01

234

Tutorial Notes Iris Recognition Tutorial @ IJCB 2014  

E-print Network

Tutorial Notes Iris Recognition Tutorial @ IJCB 2014 The notes for this tutorial are available online: www.cse.nd.edu/~kwb/IJCB_2014_Tutorial.pdf Publications related to iris recognition: www.cse.nd.edu/~kwb/publications.htm September 29, 2014 #12;Iris Recognition in the Media Iris Recognition Tutorial @ IJCB 2014 September 29

Bowyer, Kevin W.

235

An Evaluation of Iris Pattern Representations  

Microsoft Academic Search

The success of an iris recognition algorithm is partially dependent upon the iris pattern representation computed during feature extraction. Some algorithms in the literature represent iris texture by applying bandpass Alter banks to the segmented iris region. However, the selection of a bandpass filter form, as well as the particular instantiations of that form, are often presented as arbitrary choices.

J. Thornton; M. Savvides; B. V. K. Kumar

2007-01-01

236

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

237

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

238

Internetwork Chromospheric Bright Grains Observed With IRIS and SST  

NASA Astrophysics Data System (ADS)

The Interface Region Imaging Spectrograph (IRIS) reveals small-scale rapid brightenings in the form of bright grains all over coronal holes and the quiet Sun. These bright grains are seen with the IRIS 1330, 1400, and 2796 Å slit-jaw filters. We combine coordinated observations with IRIS and from the ground with the Swedish 1 m Solar Telescope (SST) which allows us to have chromospheric (Ca ii 8542 Å, Ca ii H 3968 Å, H?, and Mg ii k 2796 Å) and transition region (C ii 1334 Å, Si iv 1403 Å) spectral imaging, and single-wavelength Stokes maps in Fe i 6302 Å at high spatial (0\\buildrel{\\prime\\prime}\\over{.} 33), temporal, and spectral resolution. We conclude that the IRIS slit-jaw grains are the counterpart of so-called acoustic grains, i.e., resulting from chromospheric acoustic waves in a non-magnetic environment. We compare slit-jaw images (SJIs) with spectra from the IRIS spectrograph. We conclude that the grain intensity in the 2796 Å slit-jaw filter comes from both the Mg ii k core and wings. The signal in the C ii and Si iv lines is too weak to explain the presence of grains in the 1300 and 1400 Å SJIs and we conclude that the grain signal in these passbands comes mostly from the continuum. Although weak, the characteristic shock signatures of acoustic grains can often be detected in IRIS C ii spectra. For some grains, a spectral signature can be found in IRIS Si iv. This suggests that upward propagating acoustic waves sometimes reach all the way up to the transition region.

Martínez-Sykora, Juan; Rouppe van der Voort, Luc; Carlsson, Mats; De Pontieu, Bart; Pereira, Tiago M. D.; Boerner, Paul; Hurlburt, Neal; Kleint, Lucia; Lemen, James; Tarbell, Ted D.; Title, Alan; Wuelser, Jean-Pierre; Hansteen, Viggo H.; Golub, Leon; McKillop, Sean; Reeves, Kathy K.; Saar, Steven; Testa, Paola; Tian, Hui; Jaeggli, Sarah; Kankelborg, Charles

2015-04-01

239

Effects of Iris Surface Curvature on Iris Recognition Joseph Thompson  

E-print Network

jthomp11@nd.edu Patrick Flynn University of Notre Dame flynn@nd.edu Kevin Bowyer University of Notre Dame and shows that differences in iris curvature degrade matching ability. To our knowl- edge, no other work has

Bowyer, Kevin W.

240

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

241

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

242

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

SciTech Connect

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

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

1991-12-31

243

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

SciTech Connect

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

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

1991-01-01

244

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

245

Integration of the Empirical Exospheric GCPM Plasma Model into IRI  

NASA Technical Reports Server (NTRS)

The Global Core Plasma Model (GCPM) is an empirical description of typical thermal magnetospheric plasma densities in the plasmasphere, plasma trough, and polar cap. The GCPM makes use of the International Reference Ionosphere (IRI) for low altitudes. Densities are continuous and smooth. Plasmaspheric ion composition is also included in the GCPM. For the purpose of supporting the ionospheric community, the densities derived in the GCPM for high altitudes will be expressed as an extension of the IRI. The GCPM exospheric extension of IRI improves the topside densities and provides typical thermal plasma densities for the plasmasphere, trough, and polar cap. The GCPM is modular, having been designed for continued improvement as statistical density and composition measurements become available.

Gallagher, Dennis L.; Bilitza, Dieter

2000-01-01

246

Preliminary Probabilistic Safety Assessment of the IRIS Plant  

SciTech Connect

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

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

2002-07-01

247

Bartus Iris biometrics  

SciTech Connect

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

Johnston, R.; Grace, W.

1996-07-01

248

IRIS: Teachable Moments  

NSDL National Science Digital Library

The Education and Outreach groups of IRIS (Incorporated Research Institutions for Seismology), the University of Portland, and UNAVCO are developing a set of 'Teachable Moments', Power Point and .pdf presentations which provide a short summary of major tectonic events that can be used by educators to incorporate discussions of earthquakes and volcanic events, and resulting hazards. The presentations, generated by seismologists, geodesists, and educators, are available within a few hours to one day after the earthquake. The presentations include a variety of content allowing educators to customize the information for their classes. Common elements include US Geological Survey earthquake and volcano information, GPS time series from the EarthScope Plate Boundary Observatory (PBO) network, plate tectonic and regional tectonic maps and summaries, computer animations, seismograms, photos, and other event-specific information.

2002-04-30

249

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

250

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

SciTech Connect

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

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

2014-12-15

251

Core burnup calculation and accidents analyses of a pressurized water reactor partially loaded with rock-like oxide fuel  

Microsoft Academic Search

A rock-like oxide (ROX) fuel – light water reactor (LWR) burning system has been studied for efficient plutonium transmutation. For the improvement of small negative reactivity coefficients and severe transient behaviors of ROX fueled LWRs, a partial loading core of ROX fuel assemblies with conventional UO2 assemblies was considered. As a result, although the reactivity coefficients could be improved, the

H. Akie; Y. Sugo; R. Okawa

2003-01-01

252

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

253

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

254

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

Microsoft Academic Search

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

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

2009-01-01

255

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...

256

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

257

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

258

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

259

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

E-print Network

In: Potential of Small Nuclear Reactors for Future Clean andSmall Module (PRISM) liquid-metal reactor. Final report”. In: Office of Nuclear Reactornuclear reactor. This is accomplished by incorporating a small

Qvist, Staffan Alexander

2013-01-01

260

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

261

Prostatic carcinoma bilateral iris metastases  

PubMed Central

We described a patient with bilateral iris metastases resulted from prostatic cancer. Slit lamp and ultrasonography examination of the both eye demonstrated tumor of the iris, as an amelanotic vascular mass located on the superior temporal quadrant. On open biopsy revealed undifferentiated tissue that stained strongly positive for prostate carcinoma, confirming the diagnosis of metastasis prostate adenocarcinoma. Early diagnostic procedures are essential for the causal therapy of prostate carcinoma as the primary neoplasm. PMID:22642599

Sarenac, Tatjana S.; Janicijevic-Petrovic, Mirjana A.; Sreckovic, Suncica B.; Radovanovic, Milan R.; Vulovic, Dejan D.; Janicijevic, Katarina M.

2012-01-01

262

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

263

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

SciTech Connect

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

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

2004-10-06

264

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

265

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

Microsoft Academic Search

The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of

B. Boer; A. M. Ougouag

2010-01-01

266

Moxifloxacin and bilateral acute iris transillumination  

PubMed Central

Recent publications have alerted clinicians to a syndrome of uveitic transilluminating iris depigmentation associated with systemic fluoroquinolones and other antibiotics. Bilateral acute iris transillumination, which is associated with loss of the iris pigment epithelium and results in iris transillumination, differs from the previously described bilateral acute depigmentation of the iris, which is associated with atrophy of the iris stroma without transillumination. We present a case of fluoroquinolone-associated uveitis with anterior segment optical coherence tomography imaging to highlight some observations about this syndrome. We interpret pharmacokinetic data to help explain why oral, but not topical, moxifloxacin may cause fluoroquinolone-associated uveitis. PMID:23514193

2013-01-01

267

IRIS Earthquake Browser (IEB)  

NSDL National Science Digital Library

The IRIS Earthquake Browser (IEB) combines the DMC's large database of earthquakes with the popular Google Maps web interface. The IEB is useful both educationally and as a research tool. With the IEB you can quickly find earthquakes in any region of the globe and then import this information into the GEON Integrated Data Viewer (IDV) where the hypocenters may be visualized in three dimensions. The IEB features a simple, yet powerful, user interface. To start using it, you simply zoom and pan a map of the world just as you would with any other Google map. By default, the IEB shows the most recent 200 events in any given view. The number of events shown can be varied from 100 to 5000 while the prioritization of events (which events have priority to be shown) can be toggled between recent events and larger events. Links at the top of the page allow the user to download the data plotted on the map. The data can be viewed as an HTML table or exported as a NetCDF (Network Common Data Form) binary. The exported NetCDF data can then be imported into the GEON IDV.

268

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

NASA Astrophysics Data System (ADS)

Using modern methods of reactor physics, we performed full-scale calculations of the Oklo natural reactor. For reliability, we used recent versions of two Monte Carlo codes: the Russian code MCU-REA and the well-known international code MCNP. Both codes produced similar results. We constructed a computer model of the Oklo reactor zone RZ2 which takes into account all details of design and composition. The calculations were performed for three fresh cores with different uranium contents. Multiplication factors, reactivities, and neutron fluxes were calculated. We also estimated the temperature and void effects for the fresh core. As would be expected, we found for the fresh core a significant difference between reactor and Maxwell spectra, which had been used before for averaging cross sections in the Oklo reactor. The averaged cross section of 62149Sm and its dependence on the shift of a resonance position Er (due to variation of fundamental constants) are significantly different from previous results. Contrary to the results of previous papers, we found no evidence of a change of the samarium cross section: a possible shift of the resonance energy is given by the limits -73??Er?62 meV. Following tradition, we have used formulas of Damour and Dyson to estimate the rate of change of the fine structure constant ?. We obtain new, more accurate limits of -4×10-17??·/??3×10-17yr-1. Further improvement of the accuracy of the limits can be achieved by taking account of the core burn-up. These calculations are in progress.

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

2006-12-01

269

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

SciTech Connect

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

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

1987-04-15

270

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

SciTech Connect

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

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

2006-12-15

271

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

272

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

SciTech Connect

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

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

2006-07-01

273

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

274

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

275

Nanostructures formed in pure quartz glass under irradiation in the reactor core  

NASA Astrophysics Data System (ADS)

Optical spectroscopy and X-ray diffraction techniques were used for studying nanoscale particles grown in pure SiO2 glass under irradiation with fast neutron fluencies within 6×1016-5·1019 cm-2 and gamma-quanta ~1.8×1020 cm-2 in the reactor core in water. The neutron irradiation results in destroying of the initial ?- and ?-quartz mesoscopic order of 1.7 and 1.2 nm sizes and growing of cristobalite and tridymite nanocrystals of 16 and 8 nm sizes in the thermal peaks of displacements reapectively. The point defects (oxygen deficient E?s, E'1, E'2 and non-bridging oxygen centers) induced by the ?-irradiation are accumulated in the nanocrystals shell of 0.65-0.85 nm thickness. Interaction of close point defects at the nanocrystal-glass interface causes the splitting of optical absorption bands into the intensive (D~2-4) resonances characteristic for local interband electron transitions, having the width of 10-15 nm close to the nanocrystals' sizes and the energy depending on their structure.

Ibragimova, E. M.; Mussaeva, M. A.; Kalanov, M. U.

2014-04-01

276

Comparison of two IRI electron-density plasmasphere extensions with GPS-TEC observations  

Microsoft Academic Search

Comparisons of two model results with Global Positioning System, GPS-TEC, measurements have been carried out for different latitudinal, solar activity, magnetic activity, diurnal, and seasonal conditions. The models evaluated are the Global Core Plasma Model (GCPM-2000) and the IRI extension (IRI*) with the Russian plasmasphere model. Data from 23 observatories providing GPS-TEC and ionosonde data have been used. It is

T. L. Gulyaeva; D. L. Gallagher

2007-01-01

277

Comparison of two IRI electron-density plasmasphere extensions with GPS-TEC observations  

Microsoft Academic Search

Comparisons of two model results with Global Positioning System, GPS-TEC, measurements have been carried out for different latitudinal, solar activity, magnetic activity, diurnal, and seasonal conditions. The models evaluated are the Global Core Plasma Model (GCPM-2000) and the IRI extension (IRI?) with the Russian plasmasphere model. Data from 23 observatories providing GPS-TEC and ionosonde data have been used. It is

T. L. Gulyaeva; D. L. Gallagher

2007-01-01

278

IRIS TOXICOLOGICAL REVIEW OF ACROLEIN (2003 Final)  

EPA Science Inventory

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

279

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

280

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

281

A New Fake Iris Detection Method  

Microsoft Academic Search

Recent research works have revealed that it is not difficult to spoof an automated iris recognition system using fake iris\\u000a such as contact lens and paper print etc. Therefore, it is very important to detect fake iris as much as possible. In this\\u000a paper, we propose a new fake iris detection method based on wavelet packet transform. First, wavelet packet

Xiaofu He; Yue Lu; Pengfei Shi

2009-01-01

282

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

283

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

Microsoft Academic Search

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

Y. A. Choi; M. A. Feltus

1995-01-01

284

Neutron flux measurements in the side-core region of Hunterston B advanced gas-cooled reactor  

SciTech Connect

The core restraints of advanced gas-cooled reactors are important structural components that are required to maintain the geometric integrity of the cores. A review of neutron dosimetry for the sister stations Hunterston B and Hinkley Point B identified that earlier conservative assessments predicted high thermal neutron dose rates to key components of the restraint structure (the restraint rod welds), with the implication that some of them may be predicted to fail during a seismic event. A revised assessment was therefore undertaken [Thornton, D. A., Allen, D. A., Tyrrell, R. J., Meese, T. C., Huggon, A.P., Whiley, G. S., and Mossop, J. R., 'A Dosimetry Assessment for the Core Restraint of an Advanced Gas Cooled Reactor,' Proceedings of the 13. International Symposium on Reactor Dosimetry (ISRD-13, May 2008), World Scientific, River Edge, NJ, 2009, W. Voorbraak, L. Debarberis, and P. D'hondt, Eds., pp. 679-687] using a detailed 3D model and a Monte Carlo radiation transport program, MCBEND. This reassessment resulted in more realistic fast and thermal neutron dose recommendations, the latter in particular being much lower than had been thought previously. It is now desirable to improve confidence in these predictions by providing direct validation of the MCBEND model through the use of neutron flux measurements. This paper describes the programme of work being undertaken to deploy two neutron flux measurement 'stringers' within the side-core region of one of the Hunterston B reactors for the purpose of validating the MCBEND model. The design of the stringers and the determination of the preferred deployment locations have been informed by the use of detailed MCBEND flux calculations. These computational studies represent a rare opportunity to design a flux measurement beforehand, with the clear intention of minimising the anticipated uncertainties and obtaining measurements that are known to be representative of the neutron fields to which the vulnerable steel restraint components are exposed. (authors)

Allen, D.A. [Serco, Rutherford House, Quedgeley, Gloucester, GL2 4NF (United Kingdom); Shaw, S.E. [British Energy, Barnett Way, Barnwood, Gloucester, GL4 3RS (United Kingdom); Huggon, A.P.; Steadman, R.J.; Thornton, D.A. [Serco, Rutherford House, Quedgeley, Gloucester, GL2 4NF (United Kingdom); Whiley, G.S. [British Energy, Barnett Way, Barnwood, Gloucester, GL4 3RS (United Kingdom)

2011-07-01

285

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

SciTech Connect

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

Shadday, M.A.

2001-07-17

286

Biometric Personal Identification Based on Iris Recognition  

Microsoft Academic Search

With an increasing emphasis on security, automated personal identification based on biometrics has been receiving extensive attention over the 1990s. Iris recognition, as an emerging biometric recognition approach is becoming a very active topic in both research and practical applications. Iris recognition is the process of recognizing a person by analyzing the apparent pattern of his or her iris. A

Shimaa M. Elsherief; M. E. Allam; M. W. Fakhr

2006-01-01

287

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

288

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

289

Broken nylon iris fixation sutures.  

PubMed

Broken nylon iris sutures, used to fixate the Worst suture lens, occurred in 41 of a series of 215 eyes, a remarkably high frequency. The estimated average time to break was 27.8 months. Over three fourths of the borken sutures were discovered incidentally on a return visit. Complications related to the broken suture occurred in 21 eyes. The characteristic complication was corneal epithelial edema caused by intermittent touch of the endothelium by the broken suture; spontaneous lens dislocation was infrequent. Light and scanning electron microscopy revealed that biodegradation caused the suture break. Broken iris sutures can be avoided by fixation of the lens with a nonbiodegradable suture. PMID:517621

Cohan, B E; Pearch, A C; Schwartz, S

1979-12-01

290

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

PubMed

A comprehensive analysis has been performed to investigate the conversion of the Syrian MNSR (miniature neutron source reactor) from current HEU fuel to selected alternatives LEU and MEU fuels. For this purposes the core design calculations related to design and engineering of LEU and MEU fuels have been carried out using the codes WIMSD/4 and BORGES-part of the MTR-PC and the code CITATION. Aiming at reducing the fuel enrichment by maintaining reactor power, thermal neutron flux and excess reactivity in the same range of the current MNSR design, two fuel alternatives of LEU (UO(2)-Mg) and MEU (U(3)Si(x)-Al) have been investigated. The results indicate that the first type (UO(2)-Mg) realizes the criticality conditions with low enrichment of 20% using the similar overall design of the present HEU fuel pins, whereas the second type (U(3)Si-Al) requires increasing the enrichment up to 33%. For the purpose of reactor core lifetime extension the possibility of mixing the burnable poisons Gd(157) and Cd(113) in the fresh core has been also explored. Thus, the calculation results indicate that the long-term control effect of Cd(113) on the excess reactivity is more homogeneous over the time due to the lower burn up rate of this burnable poison. PMID:19628402

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

2009-10-01

291

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

292

Nuclear reactor  

Microsoft Academic Search

A nuclear reactor is described which comprises a reactor vessel, a core housed in the reactor vessel, an ultrasonic transducer mounted in the vicinity of the upper end of the core for emitting and receiving an ultrasonic wave pulse signal propagating above the core. A means is provided for rotating the transducer by a prescribed angle to scan horizontally the

M. Furudate; T. Miyazawa; H. Mizuguchi; K. Sasaki; N. Uesugi

1981-01-01

293

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 ...

294

Historical comparisons of IRI and early ionograms  

NASA Astrophysics Data System (ADS)

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

Rice, Donald; Sojka, Jan J.

2015-04-01

295

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

296

Analysis of overall temperature coefficient of reactivity of the VHTRC-1 core with a nuclear design code system for the high-temperature engineering test reactor  

Microsoft Academic Search

In this paper the accuracy of the nuclear design code system for the High-Temperature Engineering Test Reactor (HTTR) is evaluated for the neutronic characteristics that depend on core temperature by analyzing the overall temperature coefficients of reactivity and the effective multiplication factors obtained by an experiment in which the Very High Temperature Reactor Critical Assembly (VHTRC) is heated from ambient

K. Yamashita; I. Murata; R. Shindo

1992-01-01

297

Parametric studies on heterogeneous cores for fast breeder reactors: The Pre-Racine and Racine experimental programs  

SciTech Connect

The Pre-Racine and Racine experimental programs, which have been performed on the Masurca critical assembly at Cadarache since 1976, were designed for the study of the neutron physics characteristics of heterogeneous fast reactor cores. Geometrically simple configurations were chosen in which parameters, being typical for heterogeneous cores, were varied in a systematic manner while the basic fissile composition was kept the same. Measurements were made especially of the critical mass, the distributions of reaction rates and the spectral indices, the reactivity of sodium voiding, and control rod worths. Analyses were made independently by Commissariat a l'Energie Atomique (CEA) and DEBENE using their own calculational techniques and cross sections. No bias for core heterogeneity was found on critical mass predictions. The CEA calculations for void reactivities are consistent in heterogeneous and homogeneous configurations. For the calculation of local parameters, e.g., reaction rates and spectral indices, more sophisticated methods must be applied in heterogeneous cores, as transport effects also become more important in fissile zones with increasing fertile volume fraction. It was found at CEA that the ratio of the calculated reactivity of a central control rod to the experimental value does not change with the core size or with the presence of internal breeder zones.

Humbert, G.; Kappler, F.; Martini, M.; Norvez, G.; Rimpault, G.; Ruelle, B.; Scholtyssek, W.; Stanculescu, A.

1984-07-01

298

Assessing the IRIS Professional Development Model: Impact Beyond the Workshops  

NASA Astrophysics Data System (ADS)

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

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

2003-12-01

299

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

300

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

NASA Astrophysics Data System (ADS)

The design of a fluidized bed atomic layer deposition (ALD) reactor is described in detail. The reactor consists of three parts that have all been placed in one protective cabinet: precursor dosing, reactor, and residual gas treatment section. In the precursor dosing section, the chemicals needed for the ALD reaction are injected into the carrier gas using different methods for different precursors. The reactor section is designed in such a way that a homogeneous fluidized bed can be obtained with a constant, actively controlled, reactor pressure. Furthermore, no filters are required inside the reactor chamber, minimizing the risk of pressure increase due to fouling. The residual gas treatment section consists of a decomposition furnace to remove residual precursor and a particle filter and is installed to protect the pump. In order to demonstrate the performance of the reactor, SiO2 particles have been coated with TiO2 using tetrakis-dimethylamino titanium (TDMAT) and H2O as precursors. Experiments with varying pulse times show that saturated growth can be obtained with TDMAT pulse times larger than 600 s. Analysis of the powder with High-Angle Annular Dark-Field Scanning Transmission Electron Microscopy (HAADF-STEM) and energy dispersive X-ray spectroscopy confirmed that after 50 cycles, all SiO2 particles were coated with a 1.6 nm homogenous shell of TiO2.

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

2014-01-01

301

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

302

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

SciTech Connect

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

Conklin, J.C.

1986-08-01

303

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

NASA Astrophysics Data System (ADS)

A computer code developed to analyze the thermal, hydraulic and fuel particle performance characteristics of axisymmetric pebble bed reactors is presented. This FORTRAN 5 program is overlayed into three main segments to minimize core storage. Variable dimensioning was introduced so that storage allocation is based on the user-specified problem size, thus allowing the code to handle any problem with uniform mesh interval. This provides the flexibility to choose thermal problem mesh sizes different from those used in the physics analyses. The code can also handle a non-uniform void fraction distribution. DOE

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

1981-05-01

304

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

305

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

E-print Network

Reactors for Future Clean and Safe Energy Sources (1992),future, a majority of the world energy supply will need to be replaced by clean new energy sourcesenergy source. 1 There is, however, very limited room for future

Qvist, Staffan Alexander

2013-01-01

306

Development of a low enrichment uranium core for the MIT reactor  

E-print Network

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

Newton, Thomas Henderson

2006-01-01

307

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

308

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

Federal Register 2010, 2011, 2012, 2013, 2014

...NUCLEAR REGULATORY COMMISSION...NRC-2012-0134] Initial Test Program of Emergency...Boiling-Water Reactors AGENCY: Nuclear Regulatory Commission...SUMMARY: The U.S. Nuclear Regulatory Commission...DG-1277, ``Initial Test Program of...

2012-06-15

309

The Neutronics Design and Analysis of a 200MW(electric) Simplified Boiling Water Reactor Core  

Microsoft Academic Search

A 200-MW(electric) simplified boiling water reactor (SBWR) was designed and analyzed under sponsorship of the U.S. Department of Energy Nuclear Energy Research Initiative program. The compact size of a 200-MW(electric) reactor makes it attractive for countries with a less well developed engineering infrastructure, as well as for developed countries seeking to tailor generation capacity more closely to the growth of

Daniel R. Tinkler; Thomas J. Downar

2003-01-01

310

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

Microsoft Academic Search

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

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

1992-01-01

311

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

312

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

313

Fuel efficient hydrodynamic containment for gas core fission reactor rocket propulsion. Final report, September 30, 1992--May 31, 1995  

SciTech Connect

Gas core reactors can form the basis for advanced nuclear thermal propulsion (NTP) systems capable of providing specific impulse levels of more than 2,000 sec., but containment of the hot uranium plasma is a major problem. The initial phase of an experimental study of hydrodynamic confinement of the fuel cloud in a gas core fission reactor by means of an innovative application of a base injection stabilized recirculation bubble is presented. The development of the experimental facility, a simulated thrust chamber approximately 0.4 m in diameter and 1 m long, is described. The flow rate of propellant simulant (air) can be varied up to about 2 kg/sec and that of fuel simulant (air, air-sulfur hexafluoride) up to about 0.2 kg/sec. This scale leads to chamber Reynolds numbers on the same order of magnitude as those anticipated in a full-scale nuclear rocket engine. The experimental program introduced here is focused on determining the size, geometry, and stability of the recirculation region as a function of the bleed ratio, i.e. the ratio of the injected mass flux to the free stream mass flux. A concurrent CFD study is being carried out to aid in demonstrating that the proposed technique is practical.

Sforza, P.M.; Cresci, R.J.

1997-05-31

314

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

315

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

E-print Network

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

Sauer, Ildo Luis

1985-01-01

316

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

317

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

SciTech Connect

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 x 10{sup 20} n/cm{sup 2}. A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

Geslot, B.; Filliatre, P.; Barbot, L.; Jammes, C.; Breaud, S.; Oriol, L.; Villard, J.-F. [CEA, DEN, Cadarache, SPEx/LDCI, F-13108 Saint-Paul-lez-Durance (France); Vermeeren, L. [SCK-CEN, Boeretang 200, B-2400 Mol (Belgium); Lopez, A. Legrand [CEA, DEN, Saclay, SIREN/LECSI, F-91400 Saclay (France)

2011-03-15

318

a Novel Micromegas Detector for In-Core Nuclear Reactor Neutron Flux Measurements  

NASA Astrophysics Data System (ADS)

Future fast nuclear reactors designed for energy production and transmutation of nuclear wastes need new neutrons detectors able to measure the neutron flux over a large energy range from thermal energies to several MeV. A novel compact and very small detector, named Piccolo-Micromegas has been developed for this purpose. Description of the detector configuration especially dedicated to neutron detection inside nuclear reactor is given. The advantage of this detector over conventional neutron flux detectors and the results obtained with the first prototype are presented.

Andriamonje, S.; Aune, S.; Giganon, A.; Giomataris, I.; Pancin, J.; Riallot, M.; Blandin, C.; Breaud, S.; Geslot, B.; Jammes, C.; Kadi, Y.; Sarchiapone, L.; Ban, G.; Laborie, P.; Lecolley, J. F.; Steckmeyer, J. C.; Tillier, J.; Rosa, R.; Andriamonje, G.

2006-04-01

319

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

NASA Astrophysics Data System (ADS)

The IRIS E&O Program was initiated 10 years ago, some 15 years after the creation of the IRIS Consortium, as IRIS members increasingly recognized the fundamental need to communicate the results of scientific research more effectively and to attract more students to study Earth science. Since then, IRIS E&O has received core funding through successive 5-year cooperative agreements with NSF, based on proposals submitted by IRIS. While a small fraction of the overall Consortium budget, this consistent funding has allowed the development of strong, long-term elements within the E&O Program, including summer internships, IRIS/USGS museum displays, seismographs in schools, IRIS/SSA Distinguished Lecture series, and professional development for middle school and high school teachers. Reliable funding has allowed us to develop expertise in these areas due to the longevity of the programs and the continuous improvement resulting from ongoing evaluations. Support from Consortium members, including volunteering time and expertise, has been critical for the program, as the Consortium has to continually balance the value of E&O products versus equipment and data services for seismology research. The E&O program also provides service to the Consortium, such as PIs being able to count on and leverage IRIS resources when defining the broader impacts of their own research. The reliable base has made it possible to build on the core elements with focused and innovative proposals, allowing, for example, the expansion of our internship program into a full REU site. Developing collaborative proposals with other groups has been a key strategy where IRIS E&O's long-term viability can be combined with expertise from other organizations to develop new products and services. IRIS can offer to continue to reliably deliver and maintain products after the end of a 2-3 year funding cycle, which can greatly increase the reach of the project. Consortium backing has also allowed us to establish an educational fund in honor of the late John Lahr. This fund, which is comprised of individual donations, is being used to provide seismographs to schools along with professional development and ongoing support from the E&O program. We are also developing a plan for attracting larger private and/or foundation funds for new E&O activities, leveraging the reputation of a long-term program.

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

2009-12-01

320

Core and fuel design considerations for boiling water reactors with passive safety features  

Microsoft Academic Search

Recently there has been increasing utility and government interest in potential future nuclear units combining the characteristics of greater simplicity and passive safety features. In response to such interest, General Electric (GE) initiated studies of a boiling water reactor (BWR) with simplified power generation, safety, and heat removal systems. This paper describes the key features of GE's simplified BWR (SBWR),

L. E. Fennern; R. M. Fawcett

1990-01-01

321

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

322

Reconocimiento Automtico de Patrones de Iris  

E-print Network

.Localización del Iris 5.Comparación de Patrones de Iris 6.Experimentos Realizados 7.Problemática y retos campo (rango de enfoques), de 1 cm. de profundidad máximo. Centrado de la imagen: Ambos sistemas

Autonoma de Madrid, Universidad

323

Iris recognition: an emerging biometric technology  

Microsoft Academic Search

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

RICHARD P. WILDES

1997-01-01

324

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

325

Core Calculation of 1 MWatt PUSPATI TRIGA Reactor (RTP) using Monte Carlo MVP Code System  

NASA Astrophysics Data System (ADS)

The Monte Carlo MVP code system was adopted for the Reaktor TRIGA PUSAPTI (RTP) core calculation. The code was developed by a group of researcher of Japan Atomic Energy Agency (JAEA) first in 1994. MVP is a general multi-purpose Monte Carlo code for neutron and photon transport calculation and able to estimate an accurate simulation problems. The code calculation is based on the continuous energy method. This code is capable of adopting an accurate physics model, geometry description and variance reduction technique faster than conventional method as compared to the conventional scalar method. This code could achieve higher computational speed by several factors on the vector super-computer. In this calculation, RTP core was modeled as close as possible to the real core and results of keff flux, fission densities and others were obtained.

Karim, Julia Abdul

2008-05-01

326

WWER-1000 core and reflector parameters investigation in the LR-0 reactor  

SciTech Connect

Measurements and calculations carried out in the core and reflector of WWER-1000 mock-up are discussed: - the determination of the pin-to-pin power distribution in the core by means of gamma-scanning of fuel pins and pin-to-pin calculations with Monte Carlo code MCU-REA and diffusion codes MOBY-DICK (with WIMS-D4 cell constants preparation) and RADAR - the fast neutron spectra measurements by proton recoil method inside the experimental channel in the core and inside the channel in the baffle, and corresponding calculations in P{sub 3}S{sub 8} approximation of discrete ordinates method with code DORT and BUGLE-96 library - the neutron spectra evaluations (adjustment) in the same channels in energy region 0.5 eV-18 MeV based on the activation and solid state track detectors measurements. (authors)

Zaritsky, S. M.; Alekseev, N. I.; Bolshagin, S. N. [RRC Kurchatov Inst., 1 Kurchatov Sq., Moscow, 123182 (Russian Federation); Riazanov, D. K.; Lichadeev, V. V. [Research Inst. of Atomic Reactors, Dimitrovgrad 10, 433510 (Russian Federation); Ocmera, B. [Nuclear Research Inst., Rez, 25068 (Czech Republic); Cvachovec, F. [Univ. of Defense, 65 Kounicova st., Brno, 61200 (Czech Republic)

2006-07-01

327

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

SciTech Connect

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 [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

2005-02-06

328

Thermohydraulics in a high-temperature gas-cooled reactor primary loop during early phases of unrestricted core-heatup accidents  

SciTech Connect

In High Temperature Gas Cooled Reactor (HTGR) siting considerations, the Unrestricted Core Heatup Accidents (UCHA) are considered as accidents of highest consequence, corresponding to core meltdown accidents in light water reactors. Initiation of such accidents can be, for instance, due to station blackout, resulting in scram and loss of all main loop forced circulation, with none of the core auxiliary cooling system loops being started. The result is a slow but continuing core heatup, extending over days. During the initial phases of such UCHA scenarios, the primary loop remains pressurized, with the system pressure slowly increasing until the relief valve setpoint is reached. The major objectives of the work described here were to determine times to depressurization as well as approximate loop component temperatures up to depressurization.

Kroeger, P.G.; Colman, J.; Hsu, C.J.

1983-01-01

329

Using Neural Networks to Predict Core Parameters in a Boiling Water Reactor  

SciTech Connect

The problem of optimizing refueling in a nuclear boiling water reactor is difficult since it concerns combinatorial optimization and it is NP-Complete. In order to solve this problem, many techniques have been applied, ranging from expert systems to genetic algorithms. In most of these procedures, nuclear reactor simulators are used, which require a longer computation time, to evaluate the goodness of the proposed solutions. As the processes are iterative, many evaluations with the simulator are necessary, and this makes the process extremely slow. In this paper, the use of trained neural networks (NNs) is proposed as an alternative to the simulator, and the results of the NN training are shown in order to predict some variables of interest in the optimization, such as the effective multiplication factor and some thermal limits, related to safety aspects. Finally, a study about the effect of modifying several NN parameters is shown.

Ortiz, Juan Jose; Requena, Ignacio [Dpto. Ciencias Computacion e I.A. Univ. Granada (Spain)

2003-03-15

330

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

E-print Network

reload380 optimization using artificial ant colony connective networks. Ann. Nucl. Energy 35, 1606–1612.381 DeChaine, M., Feltus, M., 1995. Nuclear fuel management optimization using genetic algorithms. Nucl.382 Technol. 111, 109–114.383 Esquivel... for nuclear reactor fuel management optimisation. Ann. Nucl. Energy 33, 1039–406 1057.407 Khoshahval, F., Zolfaghari, A., Minuchehr, H., Sadighi, M., Norouzi, A., 2010. PWR fuel management408 optimization using continuous particle swarm intelligence. Ann. Nucl...

Hill, Natasha J.; Parks, Geoffrey T.

2014-08-24

331

Calculation of the dynamic response of reactor containment systems to full core explosions  

Microsoft Academic Search

ASTARTE, a comprehensive time-dependent two-dimensional Lagrangian hydrodynamic code has been developed to determine numerically the loadings and strains arising within the primary containment system following a high-energy reactor excursion. The detail, both temporal and spatial, which is achievable by judicious use of such a code allows the safety analyst to examine the response of individual sections of the containment. It

M. S. Cowler; N. E. Hoskin; A. G. Rowlinson

1975-01-01

332

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

333

An iris segmentation algorithm based on edge orientation for off-angle iris recognition  

NASA Astrophysics Data System (ADS)

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; Barstow, Del; Santos-Villalobos, Hector; Boehnen, Christopher

2013-03-01

334

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

335

Key Techniques and Methods for Imaging Iris in Focus  

Microsoft Academic Search

Automated iris recognition is a promising method for noninvasive verification of identity. How to acquire an iris image in focus is a key issue in iris recognition. Based on imaging properties of a simple lens, working principles of fixed focus and auto focus imaging systems are described. Key techniques for imaging iris in focus are discussed in this paper, such

Yuqing He; Jiali Cui; Tieniu Tan; Yangsheng Wang

2006-01-01

336

IRIS TOXICOLOGICAL REVIEW AND SUMMARY DOCUMENTS FOR 2-HEXANONE  

EPA Science Inventory

EPA will conduct an assessment of the noncancer health effects of 2-hexanone. The IRIS program will prepare an IRIS assessment for 2-hexanone. The IRIS assessment for 2-hexanone will consist of a Toxicological Review and an IRIS Summary. The Toxicological Review is a critical ...

337

Robust Long Range Iris Recognition from Video Using Super Resolution  

E-print Network

identifications and to control access. However, iris images ac- quired by the IOM and other long range systems are systems, (2) automatic iris mask generation of occluded regions, (3) iris matching performance enhancement, we focus on estimating a mask for the iris texture in the polar coordinate system. Unlike most

Eskenazi, Maxine

338

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

339

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

340

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

341

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

342

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

343

Fuel foaming and collapse during light water reactor core meltdown accidents  

SciTech Connect

Fuel melting in severe core damage accidents will lead to the rapid release of fission gas from the fuel matrix and the volatilization of low boiling point metallic inclusions, which can be expected to significantly influence molten fuel dynamics. A quantitative analysis of UO/sub 2/ foaming potential is based on an assessment of the time characteristics for bubble growth, surface escape, film thinning, and bubble coalescence. Analysis indicates that although the potential exists for early molten UO/sub 2/ foaming, such foams are basically unstable and tend to collapse, thereby releasing volatilized fission products from the molten fuel debris. Release of such fission products will impact radiological source term evaluation and can result in up to a 40% reduction in the residual decay heat within the core debris. This reduction in core debris heat level will influence the timing and meltdown sequence for such accidents and can impact the heat load requirements of residual heat removal systems or other engineered melt mitigation devices.

Cronenberg, A.W.; Croucher, D.W.; McDonald, P.E.

1984-11-01

344

Please cite this article in press as: Shuffler, C., et al., Thermal hydraulic analysis for grid supported pressurized water reactor cores. Nucl. Eng. Des. (2009), doi:10.1016/j.nucengdes.2008.12.028  

E-print Network

.elsevier.com/locate/nucengdes Thermal hydraulic analysis for grid supported pressurized water reactor cores C. Shuffler , J. Trant, JPlease cite this article in press as: Shuffler, C., et al., Thermal hydraulic analysis for grid supported pressurized water reactor cores. Nucl. Eng. Des. (2009), doi:10.1016/j.nucengdes.2008

Malen, Jonathan A.

345

Irradiation creep of V?Ti?Cr alloy in BR-10 reactor core instrumented experiments  

NASA Astrophysics Data System (ADS)

A thin wall tubular-type speciment of 4%Ti-4%Cr vanadium alloy was tested for creep under irradiation in BR-10 reactor at 713-723 K and at 8.6 × 10 18 n/m 2s fast neutron flux. A fluence at the end of the experiment have reached 5.8 × 10 25 n/m 2. Specimen deformation measurements were performed by a dynamometric method based on a stress relaxation control provided during irradiation under constant load applied. During the experiment 13 deformation curves were obtained for different stress levels ranged up to 165 MPa. At the same time the yield stress of the irradiated specimen was periodically determined. The irradiation creep rate has been found to be proportional to the stress up to 110-120 MPa with the module equal to 3.3 × 10 -12 dpa -1Pa -1. At higher streses, a creep process essentially accelerates. The results on V?Ti?Cr alloy are discussed in respect to data obtained for stainless steels in earlier BR-10 reactor experiments.

Troyanov, V. M.; Bulkanov, M. G.; Kruglov, A. S.; Krjuchkov, E. A.; Nikulin, M. P.; Pevchykh, J. M.; Rusanov, A. E.; Smirnoff, A. A.; Votinov, S. N.

1996-10-01

346

Development and implementation of monitoring for the reactor core of unit No. 5 of the Novovoronezh nuclear plant by local parameters  

NASA Astrophysics Data System (ADS)

In the course of upgrading the unit no. 5 reactor core of the Novovoronezh nuclear power plant, operational limits by local parameters, which limit the admissible linear power density and the relative power of fuel elements, were established. Due to applying modern computer technologies in systems of the in-core monitoring, the calculation of power density for all fuel elements in the real-time mode is implemented. To monitor the power density of fuel elements, the algorithm for determining the limiting linear power density is developed depending on the reactor core height and on the average nuclear fuel burnup. The admissible relative power of fuel elements is determined. In the course of the performed work, the excessive conservative limitations on nonuniformity of the reactor power density are excluded. The monitoring of power density by local parameters instead of indirect K q (fuel-assembly relative power) and K v (relative power of the fuel assembly section) made it possible to increase the fuel efficiency and to improve the economic parameters of fuel cycles of the unit no. 5 reactor core of the Novovoronezh nuclear power plant.

Prytkov, A. N.; Tereshchenko, A. B.; Kravchenko, Yu. N.; Boldyrev, N. V.; Pozychaniuk, I. V.; Lisitsyn, D. I.; Golubev, E. I.

2014-04-01

347

Edge detection techniques for iris recognition system  

NASA Astrophysics Data System (ADS)

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

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

2013-12-01

348

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

349

Global shielding analysis for the three-element core advanced neutron source reactor under normal operating conditions  

SciTech Connect

Two-dimensional discrete ordinates radiation transport calculations were performed for a model of the three-element core Advanced Neutron Source reactor design under normal operating conditions. The core consists of two concentric upper elements and a lower element radially centered in the annulus between the upper elements. The initial radiation transport calculations were performed with the DORT two-dimensional discrete ordinates radiation transport code using the 39-neutron-group/44-gamma-ray-group ANSL-V cross-section library, an S{sub 6} quadrature, and a P{sub 1} Legendre polynomial expansion of the cross sections to determine the fission neutron source distribution in the core fuel elements. These calculations were limited to neutron groups only. The final radiation transport calculations, also performed with DORT using the 39-neutron-group/44-gamma-ray-group ANSL-V cross-section library, an S{sub l0} quadrature, and a P{sub 3} Legendre polynomial expansion of the cross sections, produced neutron and gamma-ray fluxes over the full extent of the geometry model. Responses (or activities) at various locations in the model were then obtained by folding the appropriate response functions with the fluxes at those locations. Some comparisons were made with VENTURE-calculated (diffusion theory) 20-group neutron fluxes that were summed into four broad groups. Tne results were in reasonably good agreement when the effects of photoneutrons were not included, thus verifying the physics model upon which the shielding model was based. Photoneutrons increased the fast-neutron flux levels deep within the D{sub 2}0 several orders of magnitude. Results are presented as tables of activity values for selected radial and axial traverses, plots of the radial and axial traverse data, and activity contours superimposed on the calculational geometry model.

Slater, C.O.; Bucholz, J.A.

1995-08-01

350

Three-Dimensional Full Core Power Calculations for Pressurized Water Reactors  

SciTech Connect

We have implemented a new multilevel parallel decomposition in the Denovo discrete ordinates radiation transport code. In concert with Krylov subspace iterative solvers, the multilevel decomposition allows concurrency over energy in addition to space-angle. The original space-angle partitioning in Denovo placed an eective limit on the scalability of the transport solver that was highly dependent on the problem size. The added phase-space concurrency combined with the high-performance Krylov solvers has enabled weak scaling to 100K cores on the Jaguar XT5 supercomputer. Furthermore, the multilevel decomposition provides enough concurrency to scale to exascale computing and beyond.

Evans, Thomas M [ORNL] [ORNL; Davidson, Gregory G [ORNL] [ORNL; Slaybaugh, Rachel N [ORNL] [ORNL

2010-01-01

351

ZPR-6 assembly 7 high {sup 240}Pu core experiments : a fast reactor core with mixed (Pu,U)-oxide fuel and a centeral high{sup 240}Pu zone.  

SciTech Connect

ZPR-6 Assembly 7 (ZPR-6/7) encompasses a series of experiments performed at the ZPR-6 facility at Argonne National Laboratory in 1970 and 1971 as part of the Demonstration Reactor Benchmark Program (Reference 1). Assembly 7 simulated a large sodium-cooled LMFBR with mixed oxide fuel, depleted uranium radial and axial blankets, and a core H/D near unity. ZPR-6/7 was designed to test fast reactor physics data and methods, so configurations in the Assembly 7 program were as simple as possible in terms of geometry and composition. ZPR-6/7 had a very uniform core assembled from small plates of depleted uranium, sodium, iron oxide, U{sub 3}O{sub 8} and Pu-U-Mo alloy loaded into stainless steel drawers. The steel drawers were placed in square stainless steel tubes in the two halves of a split table machine. ZPR-6/7 had a simple, symmetric core unit cell whose neutronic characteristics were dominated by plutonium and {sup 238}U. The core was surrounded by thick radial and axial regions of depleted uranium to simulate radial and axial blankets and to isolate the core from the surrounding room. The ZPR-6/7 program encompassed 139 separate core loadings which include the initial approach to critical and all subsequent core loading changes required to perform specific experiments and measurements. In this context a loading refers to a particular configuration of fueled drawers, radial blanket drawers and experimental equipment (if present) in the matrix of steel tubes. Two principal core configurations were established. The uniform core (Loadings 1-84) had a relatively uniform core composition. The high {sup 240}Pu core (Loadings 85-139) was a variant on the uniform core. The plutonium in the Pu-U-Mo fuel plates in the uniform core contains 11% {sup 240}Pu. In the high {sup 240}Pu core, all Pu-U-Mo plates in the inner core region (central 61 matrix locations per half of the split table machine) were replaced by Pu-U-Mo plates containing 27% {sup 240}Pu in the plutonium component to construct a central core zone with a composition closer to that in an LMFBR core with high burnup. The high {sup 240}Pu configuration was constructed for two reasons. First, the composition of the high {sup 240}Pu zone more closely matched the composition of LMFBR cores anticipated in design work in 1970. Second, comparison of measurements in the ZPR-6/7 uniform core with corresponding measurements in the high {sup 240}Pu zone provided an assessment of some of the effects of long-term {sup 240}Pu buildup in LMFBR cores. The uniform core version of ZPR-6/7 is evaluated in ZPR-LMFR-EXP-001. This document only addresses measurements in the high {sup 240}Pu core version of ZPR-6/7. Many types of measurements were performed as part of the ZPR-6/7 program. Measurements of criticality, sodium void worth, control rod worth and reaction rate distributions in the high {sup 240}Pu core configuration are evaluated here. For each category of measurements, the uncertainties are evaluated, and benchmark model data are provided.

Lell, R. M.; Morman, J. A.; Schaefer, R.W.; McKnight, R.D.; Nuclear Engineering Division

2009-02-23

352

[Iridoschisis, a special form of iris atrophy].  

PubMed

Iridoschisis is a rare degenerative disease characterized by the separation of the anterior iris stroma from the posterior layer. The anterior layer splits into strands, and the free ends float freely in the anterior chamber. We report the case of a 57-year-old man, in whom we incidentally discovered isolated unilateral iris atrophy. The patient had no history of the common causes of atrophy (herpes, pigment dispersion, ocular trauma, etc.). During follow-up, the atrophy gradually worsened, with an increase in the number and bilaterality of the lesions. Ultrasound biomicroscopy (UBM) and optical coherence tomography (OCT) of anterior chamber showed thinning of the anterior iris and cleavage of the iris into two layers, an imaging result which, to our knowledge, has not yet been reported in the literature. Familiarity with iridoschisis is important, due to its frequent association with glaucoma, so that appropriate screening can be carried out at the time of diagnosis and on follow-up. PMID:23261208

Agard, E; Malcles, A; El Chehab, H; Ract-Madoux, G; Swalduz, B; Aptel, F; Denis, P; Dot, C

2013-04-01

353

Securing Iris Recognition Systems Against Masquerade Attacks  

E-print Network

security enhancing approach. Keywords: Iris recognition, Inverse Biometrics, Security, Vulnerabilities towards improving the matching accuracy of biometric systems as assessed by various performance metrics.2 However, other aspects of a biometric system are relatively under explored. In particular, only recently

Autonoma de Madrid, Universidad

354

IRIS TOXICOLOGICAL REVIEW OF HEXAVALENT CHROMIUM  

EPA Science Inventory

EPA is conducting a peer review of the scientific basis supporting the human health hazard and dose-response assessment of hexavalent chromium that will appear on the Integrated Risk Information System (IRIS) database. ...

355

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...

356

Burn-up Analysis and Determination of Equilibrium Core Configuration for Tehran Research Reactor at 7.5 MW Power Level  

SciTech Connect

This technical report presents burn-up and in-core fuel management calculations to determine a configuration for the equilibrium core for Tehran Research Reactor (TRR) at upgraded power level of 7.5 MW. Two different equilibrium core configurations have been concluded at this stage of design analysis; one equilibrium core with neutron trap and one without neutron trap. According to the preliminary fuel management calculations for the core configuration consisting of 27 Fuel Elements at 7.5 MW rated power, considering burn-up analysis, up to two neutron trap locations could be introduced in the central parts of the equilibrium core with a cycle length equal to 12 days that would satisfy the operational conditions. For the equilibrium core without neutron trap, a core with cycle length equal to 15 days gives satisfactory results. To cross-check the results of the CITVAP diffusion calculations with a Monte Carlo code such as MCNP-4B, the number densities calculated by CITVAP for the burned-up core have been provided for MCNP-4B through an auxiliary code called WIMSND. The results obtained show good agreement between these two different schemes. (authors)

Afshar, Ebrahim; Shahidi, Alireza; Zaker, Mohammad [Reactor Research and Operation Department, Nuclear Research Center Atomic Energy Organization of Iran, North Kargar Ave., P.O. Box 14155-1339, Tehran (Iran, Islamic Republic of)

2004-07-01

357

Proton beam radiotherapy of iris melanoma  

Microsoft Academic Search

Purpose: To report on outcomes after proton beam radiotherapy of iris melanoma. Methods and Materials: Between 1993 and 2004, 88 patients with iris melanoma received proton beam radiotherapy, with 53.1 Gy in 4 fractions. Results: The patients had a mean age of 52 years and a median follow-up of 2.7 years. The tumors had a median diameter of 4.3 mm,

Bertil. Damato; Andrzej Kacperek; Mona Chopra; Martin A. Sheen; Ian R. Campbell; R. Douglas Errington

2005-01-01

358

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

359

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

360

Vertical TEC representation by IRI 2012 and IRI Plas models for European midlatitudes  

NASA Astrophysics Data System (ADS)

Vertical total electron content (vTEC) values computed using IRI-2012 and IRI Plas models have been compared with diurnal GPS vTEC data derived from European mid-latitude GPS station Potsdam. Comparative data-model analysis does not reveal good performance in vTEC representation. It was found that new extension of IRI model - IRI Plas - cannot represent correctly the vTEC variations over European midlatitudes and mainly overestimates GPS vTEC especially for low and moderate solar activity. In order to estimate the source of the data-model discrepancies, the case-study with detailed analysis of the model simulated electron density profiles was done. It was obtained that all models do not represent correctly the topside profile part and tend to overestimate the electron density higher than F2 peak. So, the main problem of the IRI vTEC representation is not situated in the plasmaspheric part, its absence in IRI model or its presence in IRI Plas model, the main source of the resulted discrepancies is still in the IRI topside ionosphere representation.

Zakharenkova, I. E.; Cherniak, Iu. V.; Krankowski, A.; Shagimuratov, I. I.

2015-04-01

361

Methodology for Identification of the Coolant Thermalhydraulic Regimes in the Core of Nuclear Reactors  

SciTech Connect

The paper deals with one of the acute for the nuclear energy problem of accident regimes of NPPs recognition diagnostics using noise signal diagnostics methodology. The methodology intends transformation of the random noise signals of the main technological parameters at the exit of a nuclear facility (neutron flow, dynamic pressure etc.) which contain the important information about the technical status of the equipment. The effective algorithms for identification of random processes wore developed. After proper transformation its were considered as multidimensional random vectors. Automatic classification of these vectors in the developed algorithms is realized on the basis of the probability function in particular Bayes classifier and decision functions. Till now there no mathematical models for thermalhydraulic regimes of fuel assemblies recognition on the acoustic and neutron noises parameters in the core of nuclear facilities. The two mathematical models for analysis of the random processes submitted to the automatic classification is proposed, i.e. statistical (using Bayes classifier of acoustic spectral density diagnosis signals) and geometrical (on the basis of formation in the featured space of dividing hyper-plane). The theoretical basis of the bubble boiling regimes in the fuel assemblies is formulated as identification of these regimes on the basis of random parameters of auto spectral density of acoustic noise (ASD) measured in the fuel assemblies (dynamic pressure in the upper plenum in the paper). The elaborated algorithms allow recognize realistic status of the fuel assemblies. For verification of the proposed mathematical models the analysis of experimental measurements was carried out. The research of the boiling onset and definition of the local values of the flow parameters in the seven-beam fuel assembly (length of 1.3 m, diameter of 6 mm) have shown the correct identification of the bubble boiling regimes. The experimental measurements on real WWER core assemblies were analysed as well. On the basis of model of Bayes classifier for bubble structure of two-phase flow in fuel assemblies of WWER-440 (intends usage of 28 dimensional accidental realizations of ASD of neutron noise) the reliable identification of the pointed regimes of fuel assemblies in WWERs up to 98% was obtained. On the basis of geometrical mathematical model of identification at essentially more limited volume of teaching sampling the recognition of ASD realizations of the neutron noise of the same both dimensions and quantity of the reliability of correct identification of these parameters was up to 92%. The recognition of the pointed thermalhydraulic parameters was carried out on the basis of experimental research of ASD of acoustic noise parameters of the experimental fuel assembly with electrically heated imitators using the two recognition models - statistical and geometrical. It confirmed high efficiency of the algorithms developed. The average reliability of identification of the first vapor bubbles activation regime at the heat transfer surface was not low then 90%. (authors)

Sharaevsky, L.G.; Sharaevskaya, E.I.; Domashev, E.D.; Arkhypov, A.P.; Kolochko, V.N. [Ministry of Fuel and Energy of Ukraine, 30, Khreshchatyk str., Kyiv 01601 (Ukraine)

2002-07-01

362

Nuclear reactor  

SciTech Connect

A nuclear reactor of the type which uses liquid metal as primary and secondary coolants, and wherein the reactor vessel contains a core and a plurality of vertically extending cylindrical intermediate heat exchangers for carrying out heat exchange between the primary and secondary coolants; primary coolant circulation pumps disposed outside of the reactor vessel; a pipe for conducting to the circulation pump the primary coolant which has passed through the intermediate heat exchangers after leaving the core; and a pipe for guiding the primary coolant discharged from the circulation pump to the core through the reactor vessel.

Sato, M.

1984-08-14

363

The Interface Region Imaging Spectrograph (IRIS)  

NASA Astrophysics Data System (ADS)

The solar chromosphere and transition region (TR) form a highly structured and dynamic interface region between the photosphere and the corona. This region not only acts as the conduit of all mass and energy feeding into the corona and solar wind, it also requires an order of magnitude more energy to heat than the corona. Nevertheless, the chromosphere remains poorly understood, because of the complexity of the required observational and analytical tools: the interface region is highly complex with transitions from optically thick to optically thin radiation, from pressure to magnetic field domination, and large density and temperature contrasts on small spatial scales. The Interface Region Imaging Spectrograph (IRIS) was selected for a NASA SMEX mission in 2009 and is scheduled to launch in early 2013. IRIS addresses critical questions: (1) Which types of non-thermal energy dominate in the chromosphere and beyond? (2) How does the chromosphere regulate mass and energy supply to the corona and heliosphere? (3) How do magnetic flux and matter rise through the lower atmosphere, and what role does flux emergence play in flares and mass ejections? These questions are addressed with a high-resolution near and far UV imaging spectrometer sensitive to emission from plasma at temperatures between 5,000 K and 10 MK. IRIS has a field-of-view of 120 arcsec, a spatial resolution of 0.4 arcsec, and velocity resolution of 0.5 km/s. The IRIS investigation includes a strong numerical modeling component based on advanced radiative MHD codes to facilitate interpretation of observations. We will describe the IRIS instrumentation and numerical modeling, and present the status of the IRIS observatory development. We will highlight some of the issues that IRIS observations can help resolve.

De Pontieu, B.; Title, A. M.; Lemen, J. R.; Wuelser, J.; Tarbell, T. D.; Schrijver, C.; Golub, L.; Kankelborg, C. C.; Hansteen, V. H.; Carlsson, M.

2012-12-01

364

The Interface Region Imaging Spectrograph (IRIS)  

NASA Astrophysics Data System (ADS)

The solar chromosphere and transition region (TR) form a highly structured and dynamic interface region between the photosphere and the corona. This region not only acts as the conduit of all mass and energy feeding into the corona and solar wind, it also requires an order of magnitude more energy to heat than the corona. Nevertheless, the chromosphere remains poorly understood, because of the complexity of the required observational and analytical tools: the interface region is highly complex with transitions from optically thick to optically thin radiation, from pressure to magnetic field domination, and large density and temperature contrasts on small spatial scales. The Interface Region Imaging Spectrograph (IRIS) was selected for a NASA SMEX mission in 2009 and is scheduled to launch on 26-June-2013 (with first light scheduled for mid July). IRIS addresses critical questions: (1) Which types of non-thermal energy dominate in the chromosphere and beyond? (2) How does the chromosphere regulate mass and energy supply to the corona and heliosphere? (3) How do magnetic flux and matter rise through the lower atmosphere, and what role does flux emergence play in flares and mass ejections? These questions are addressed with a high-resolution near and far UV imaging spectrometer sensitive to emission from plasma at temperatures between 5,000 K and 10 MK. IRIS has a field-of-view of 120 arcsec, a spatial resolution of 0.4 arcsec, and velocity resolution of 0.5 km/s. The IRIS investigation includes a strong numerical modeling component based on advanced radiative MHD codes to facilitate interpretation of observations. We describe the IRIS instrumentation and numerical modeling, and present the plans for observations, calibration and data distribution. We will highlight some of the issues that IRIS observations can help resolve. More information can be found at http://iris.lmsal.com

De Pontieu, Bart; Title, A. M.; Lemen, J.; Wuelser, J.; Tarbell, T. D.; Schrijver, C. J.; Golub, L.; Kankelborg, C.; Carlsson, M.; Hansteen, V. H.; Worden, S.; IRIS Team

2013-07-01

365

Nuclear reactor  

SciTech Connect

A nuclear reactor is described which comprises a reactor vessel, a core housed in the reactor vessel, an ultrasonic transducer mounted in the vicinity of the upper end of the core for emitting and receiving an ultrasonic wave pulse signal propagating above the core. A means is provided for rotating the transducer by a prescribed angle to scan horizontally the ultrasonic wave emitted from the transducer. A plurality of reflective means are mounted in the vicinity of the upper end of the core in a manner to face the transducer for reflecting the ultrasonic wave signal emitted from the transducer.

Furudate, M.; Miyazawa, T.; Mizuguchi, H.; Sasaki, K.; Uesugi, N.

1981-09-22

366

NUCLEAR REACTOR  

Microsoft Academic Search

High temperature reactors which are uniquely adapted to serve as the ; heat source for nuclear pcwered rockets are described. The reactor is comprised ; essentially of an outer tubular heat resistant casing which provides the main ; coolant passageway to and away from the reactor core within the casing and in ; which the working fluid is preferably hydrogen

Grebe

1959-01-01

367

Proton beam radiotherapy of iris melanoma  

SciTech Connect

Purpose: To report on outcomes after proton beam radiotherapy of iris melanoma. Methods and Materials: Between 1993 and 2004, 88 patients with iris melanoma received proton beam radiotherapy, with 53.1 Gy in 4 fractions. Results: The patients had a mean age of 52 years and a median follow-up of 2.7 years. The tumors had a median diameter of 4.3 mm, involving more than 2 clock hours of iris in 32% of patients and more than 2 hours of angle in 27%. The ciliary body was involved in 20%. Cataract was present in 13 patients before treatment and subsequently developed in another 18. Cataract had a 4-year rate of 63% and by Cox analysis was related to age (p = 0.05), initial visual loss (p < 0.0001), iris involvement (p < 0.0001), and tumor thickness (p < 0.0001). Glaucoma was present before treatment in 13 patients and developed after treatment in another 3. Three eyes were enucleated, all because of recurrence, which had an actuarial 4-year rate of 3.3% (95% CI 0-8.0%). Conclusions: Proton beam radiotherapy of iris melanoma is well tolerated, the main problems being radiation-cataract, which was treatable, and preexisting glaucoma, which in several patients was difficult to control.

Damato, Bertil [St. Paul's Eye Unit, Royal Liverpool University Hospital, Liverpool (United Kingdom)]. E-mail: Bertil@damato.co.uk; Kacperek, Andrzej [Douglas Cyclotron, Clatterbridge Centre for Oncology, Bebington, Wirral (United Kingdom); Chopra, Mona [Douglas Cyclotron, Clatterbridge Centre for Oncology, Bebington, Wirral (United Kingdom); Sheen, Martin A. [Douglas Cyclotron, Clatterbridge Centre for Oncology, Bebington, Wirral (United Kingdom); Campbell, Ian R. [IC Statistical Services, Wirral (United Kingdom); Errington, R. Douglas [Douglas Cyclotron, Clatterbridge Centre for Oncology, Bebington, Wirral (United Kingdom)

2005-09-01

368

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

369

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

370

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

371

Sector iris hemangioma in association with diffuse choroidal hemangioma.  

PubMed

Two patients referred for iris lesions were found to have sector hemangioma of the iris stroma in contiguity with diffuse choroidal hemangioma. Neither patient had other manifestations of Sturge-Weber syndrome. PMID:25727597

Shields, Carol L; Atalay, Hatice Tuba; Wuthisiri, Wadakarn; Levin, Alex V; Lally, Sara E; Shields, Jerry A

2015-02-01

372

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

373

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

374

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

375

Incorporated Research Institutions for Seismology (IRIS) Consortium  

NSDL National Science Digital Library

The Incorporated Research Institutions for Seismology (IRIS) is a consortium of 91 US Universities with research interests in seismology. IRIS was established in an effort "to develop and operate the infrastructure needed for the acquisition and distribution of high quality seismic data." One highlight is the Data Management System, which incorporates six data collection centers to coordinate data inflow from the 128 seismic recording stations that make up IRIS's Global Seismographic Network (GSN). Other features include a seismic monitor link for a quick view of current seismic activity. The SeismiQuery Database allows users to search out available data by day, month, station, event, and more. Further, a station book "contains information about stations from all networks that contribute data." Finally, this fine site also features Special Event Pages, an excellent collection of links to specific sites, graphics, and general information on recent earthquakes (see the September 1, 1999 Scout Report for Science & Engineering).

376

Tunable liquid iris actuated using electrowetting effect  

NASA Astrophysics Data System (ADS)

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

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

2014-05-01

377

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

378

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

379

Increased Iris-lens Contact Following Spontaneous Blinking: Mathematical Modeling  

PubMed Central

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

Amini, Rouzbeh; Jouzdani, Sara

2012-01-01

380

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...

381

Iris recognition system by using support vector machines  

Microsoft Academic Search

In recent years, with the increasing demands of security in our networked society, biometric systems for user verification are becoming more popular. Iris recognition system is a new technology for user verification. In this paper, the CASIA iris database is used for individual userpsilas verification by using support vector machines (SVMs) which based on the analysis of iris code as

Hasimah Ali; Momoh J. E. Salami; Wahyudi

2008-01-01

382

Light stimulation of iris tyrosinase in vivo. [Rabbits  

SciTech Connect

This paper presents evidence that light stimulates tyrosinase activity in iris melanocytes in rabbits. Levels of iris tyrosinase were found to be greater in eyes of rabbits exposed to light for 6 weeks than in eyes of rabbits maintained in darkness. Despite increasing tyrosinase levels, exposure to light produced no clinically observable change in iris color.

Dryja, T.P.; Kimball, G.P.; Albert, D.M.

1980-05-01

383

Cataract influence on iris recognition performance  

NASA Astrophysics Data System (ADS)

This paper presents the experimental study revealing weaker performance of the automatic iris recognition methods for cataract-affected eyes when compared to healthy eyes. There is little research on the topic, mostly incorporating scarce databases that are often deficient in images representing more than one illness. We built our own database, acquiring 1288 eye images of 37 patients of the Medical University of Warsaw. Those images represent several common ocular diseases, such as cataract, along with less ordinary conditions, such as iris pattern alterations derived from illness or eye trauma. Images were captured in near-infrared light (used in biometrics) and for selected cases also in visible light (used in ophthalmological diagnosis). Since cataract is a disorder that is most populated by samples in the database, in this paper we focus solely on this illness. To assess the extent of the performance deterioration we use three iris recognition methodologies (commercial and academic solutions) to calculate genuine match scores for healthy eyes and those influenced by cataract. Results show a significant degradation in iris recognition reliability manifesting by worsening the genuine scores in all three matchers used in this study (12% of genuine score increase for an academic matcher, up to 175% of genuine score increase obtained for an example commercial matcher). This increase in genuine scores affected the final false non-match rate in two matchers. To our best knowledge this is the only study of such kind that employs more than one iris matcher, and analyzes the iris image segmentation as a potential source of decreased reliability

Trokielewicz, Mateusz; Czajka, Adam; Maciejewicz, Piotr

2014-11-01

384

High flux reactor  

Microsoft Academic Search

A high flux nuclear reactor is described comprising: (a) a pressure vessel including reactor coolant inlet means at the first end thereof and reactor coolant outlet means at the second end thereof; (b) a reactor coolant; (c) a first core segment housed within the pressure vessel, the first core segment including a plurality of concentric, circumferential fuel plates, the spacing

J. A. Lake; R. L. Heath; J. L. Liebenthal; D. R. DeBoisblanc; C. F. Leyse; K. Parsons; J. M. Ryskamp; R. P. Wadkins; Y. D. Harker; G. N. Fillmore

1988-01-01

385

IVA2: A computer code for modeling of transient three-dimensional three phase, three component, flows using three velocity fields in cylindrical geometry with arbitrary internals including nuclear reactor PWR\\/BWR-core  

Microsoft Academic Search

Input data, contents of the COMMON blocks, and functions of the IVA2\\/001 routines are described. The nonformal description of the current IVA2\\/001 constitutive package and the reactor core model are given.

N. I. Kolev

1986-01-01

386

Irrigating iris retractor for complicated cataract surgery.  

PubMed

We describe an irrigating iris retractor for cataract surgery in eyes with small pupils. The retractor, a modified irrigating handpiece, has 2 standard side-port irrigating openings and a smooth, button-like iris hook in the front. The hook is inserted into the margin of a small pupil and retracts the pupil peripherally to allow visualization of cortical remnants in the equatorial area of the lens capsule. The capsule can then be safely cleaned using the aspirating handpiece in the surgeon's other hand. PMID:19251131

Böhm, Peter; Horvath, Jozef; Zahorcova, Miriam

2009-03-01

387

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

388

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

389

IRIS Resolving Unresolved Structure - Duration: 11 seconds.  

NASA Video Gallery

NASA’s IRIS, which is able to look at a low layer of the sun’s atmosphere in unprecedented resolution, reveals details in the bright loops seen over the sun’s limb that have never been witnessed be...

390

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...

391

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...

392

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...

393

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.

Tom Bridgman

2001-10-09

394

Pigment Melanin: Pattern for Iris Recognition  

Microsoft Academic Search

Recognition of iris based on visible light (VL) imaging is a difficult problem because of the light reflection from the cornea. Nonetheless, pigment melanin provides a rich feature source in VL, which is unavailable in near-infrared (NIR) imaging. This is due to the biological spectroscopy of eumelanin, a chemical not stimulated in NIR. In this case, a plausible solution to

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

2010-01-01

395

Nuclear reactor construction with bottom supported reactor vessel  

Microsoft Academic Search

This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side

Sharbaugh

1987-01-01

396

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

397

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

398

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

399

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

SciTech Connect

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, E-mail: d.setyawan@bapeten.go.id [Center for Assessment of Regulatory System and Technology for Nuclear Installations and Materials, Indonesian Nuclear Energy Regulatory Agency (BAPETEN), Jl. Gajah Mada No. 8 Jakarta 10120 (Indonesia); Rohman, Budi [Licensing Directorate for Nuclear Installations and Materials, Indonesian Nuclear Energy Regulatory Agency (BAPETEN), Jl. Gajah Mada No. 8 Jakarta 10120 (Indonesia)

2014-09-30

400

Coaxial optical structure for iris recognition from a distance  

NASA Astrophysics Data System (ADS)

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

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

2011-05-01

401

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

402

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

SciTech Connect

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., E-mail: gorpavel@vver.kiae.ru; Kotsarev, A. V.; Lizorkin, M. P. [National Research Center Kurchatov Institute (Russian Federation)

2014-12-15

403

Neutronics methods, models, and applications at the Idaho National Engineering Laboratory for the advanced neutron source reactor three-element core design  

SciTech Connect

A summary of the methods and models used to perform neutronics analyses on the Advanced Neutron Source reactor three-element core design is presented. The applications of the neutral particle Monte Carlo code MCNP are detailed, as well as the expansion of the static role of MCNP to analysis of fuel cycle depletion calculations. Results to date of these applications are presented also. A summary of the calculations not yet performed is also given to provide a {open_quotes}to-do{close_quotes} list if the project is resurrected.

Wemple, C.A.; Schnitzler, B.G.; Ryskamp, J.M. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

1995-08-01

404

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

405

The Interface Region Imaging Spectrograph (IRIS)  

NASA Astrophysics Data System (ADS)

The Interface Region Imaging Spectrograph (IRIS) small explorer spacecraft provides simultaneous spectra and images of the photosphere, chromosphere, transition region, and corona with 0.33 - 0.4 arcsec spatial resolution, two-second temporal resolution, and 1 km s-1 velocity resolution over a field-of-view of up to 175 arcsec × 175 arcsec. IRIS was launched into a Sun-synchronous orbit on 27 June 2013 using a Pegasus-XL rocket and consists of a 19-cm UV telescope that feeds a slit-based dual-bandpass imaging spectrograph. IRIS obtains spectra in passbands from 1332 - 1358 Å, 1389 - 1407 Å, and 2783 - 2834 Å, including bright spectral lines formed in the chromosphere (Mg ii h 2803 Å and Mg ii k 2796 Å) and transition region (C ii 1334/1335 Å and Si iv 1394/1403 Å). Slit-jaw images in four different passbands (C ii 1330, Si iv 1400, Mg ii k 2796, and Mg ii wing 2830 Å) can be taken simultaneously with spectral rasters that sample regions up to 130 arcsec × 175 arcsec at a variety of spatial samplings (from 0.33 arcsec and up). IRIS is sensitive to emission from plasma at temperatures between 5000 K and 10 MK and will advance our understanding of the flow of mass and energy through an interface region, formed by the chromosphere and transition region, between the photosphere and corona. This highly structured and dynamic region not only acts as the conduit of all mass and energy feeding into the corona and solar wind, it also requires an order of magnitude more energy to heat than the corona and solar wind combined. The IRIS investigation includes a strong numerical modeling component based on advanced radiative-MHD codes to facilitate interpretation of observations of this complex region. Approximately eight Gbytes of data (after compression) are acquired by IRIS each day and made available for unrestricted use within a few days of the observation.

De Pontieu, B.; Title, A. M.; Lemen, J. R.; Kushner, G. D.; Akin, D. J.; Allard, B.; Berger, T.; Boerner, P.; Cheung, M.; Chou, C.; Drake, J. F.; Duncan, D. W.; Freeland, S.; Heyman, G. F.; Hoffman, C.; Hurlburt, N. E.; Lindgren, R. W.; Mathur, D.; Rehse, R.; Sabolish, D.; Seguin, R.; Schrijver, C. J.; Tarbell, T. D.; Wülser, J.-P.; Wolfson, C. J.; Yanari, C.; Mudge, J.; Nguyen-Phuc, N.; Timmons, R.; van Bezooijen, R.; Weingrod, I.; Brookner, R.; Butcher, G.; Dougherty, B.; Eder, J.; Knagenhjelm, V.; Larsen, S.; Mansir, D.; Phan, L.; Boyle, P.; Cheimets, P. N.; DeLuca, E. E.; Golub, L.; Gates, R.; Hertz, E.; McKillop, S.; Park, S.; Perry, T.; Podgorski, W. A.; Reeves, K.; Saar, S.; Testa, P.; Tian, H.; Weber, M.; Dunn, C.; Eccles, S.; Jaeggli, S. A.; Kankelborg, C. C.; Mashburn, K.; Pust, N.; Springer, L.; Carvalho, R.; Kleint, L.; Marmie, J.; Mazmanian, E.; Pereira, T. M. D.; Sawyer, S.; Strong, J.; Worden, S. P.; Carlsson, M.; Hansteen, V. H.; Leenaarts, J.; Wiesmann, M.; Aloise, J.; Chu, K.-C.; Bush, R. I.; Scherrer, P. H.; Brekke, P.; Martinez-Sykora, J.; Lites, B. W.; McIntosh, S. W.; Uitenbroek, H.; Okamoto, T. J.; Gummin, M. A.; Auker, G.; Jerram, P.; Pool, P.; Waltham, N.

2014-07-01

406

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

E-print Network

In accordance with a 1986 NRC ruling, the MIT Research Reactor (MITR) is planning on converting from the use of highly enriched uranium (HEU) to low enriched uranium (LEU) for fuel. A component of the conversion analysis ...

Plumer, Kevin E. (Kevin Edward)

2011-01-01

407

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

Code of Federal Regulations, 2013 CFR

...such a model that affects the temperature calculation, the...such a model that affects the temperature calculation, the...from the chemical reaction of the cladding with...reactor coolant, at a rate in excess of...

2013-01-01

408

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

Code of Federal Regulations, 2014 CFR

...such a model that affects the temperature calculation, the...such a model that affects the temperature calculation, the...from the chemical reaction of the cladding with...reactor coolant, at a rate in excess of...

2014-01-01

409

78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors  

Federal Register 2010, 2011, 2012, 2013, 2014

...NUCLEAR REGULATORY COMMISSION...NRC-2012-0134] Initial Test Program of Emergency...Boiling-Water Reactors AGENCY: Nuclear Regulatory Commission...SUMMARY: The U.S. Nuclear Regulatory Commission...79.1, ``Initial Test Program of...

2013-10-24

410

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

SciTech Connect

A generalized non local-equilibrium, three-equation model was developed for the macroscopic description of two-phase flow heat and mass transfer in porous media subjected to phase change. Six pore-scale closure problems were proposed to determine all the effective transport coefficients for representative unit cells. An improved model is presented in this paper with the perspective of application to intense boiling phenomena. The objective of this paper is to present application of this model to the simulation of reflooding of severely damaged nuclear reactor cores. In case of accident at a nuclear power plant, water sources may not be available for a long period of time and the core heats up due to the residual power. Any attempt to inject water during core degradation can lead to quenching and further fragmentation of the core material. The fragmentation of fuel rods and melting of reactor core materials may result in the formation of a {sup d}ebris bed{sup .} The typical particle size in a debris bed might reach few millimeters (characteristic length-scale: 1 to 5 mm), which corresponds to a high permeability porous medium. The proposed two-phase flow model is implemented in the ICARECATHARE code, developed by IRSN to study severe accident scenarios in pressurized water reactors. Currently, the French IRSN has set up two experimental facilities to study debris bed reflooding, PEARL and PRELUDE, with the objective to validate safety models. The PRELUDE program studies the complex two phase flow of water and steam in a porous medium (diameter 180 mm, height 200 mm), initially heated to a high temperature (400 deg. C or 700 deg. C). The series of PRELUDE experiments achieved in 2010 constitute a significant complement to the database of high temperature bottom reflood experimental data. They provide relevant data to understand the progression of the quench front and the intensity of heat transfer. Modeling accurately these experiments required improvements to the reflooding model, especially in terms of the existence of various saturation regimes. The improved two-phase flow model shows a good agreement with PRELUDE experimental results.

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

2012-05-15

411

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

NASA Astrophysics Data System (ADS)

A generalized non local-equilibrium, three-equation model was developed for the macroscopic description of two-phase flow heat and mass transfer in porous media subjected to phase change. Six pore-scale closure problems were proposed to determine all the effective transport coefficients for representative unit cells. An improved model is presented in this paper with the perspective of application to intense boiling phenomena. The objective of this paper is to present application of this model to the simulation of reflooding of severely damaged nuclear reactor cores. In case of accident at a nuclear power plant, water sources may not be available for a long period of time and the core heats up due to the residual power. Any attempt to inject water during core degradation can lead to quenching and further fragmentation of the core material. The fragmentation of fuel rods and melting of reactor core materials may result in the formation of a "debris bed". The typical particle size in a debris bed might reach few millimeters (characteristic length-scale: 1 to 5 mm), which corresponds to a high permeability porous medium. The proposed two-phase flow model is implemented in the ICARECATHARE code, developed by IRSN to study severe accident scenarios in pressurized water reactors. Currently, the French IRSN has set up two experimental facilities to study debris bed reflooding, PEARL and PRELUDE, with the objective to validate safety models. The PRELUDE program studies the complex two phase flow of water and steam in a porous medium (diameter 180 mm, height 200 mm), initially heated to a high temperature (400°C or 700°C). The series of PRELUDE experiments achieved in 2010 constitute a significant complement to the database of high temperature bottom reflood experimental data. They provide relevant data to understand the progression of the quench front and the intensity of heat transfer. Modeling accurately these experiments required improvements to the reflooding model, especially in terms of the existence of various saturation regimes. The improved two-phase flow model shows a good agreement with PRELUDE experimental results.

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

2012-05-01

412

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

413

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

414

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

415

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

416

Photometry Analysis using Oukaimeden IRIS Data  

NASA Astrophysics Data System (ADS)

Since 1988, a Moroccan astronomers began a site evaluation of the Oukaimeden (2600m) in the mountain ATLAS chain. The site has been selected, in parallel, for installation of IRIS helioseismograph as one of the seven stations by the world. The Oukaimeden appeared to be a good site, when compared to other former observatories, in term of daytime sky transparency, photometry and extinction and ground climatology. We present here a results of one year (1997) clouds cover, extinction coefficient and transparency fluctuation's measurement, using a total solar intensity provided by IRIS helioseismograph. We also compared the results with the same works mad by Hill et al (1994) in the GONG site testing campaign and Benkhaldoun et al (1993) using a simple flux integration photometer.

Benkhaldoun, Z.; Siher, E.

417

Iris transillumination defect and microhyphema syndrome.  

PubMed

We present a previously undescribed delayed complication attributable to sulcus-fixated posterior chamber lenses with elliptical polypropylene haptics containing a 10 degrees anterior angulation. Clinical signs of this complication are crescent-shaped iris transillumination defects overlying the lens haptics in the peripheral iris; in some cases these are associated with single or recurrent visually significant microhyphemas. This series describes 41 eyes that contain these transillumination defects; eight of the eyes have had lens-induced intraocular hemorrhage. We estimate the overall incidence of transillumination defects in our sulcus-fixated posterior chamber lens patient population to be between 5% and 15%. Those patients who have had lens-induced hemorrhage represent slightly greater than 1%, which is higher than our incidence of cystoid macular edema or retinal detachment. It is important for all ophthalmologists to be aware of this syndrome in evaluating patients with posterior chamber lenses who present with a transient obscuration of vision. PMID:6094408

Johnson, S H; Kratz, R P; Olson, P F

1984-01-01

418

Opiate receptors in the rabbit iris  

Microsoft Academic Search

Evidence pointing to the presence of opiate receptors in the rabbit iris was obtained in an in vivo study of the effects on the pupil of the intraocular injection of morphine (10–100 µg) and d-Ala2-met-enkephalinamide (d-Ala-E) (5–50 µg). Both opiates induced a significant dose-dependent decrease in pupillary size and an appreciable fluctuation of pupillary size. I.v. administration of naloxone (0.5

Filippo Drago; Giovanni Gorgone; Francesco Spina; Giovambattista Panissidi; Alberto Dal Bello; Ferruccio Moro; Umberto Scapagnini

1980-01-01

419

Design of an IRI track system  

Microsoft Academic Search

This paper presents the algorithm of an IR image track system, its hardware and software design. TMS32020 (or C25), as the main tools of image processing, makes the IRI track experiment system closed-loop. Track algorithm and processing speed are tested in the dynamic tracking situation. The experiment results show that the system can track the target in real-time, the period

Lan Tao

1993-01-01

420

IRIS - A Community-Based Facility to Support Research in Seismology  

NASA Astrophysics Data System (ADS)

The IRIS Consortium was established in 1984 in response to growing pressure from the research community for enhanced facilities in global and lithospheric seismology. At the same time, the National Science Foundation was encouraging improvements in technology and infrastructure that were sorely needed to ensure the future health of the nation's research endeavors. The governance of IRIS and growth of the facility programs have been guided by strong involvement of the research community. The IRIS management governance and structure serves as an interface between the scientific community, funding agencies, and the programs of IRIS. The structure is designed to focus scientific talent on common objectives, to encourage broad participation, and to efficiently manage its programs. IRIS is governed by a Board of Directors consisting of representatives from each of IRIS' 99 member institutions. Operational policies are set by an Executive Committee elected by the Board of Directors. The Executive Committee, in turn, appoints members to the Planning Committee, the Program Coordination Committee, and the four Standing Committees that provide oversight of the Global Seismographic Network (GSN), the Program of Array Seismic Studies of the Continental Lithosphere (PASSCAL), the Data Management System (DMS), and the Education and Outreach Program (E&O). In addition, special advisory committees and ad hoc working groups are convened for special tasks. Development of the IRIS programs has rested on strong core support from the Instrumentation and Facilities Program of the National Science Foundation's Earth Science Division, augmented with funding from the Department of Defense, Department of Energy, member universities, and private organizations. Close collaboration with the US Geological Survey and other national and foreign institutions has greatly extended the geographical coverage and strengthened the intellectual input that is essential to guiding the evolution of the IRIS programs. In the 18 years since the founding of IRIS, the core programs have grown to meet most of the original design goals and the Consortium continues to evolve in response to the community's changing needs. The GSN has now 126 permanent seismic stations distributed throughout the world with real-time connectivity to nearly 90 sites and dial-up links to most others. In addition to seismometers, microbarographs are installed at 19 sites and GPS instrumentation is located at 16 sites. The PASSCAL program supports between 50 and 60 experiments per year, from a lending pool of 250 broadband seismic sensor systems, and over 800 higher frequency systems for active source experiments. All seismic data acquired under the GSN and PASSCAL programs are made openly and freely available to anyone on the Internet, through the DMS. The DMS currently receives over 6Tb per year, and is able to service most requests for data within hours. Shipments in 2002 serviced nearly 60,000 requests, comprising nearly 1Tb of data, made to 563 different seismologists from 145 institutions in 33 countries around the world. The E&O program is relatively young, yet is making considerable inroads through its museum partnership (reaching 8,000,000 people per year); distribution of inexpensive seismographs; development of teaching modules and other educational materials for schools; technical support to internet-enabled school-based networks; workshops for geoscience educators; and undergraduate summer internships.

Ingate, S.; Ahern, T.; Butler, R.; Fowler, J.; Simpson, D.; Taber, J.; van der Vink, G.

2002-12-01

421

A standardized approach for iris color determination.  

PubMed

Latanoprost, the phenyl-substituted prostaglandin F2alpha, has been found to be an effective agent for glaucoma therapy. This prostaglandin derivative exerts ocular hypotensive activity but is also associated with an untoward side effect, namely iris color changes. Latanoprost provoked iris color changes in cynomolgus monkeys and in multicenter clinical trials. Until now photographs were taken and compared with color plates to document these changes. The disadvantage of this method is obvious, i.e., the color luminance varies between measurements due to changes in the developer. Furthermore, subjective comparison of color changes relative to color plates rendered judgment subject to impression and opinion rather than to objective data. Therefore, a computerized method using a 3-CCD video camera attached to a slit lamp was developed. The signals were transferred to a computer and a single frame, which was "frozen" by means of a "grabber card." Camera and the computer had previously been calibrated and color plates were measured to check the standard conditions. They were evaluated by a software program displaying average color (as red, green, and blue values) of the selected area. This method provides a fast and accurate way to quantify color changes in the iris of both experimental animals and clinical trials. PMID:12573949

Niggemann, Birgit; Weinbauer, Gerhard; Vogel, Friedhelm; Korte, Rainhart

2003-01-01

422

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

423

HyspIRI Low Latency Concept and Benchmarks  

NASA Technical Reports Server (NTRS)

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

Mandl, Dan

2010-01-01

424

Linking genotype to phenotype: the International Rice Information System (IRIS)  

Microsoft Academic Search

The International Rice Information System (IRIS, http:\\/\\/www.iris.irri.org) is the rice implementation of the International Crop Information System (ICIS, http:\\/\\/www.icis.cgiar.org), a database system for the management and integration of global information on genetic resources and germplasm improvement for any crop. Building upon the germplasm genealogy and field data components of ICIS, IRIS is being extended to handle diverse rice genomics data

Richard M. Bruskiewich; Alexander B. Cosico; William Eusebio; Arllet M. Portugal; Luralyn M. Ramos; Ma. Teresa Reyes; May Ann B. Sallan; Victor Jun M. Ulat; Xusheng Wang; Kenneth L. Mcnally; Ruaraidh Sackville Hamilton; Christopher Graham Mclaren

2003-01-01

425

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

426

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

427

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

428

A Novel Method To On-Line Monitor Reactor Nuclear Power And In-Core Thermal Environments  

Microsoft Academic Search

For current nuclear power plants, nuclear power can not be directly measured and in-core fuel thermal environments can not be monitored due to the unavailability of an appropriate measurement technology and the inaccessibility of the fuel. If the nuclear deposited power and the in-core thermal conditions (i.e. fuel or coolant temperature and heat transfer coefficient) can be monitored in-situ, then

Hanying Liu; Don W. Miller; Dongxu Li; Thomas D. Radcliff

2002-01-01

429

An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the Molten Salt Fast Reactor  

NASA Astrophysics Data System (ADS)

In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR). This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation. The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution. Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions. The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.

Aufiero, M.; Cammi, A.; Fiorina, C.; Leppänen, J.; Luzzi, L.; Ricotti, M. E.

2013-10-01

430

PRA-Based SMA: the First Tool toward a Risk-Informed Approach to the Seismic Design of the IRIS  

Microsoft Academic Search

International Reactor Innovative and Secure (IRIS) is an advanced, modular, medium-power PWR with an integral primary system layout. As part of the “safety-by-design_” philosophy that inspired the project from the very beginning, a risk-informed approach to its design phase is being adopted and a probabilistic risk assessment (PRA) is being used as an active tool in pursuing an advanced level

Yuji KUMAGAI; Andrea MAIOLI; Marco E. RICOTTI; Hisashi NINOKATA; Mario D. CARELLI

2007-01-01

431

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

432

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

Microsoft Academic Search

A new computer code called MATADOR (Methods for the Analysis of Transport And Deposition Of Radionuclides) has been developed to replace the CORRAL computer code which was written for the Reactor Safety Study (WASH-1400). This report contains a detailed description of the models used in MATADOR. MATADOR is intended for use in system risk studies to analyze radionuclide transport and

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

1985-01-01

433

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

Microsoft Academic Search

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 ZrO2 as inert matrix. Adding a burnable poison to such a fuel proves to be necessary.

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

1999-01-01

434

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

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

Effect of lens implants on iris fluorescein angiography and the iris pigment layer.  

PubMed

Fourteen eyes underwent cataract operation and implantation of an iris fixated (Medallion) pseudophakos. Eleven had an uncomplicated per- and post-operative course (group 1), 3 had a subluxation of the lens implant (group 2). The control group (group 3) consisted of the 14 contralateral eyes from group 1 and 2. Two cadaver eyes had a similar lens implanted (group 4). Groups 1-3 were investigated by bilateral simultaneous iris fluorescein angiography and retroiridal stereo transillumination photography. In group 1, one case with pupillary and diffuse semiperipheral leakage areas was observed, in a clinically healthy eye 15 months after the operation. An other one had some leakage at the pupillary border, but not in the area of contact with the loops. Rubbing between iris surface and some parts of the lens did not provoke local leakage. No new vessel formation was found. Pigment layer defects were observed in every operated eye, but no leakage was related to them except in the two pupillary areas mentioned. Mayor pigment defects did not progress in a quiet eye. The iris seems to be to some degree resistant to contact with the parts of the intraocular lens implant. PMID:4050356

Krause, U

1985-08-01

437

Direct Attacks Using Fake Images in Iris Verification  

E-print Network

if an impostor "steals" it). Moreover, biometric systems are vulner- able to external attacks which could points to biometric recogni- tion systems. These vulnerability points, depicted in Figure 1, can broadly, the vulnerabilities of iris-based recogni- tion systems to direct attacks are studied. A database of fake iris images

Autonoma de Madrid, Universidad

438

From the Iriscode to the Iris: A New Vulnerability  

E-print Network

vulnerabilities · Competitions: · Standards: 5/43 #12;2. Biometrics #12;Biometric systems 7/43 #12;BiometricFrom the Iriscode to the Iris: A New Vulnerability Of Iris Recognition Systems Javier Galbally to bypass 8/43 #12;Attacks to Biometric Systems · Possible points of attack to a biometric system. Sensor

Autonoma de Madrid, Universidad

439

The IRIS network site at the Wilcox Solar Observatory  

NASA Technical Reports Server (NTRS)

The site for the International Research on the Interior of the Sun (IRIS) instrument housed at the Wilcox Solar Observatory at Stanford University (near San Francisco, USA) is described together with the instrument operation procedure. The IRIS instrument, which measures global oscillations of the sun, operates continuously every clear day since it was installed in August 1987.

Hoeksema, J. T.; Scherrer, P. H.

1991-01-01

440

Comparison of IRI2001 With TOPEX TEC Measurements  

Microsoft Academic Search

The International Reference Ionosphere (IRI) is an international joint project of the Committee on Space Research (COSPAR) and the International Union of Radio Science (URSI). As one of the most comprehensive empirical models of the ionosphere, the IRI provides the electron density, electron temperature, ion temperature, and ion composition in the altitude range from about 50 km to 2000 km,

G. Jee; R. W. Schunk; L. Scherliess

2003-01-01