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1

Generic small modular reactor plant design.  

SciTech Connect

This report gives an overview of expected design characteristics, concepts, and procedures for small modular reactors. The purpose of this report is to provide those who are interested in reducing the cost and improving the safety of advanced nuclear power plants with a generic design that possesses enough detail in a non-sensitive manner to give merit to their conclusions. The report is focused on light water reactor technology, but does add details on what could be different in a more advanced design (see Appendix). Numerous reactor and facility concepts were used for inspiration (documented in the bibliography). The final design described here is conceptual and does not reflect any proposed concept or sub-systems, thus any details given here are only relevant within this report. This report does not include any design or engineering calculations.

Lewis, Tom Goslee,; Cipiti, Benjamin B.; Jordan, Sabina Erteza; Baum, Gregory A.

2012-12-01T23:59:59.000Z

2

Distributed computing and nuclear reactor analysis  

SciTech Connect

Large-scale scientific and engineering calculations for nuclear reactor analysis can now be carried out effectively in a distributed computing environment, at costs far lower than for traditional mainframes. The distributed computing environment must include support for traditional system services, such as a queuing system for batch work, reliable filesystem backups, and parallel processing capabilities for large jobs. All ANL computer codes for reactor analysis have been adapted successfully to a distributed system based on workstations and X-terminals. Distributed parallel processing has been demonstrated to be effective for long-running Monte Carlo calculations.

Brown, F.B.; Derstine, K.L.; Blomquist, R.N.

1994-03-01T23:59:59.000Z

3

LIMITED POWER BURSTS IN DISTRIBUTED MODELS OF NUCLEAR REACTORS  

E-Print Network (OSTI)

LIMITED POWER BURSTS IN DISTRIBUTED MODELS OF NUCLEAR REACTORS M. V. Bazhenov and E. F. Sabaev UDC employed for analyzing reactor dynamics. Equations of this type are used for analyzing the stability of the reactor power, etc. Among these problems the question of the boundedness of reactor power bursts

Bazhenov, Maxim

4

Apparatus and method for quantitative assay of generic transuranic wastes from nuclear reactors  

DOE Patents (OSTI)

A combination of passive and active neutron measurements which yields quantitative information about the isotopic composition of transuranic wastes from nuclear power or weapons material manufacture reactors is described. From the measurement of prompt and delayed neutron emission and the incidence of two coincidentally emitted neutrons from induced fission of fissile material in the sample, one can quantify /sup 233/U, /sup 235/U and /sup 239/Pu isotopes in waste samples. Passive coincidence counting, including neutron multiplicity measurement and determination of the overall passive neutron flux additionally enables the separate quantitative evaluation of spontaneous fission isotopes such as /sup 240/Pu, /sup 244/Cm and /sup 252/Cf, and the spontaneous alpha particle emitter /sup 241/Am. These seven isotopes are the most important constituents of wastes from nuclear power reactors and once the mass of each isotope present is determined by the apparatus and method of the instant invention, the overall alpha particle activity can be determined to better than 1 nCi/g from known radioactivity data. Therefore, in addition to the quantitative analysis of the waste sample useful for later reclamation purposes, the alpha particle activity can be determined to decide whether permanent low-level burial is appropriate for the waste sample.

Caldwell, J.T.; Kunz, W.E.; Atencio, J.D.

1982-03-31T23:59:59.000Z

5

Quantification of model mismatch errors of the dynamic energy distribution in a stirred-tank reactor  

E-Print Network (OSTI)

QUANTIFICATION OF MODEL MISMATCH ERRORS OF THE DYNAMIC ENERGY DISTRIBUTION IN A STIRRED- TANK REACTOR A Thesis by MARK RAYMOND KIMMICH Submitted to the Graduate College of Texas AkM University in partial fulfillment of the requirement... for the degree of MASTER OF SCIENCE August 198i Major Subject: Chemical Engineering QUANTIFICATION OF MODEL MISMATCH ERRORS OF THE DYNAMIC ENERGY DISTRIBUTION IN A STIRRED-TANK REACTOR A Thesis by MARK RAYMOND KIMMICH Approved as to style and content by...

Kimmich, Mark Raymond

1987-01-01T23:59:59.000Z

6

Residence Time Distribution Measurement and Analysis of Pilot-Scale Pretreatment Reactors for Biofuels Production: Preprint  

SciTech Connect

Measurement and analysis of residence time distribution (RTD) data is the focus of this study where data collection methods were developed specifically for the pretreatment reactor environment. Augmented physical sampling and automated online detection methods were developed and applied. Both the measurement techniques themselves and the produced RTD data are presented and discussed.

Sievers, D.; Kuhn, E.; Tucker, M.; Stickel, J.; Wolfrum, E.

2013-06-01T23:59:59.000Z

7

Fabrication Flaw Density and Distribution In Repairs to Reactor Pressure Vessel and Piping Welds  

SciTech Connect

The Pacific Northwest National Laboratory is developing a generalized fabrication flaw distribution for the population of nuclear reactor pressure vessels and for piping welds in U.S. operating reactors. The purpose of the generalized flaw distribution is to predict component-specific flaw densities. The estimates of fabrication flaws are intended for use in fracture mechanics structural integrity assessments. Structural integrity assessments, such as estimating the frequency of loss-of-coolant accidents, are performed by computer codes that require, as input, accurate estimates of flaw densities. Welds from four different reactor pressure vessels and a collection of archived pipes have been studied to develop empirical estimates of fabrication flaw densities. This report describes the fabrication flaw distribution and characterization in the repair weld metal of vessels and piping. This work indicates that large flaws occur in these repairs. These results show that repair flaws are complex in composition and sometimes include cracks on the ends of the repair cavities. Parametric analysis using an exponential fit is performed on the data. The relevance of construction records is established for describing fabrication processes and product forms. An analysis of these records shows there was a significant change in repair frequency over the years when these components were fabricated. A description of repair flaw morphology is provided with a discussion of fracture mechanics significance. Fabrication flaws in repairs are characterized using optimized-access, high-sensitivity nondestructive ultrasonic testing. Flaw characterizations are then validated by other nondestructive evaluation techniques and complemented by destructive testing.

GJ Schuster, FA Simonen, SR Doctor

2008-04-01T23:59:59.000Z

8

Multi-Site Capacity, Production and Distribution Planning with Reactor Modifications: MILP Model, Bi-level Decomposition  

E-Print Network (OSTI)

-site production facilities to meet the demands of multiple geographically distributed markets. Potential capacityMulti-Site Capacity, Production and Distribution Planning with Reactor Modifications: MILP Model mixed-integer linear programming (MILP) model for the simultaneous capacity, production and distribution

Grossmann, Ignacio E.

9

Spatial Distribution of Total, Ammonia-Oxidizing, and Denitrifying Bacteria in Biological Wastewater Treatment Reactors for Bioregenerative Life Support  

Science Journals Connector (OSTI)

...bacteria perform recycling of various elements...fixed-film biological wastewater treatment reactors...The high recycling rate provided...distribution of wastewater throughout the...treating and recycling wastewater for consumption...

Yuko Sakano; Karen D. Pickering; Peter F. Strom; Lee J. Kerkhof

2002-05-01T23:59:59.000Z

10

A Receding Horizon Controller with a Parameter Estimator for Nuclear Reactor Power Distribution  

SciTech Connect

A receding horizon control method is used to solve on-line, at each time step, an optimization problem for a finite future interval and to implement the first optimal control input as the current control input. The receding horizon control method is combined with a parameter estimator to overcome the problems of the linear modeling and time-varying characteristics of a process. It is a suitable control strategy for time-varying systems, in particular, because the parameter estimator identifies a controller design model recursively at each time step, and also the receding horizon controller recalculates an optimal input at each time step by using newly measured signals. The proposed controller is applied to the axial power distribution control in a pressurized water reactor. The reactor dynamics model used for computer simulations is a two-point xenon oscillation model in which the reactor core is axially divided into two regions (upper and lower halves) and each region is assumed to have a single input and a single output and to be coupled with the other region. It is shown from numerical simulations that the proposed controller exhibits very fast tracking responses due to the step and ramp changes of axial target shape and also works well in a time-varying parameter condition.

Na, Man Gyun; Jung, Dong Won; Lee, Sun Mi [Paul Scherrer Institute (Switzerland)

2004-09-15T23:59:59.000Z

11

Design of a Receding Horizon Control System for Nuclear Reactor Power Distribution  

SciTech Connect

A receding horizon control method is applied to the axial power distribution control in a pressurized water reactor. The basic concept of receding horizon control is to solve on-line, at each sampling instant, an optimization problem for a finite future and to implement the first optimal control input as the current control input. Thus, it is a suitable control strategy for time-varying systems. The reactor model used for computer simulations is a two-point xenon oscillation model based on the nonlinear xenon and iodine balance equations and a one-group, one-dimensional, neutron diffusion equation with nonlinear power reactivity feedback that adequately describes axial oscillations and treats the nonlinearities explicitly. The reactor core is axially divided into two regions, and each region has one input and one output and is coupled with the other region. Through numerical simulations, it is shown that the proposed control algorithm exhibits very fast tracking responses due to the step and ramp changes of axial target shape and also works well in a time-varying parameter condition.

Na, Man Gyun [Chosun University (Korea, Republic of)

2001-07-15T23:59:59.000Z

12

Radiation safety assessment of a system of small reactors for distributed energy  

Science Journals Connector (OSTI)

......of safety or reliability. On the other...said that marine reactors are appropriate...300 MW thermal reactor power was used. The analysis for the normal...9 108 Sv h1. Analysis of effective...immediately after reactor shutdown was......

N. Odano; T. Ishida

2005-12-20T23:59:59.000Z

13

Standard for the determination of steady-state neutron reaction-rate distributions and reactivity of nuclear power reactors  

SciTech Connect

American National Standard ANSI/ANS*-19.3-2005 [1] covers 'The Determination of Steady-State Neutron Reaction-Rate Distributions and Reactivity of Nuclear Power Reactors'. The 2005 version is a new revision of this Standard, which had previously been issued in 1995. In this revision, the sections on the various types of power reactors have been updated to cover the latest methodologies of calculation in current use, and a section on HWR [CANDU{sup R}] reactors has been added. Also, the sections on verification and validation were revised to more fully define, discuss, and distinguish between these topics, and describe actions related to them. (authors)

Rouben, B. [Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga, Ont. L5K 1B2 (Canada)

2006-07-01T23:59:59.000Z

14

Efficient Generation of Generic Entanglement  

E-Print Network (OSTI)

We find that generic entanglement is physical, in the sense that it can be generated in polynomial time from two-qubit gates picked at random. We prove as the main result that such a process generates the average entanglement of the uniform (Haar) measure in at most $O(N^3)$ steps for $N$ qubits. This is despite an exponentially growing number of such gates being necessary for generating that measure fully on the state space. Numerics furthermore show a variation cut-off allowing one to associate a specific time with the achievement of the uniform measure entanglement distribution. Various extensions of this work are discussed. The results are relevant to entanglement theory and to protocols that assume generic entanglement can be achieved efficiently.

R. Oliveira; O. C. O. Dahlsten; M. B. Plenio

2007-04-03T23:59:59.000Z

15

A generic material flow control model applied in two industrial sectors  

Science Journals Connector (OSTI)

This paper addresses the problem of generic planning and control of automated material handling systems (AMHSs). We build upon previous work to provide a proof of concept for generic control of AMHSs in different domains. We present a generic control ... Keywords: Automated material handling systems (AMHSs), Baggage Handling, Distribution, Generic control architecture, Real-time scheduling

S. W. A. Haneyah; P. C. Schuur; J. M. J. Schutten; W. H. M. Zijm

2013-08-01T23:59:59.000Z

16

Radiation safety assessment of a system of small reactors for distributed energy  

Science Journals Connector (OSTI)

......condition that the reactor was operated at full...Coolant water 55.0 Reactor vessel 30.0 Containment...at utilisation of nuclear energy other than electricity generation(3). Two PSRD units of 100 MW thermal reactor power are located......

N. Odano; T. Ishida

2005-12-20T23:59:59.000Z

17

Generic programming in Scala  

E-Print Network (OSTI)

Generic programming is a programming methodology that aims at producing reusable code, defined independently of the data types on which it is operating. To achieve this goal, that particular code must rely on a set of requirements known as concepts...

N'guessan, Olayinka

2007-04-25T23:59:59.000Z

18

Generic Deep Geologic Disposal Safety Case | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Deep Geologic Disposal Safety Case Deep Geologic Disposal Safety Case Generic Deep Geologic Disposal Safety Case The Generic Deep Geologic Disposal Safety Case presents generic information that is of use in understanding potential deep geologic disposal options in the U.S. for used nuclear fuel (UNF) from reactors and high-level radioactive waste (HLW). Potential disposal options include mined disposal in a variety of geologic media (e.g., salt, shale, granite), and deep borehole disposal in basement rock. The Generic Safety Case is intended to be a source of information to provide answers to questions that may arise as the U.S. works to develop strategies to dispose of current and future inventories of UNF and HLW. DOE is examining combinations of generic geologic media and facility designs that could potentially support

19

High-temperature reactor fuel fission product release and distribution at 1600 to 1800 degrees C  

SciTech Connect

The essential feature of small, modular high-temperature reactors (HTRs) is the inherent limitation in maximum accident temperature to below 1600{degrees} C combined with the ability of coated particle fuel to retain all safety-relevant fission products under these conditions. To demonstrate this ability, spherical fuel elements with modern TRISO particles are irradiated and subjected to heating tests. Even after extended heating times at 1600{degrees} C, fission product release does not exceed the already low values projected for normal operating conditions. In this paper details of fission product distribution within spherical fuel elements heated at constant temperatures of 1600, 1700, and 1800{degrees} C are presented. The measurements confirm the silicon carbide (SiC) coating layer as the most important fission product barrier up to 1800{degrees} C. If the SiC fails (or is defective), the following transport properties at 1600 to 1800{degrees} C can be observed; cesium shows the fastest release from the UO{sub 2} kernel but is highly sorbed in the buffer layer of the particle and in the matrix graphite of the sphere; strontium is retained strongly both in the UO{sub 2} kernels and in matrix graphite, but can penetrate SiC in some cases where cesium is still completely retained; only if all coating layers are breached can iodine and noble gases be released. For the first 100 h at 1600{degrees} C (enveloping all possible accident scenarios of small HTRs), these fission products are almost completely retained in the coated particles.

Schenk, W.; Nabielek, H. (Forschungszentrum Juelich, Postfach 1913, W-5170 Juelich (DE))

1991-12-01T23:59:59.000Z

20

An interpretation of information gained from residence time distribution studies for operation of biological reactors  

E-Print Network (OSTI)

comparison of the same two models, that there was no definitive parameter by which to choose; the single exception being a direct comparison of the predicted to the actual conversions from operating reactors. Comparison itself is a formidable and tinre.... (May 1971) Marlow Lee Dodge, B. A. , Rockford College Directed by: Dr. Robert L. Irvine Most rational designs of biological reactors include the use of mass balances and an assumption of a particular hydraulic descrip- tion such as plug...

Dodge, Marlow Lee

2012-06-07T23:59:59.000Z

Note: This page contains sample records for the topic "reactor generic distributed" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

Determination of the relative power density distribution in a heterogeneous reactor from the results of measurements of the reactivity effects and the neutron importance function  

SciTech Connect

A method for experimental determination of the relative power density distribution in a heterogeneous reactor based on measurements of fuel reactivity effects and importance of neutrons from a californium source is proposed. The method was perfected on two critical assembly configurations at the NARCISS facility of the Kurchatov Institute, which simulated a small-size heterogeneous nuclear reactor. The neutron importance measurements were performed on subcritical and critical assemblies. It is shown that, along with traditionally used activation methods, the developed method can be applied to experimental studies of special features of the power density distribution in critical assemblies and reactors.

Bobrov, A. A.; Glushkov, E. S.; Zimin, A. A.; Kapitonova, A. V.; Kompaniets, G. V.; Nosov, V. I., E-mail: rpp@adis.vver.kiae.ru; Petrushenko, R. P.; Smirnov, O. N. [Russian Research Centre Kurchatov Institute (Russian Federation)

2012-12-15T23:59:59.000Z

22

Development of Generic Background  

NLE Websites -- All DOE Office Websites (Extended Search)

Kentucky Guidance for Ambient Background Assessment Kentucky Guidance for Ambient Background Assessment January 8, 2004 Natural Resources and Environmental Protection Cabinet Introduction This guidance document is intended to assist in comparing site data and background data for sites undergoing environmental assessment. These procedures provide a simplified statistical procedure for determining if the site data is part of the background population. It also provides generic statewide background values for inorganic chemicals that may be used in lieu of collecting site-specific background samples. The statistical procedures may be used for site- specific data or the generic statewide values in Tables 1 and 2. This guidance does not preclude other appropriate statistical comparisons from being made, but rather a simplified screening

23

Cosmogony of Generic Structures  

E-Print Network (OSTI)

The problem of formation of generic structures in the Universe is addressed, whereby first the kinematics of inertial continua for coherent initial data is considered. The generalization to self--gravitating continua is outlined focused on the classification problem of singularities and metamorphoses arising in the density field. Self--gravity gives rise to an internal hierarchy of structures, and, dropping the assumption of coherence, also to an external hierarchy of structures dependent on the initial power spectrum of fluctuations.

T. Buchert

1994-12-19T23:59:59.000Z

24

Radiation safety assessment of a system of small reactors for distributed energy  

Science Journals Connector (OSTI)

......inventory (t) 19.8 Fuel Outer diameterpitch...in UO2 No. of fuel assembles 69 Control...3014.0 Reactor vessel Inner diameterheight...of the energy consumption area, aiming...For design of a fuel exchange facility...of the energy consumption area because no...Development of in-vessel type control rod......

N. Odano; T. Ishida

2005-12-20T23:59:59.000Z

25

Generation -IV Reactor Concepts  

NLE Websites -- All DOE Office Websites (Extended Search)

Generation-IV Reactor Concepts Generation-IV Reactor Concepts Thomas H. Fanning Argonne National Laboratory 9700 South Cass Avenue Argonne, Illinois 60439, USA The Generation-IV International Forum (GIF) is a multi-national research and development (R&D) collaboration. The GIF pursues the development of advanced, next generation reactor technology with goals to improve: a) sustainability (effective fuel utilization and minimization of waste) b) economics (competitiveness with respect to other energy sources) c) safety and reliability (e.g., no need for offsite emergency response), and d) proliferation resistance and physical protection The GIF Technology Roadmap exercise selected six generic systems for further study: the Gas- cooled Fast Reactor (GFR), the Lead-cooled Fast Reactor (LFR), the Molten Salt Reactor (MSR),

26

Study of power distribution in the CZP, HFP and normal operation states of VVER-1000 (Bushehr) nuclear reactor core by coupling nuclear codes  

Science Journals Connector (OSTI)

Abstract In this research, the simulation of one-sixth of VVER-1000 (Bushehr) reactor core is carried out by WIMS-D4 nuclear code, based on symmetry of core and also by information obtained from FSAR. The cross sections of some nuclides are obtained by WIMS-D4 from the beginning of cycle (BOC) to the end of cycle (EOC), and they are transferred into the CITATION code as inputs. In the next stage, the amounts of neutron fluxes and power of reactor core are obtained by CITATION code in the CZP and HFP states. Then, the received products are returned again into the extended program cycle, thereby distributions of neutron fluxes and power are finally depicted. In the meantime, the space distribution of neutron fluxes and power throughout the core are presented during the normal operation by this simulation. It can be inferred that if the reactor operation continues, a flat power distribution will be made in the reactor core that might cause maximum power.

Mohsen Rafiei Karahroudi; Seyed Alireza Mousavi Shirazi

2015-01-01T23:59:59.000Z

27

Design of generic coal conversion facilities: Process release---Direct coal liquefaction  

SciTech Connect

The direct liquefaction portion of the PETC generic direct coal liquefaction process development unit (PDU) is being designed to provide maximum operating flexibility. The PDU design will permit catalytic and non-catalytic liquefaction concepts to be investigated at their proof-of-the-concept stages before any larger scale operations are attempted. The principal variations from concept to concept are reactor configurations and types. These include thermal reactor, ebullating bed reactor, slurry phase reactor and fixed bed reactor, as well as different types of catalyst. All of these operating modes are necessary to define and identify the optimum process conditions and configurations for determining improved economical liquefaction technology.

Not Available

1991-09-01T23:59:59.000Z

28

Taint-Exchange: a Generic System for Cross-process and Cross-host Taint Tracking  

E-Print Network (OSTI)

Taint-Exchange: a Generic System for Cross-process and Cross-host Taint Tracking Angeliki Zavou also utilized to track data across processes and hosts to shed light on the interaction of distributed components, but also for security purposes. This paper presents Taint-Exchange, a generic cross- process

Yang, Junfeng

29

Vertical Integration and Market Entry in the Generic Pharmaceutical Industry  

E-Print Network (OSTI)

in the Generic Pharmaceutical Industry . 2.2.1 Marketingin the Generic Pharmaceutical Industry 3.4 EconometricIntegration in the Generic Pharmaceutical Industry 2.1

Kubo, Kensuke

2011-01-01T23:59:59.000Z

30

Nuclear Reactor (atomic reactor)  

Science Journals Connector (OSTI)

A nuclear reactor splits Uranium or Plutonium nuclei, and the...235 is fissionable but more than 99% of the naturally occurring Uranium is U238 that makes enrichment mandatory. In some reactors U238 and Thorium23...

2008-01-01T23:59:59.000Z

31

Performance and Safety Analysis of a Generic Small Modular Reactor  

E-Print Network (OSTI)

for spent fuel from a Westinghouse AP1000. The results showed that from a fuel material standpoint, the SMR and AP1000 had effectively the same PR value. Unable to analyze security systems and methods employed at specific nuclear power plant sites...

Kitcher, Evans Damenortey, 1987-

2012-11-07T23:59:59.000Z

32

Descriptive Model of Generic WAMS  

SciTech Connect

The Department of Energy’s (DOE) Transmission Reliability Program is supporting the research, deployment, and demonstration of various wide area measurement system (WAMS) technologies to enhance the reliability of the Nation’s electrical power grid. Pacific Northwest National Laboratory (PNNL) was tasked by the DOE National SCADA Test Bed Program to conduct a study of WAMS security. This report represents achievement of the milestone to develop a generic WAMS model description that will provide a basis for the security analysis planned in the next phase of this study.

Hauer, John F.; DeSteese, John G.

2007-06-01T23:59:59.000Z

33

GLAD: A Generic LAttice Debugger  

SciTech Connect

Today, numerous simulation and analysis codes exist for the design, commission, and operation of accelerator beam lines. There is a need to develop a common user interface and database link to run these codes interactively. This paper will describe a proposed system, GLAD (Generic LAttice Debugger), to fulfill this need. Specifically, GLAD can be used to find errors in beam lines during commissioning, control beam parameters during operation, and design beam line optics and error correction systems for the next generation of linear accelerators and storage rings.

Lee, M.J.

1991-11-01T23:59:59.000Z

34

Nuclear power plant Generic Aging Lessons Learned (GALL). Appendix B  

SciTech Connect

The purpose of this generic aging lessons learned (GALL) review is to provide a systematic review of plant aging information in order to assess materials and component aging issues related to continued operation and license renewal of operating reactors. Literature on mechanical, structural, and thermal-hydraulic components and systems reviewed consisted of 97 Nuclear Plant Aging Research (NPAR) reports, 23 NRC Generic Letters, 154 Information Notices, 29 Licensee Event Reports (LERs), 4 Bulletins, and 9 Nuclear Management and Resources Council Industry Reports (NUMARC IRs) and literature on electrical components and systems reviewed consisted of 66 NPAR reports, 8 NRC Generic Letters, 111 Information Notices, 53 LERs, 1 Bulletin, and 1 NUMARC IR. More than 550 documents were reviewed. The results of these reviews were systematized using a standardized GALL tabular format and standardized definitions of aging-related degradation mechanisms and effects. The tables are included in volume s 1 and 2 of this report. A computerized data base has also been developed for all review tables and can be used to expedite the search for desired information on structures, components, and relevant aging effects. A survey of the GALL tables reveals that all ongoing significant component aging issues are currently being addressed by the regulatory process. However, the aging of what are termed passive components has been highlighted for continued scrutiny. This report consists of Volume 2, which consists of the GALL literature review tables for the NUMARC Industry Reports reviewed for the report.

Kasza, K.E.; Diercks, D.R.; Holland, J.W.; Choi, S.U. [and others

1996-12-01T23:59:59.000Z

35

FIBWR: a steady-state core flow distribution code for boiling water reactors code verification and qualification report. Final report  

SciTech Connect

A steady-state core flow distribution code (FIBWR) is described. The ability of the recommended models to predict various pressure drop components and void distribution is shown by comparison to the experimental data. Application of the FIBWR code to the Vermont Yankee Nuclear Power Station is shown by comparison to the plant measured data.

Ansari, A.F.; Gay, R.R.; Gitnick, B.J.

1981-07-01T23:59:59.000Z

36

Generic physical protection logic trees  

SciTech Connect

Generic physical protection logic trees, designed for application to nuclear facilities and materials, are presented together with a method of qualitative evaluation of the trees for design and analysis of physical protection systems. One or more defense zones are defined where adversaries interact with the physical protection system. Logic trees that are needed to describe the possible scenarios within a defense zone are selected. Elements of a postulated or existing physical protection system are tagged to the primary events of the logic tree. The likelihood of adversary success in overcoming these elements is evaluated on a binary, yes/no basis. The effect of these evaluations is propagated through the logic of each tree to determine whether the adversary is likely to accomplish the end event of the tree. The physical protection system must be highly likely to overcome the adversary before he accomplishes his objective. The evaluation must be conducted for all significant states of the site. Deficiencies uncovered become inputs to redesign and further analysis, closing the loop on the design/analysis cycle.

Paulus, W.K.

1981-10-01T23:59:59.000Z

37

Techniques in Active and Generic Software Libraries  

E-Print Network (OSTI)

the construction of algorithms for one domain entirely in terms of formalisms from a second domain; the construction of generic algorithms for algorithmic differentiation, implemented as an active library in Spad, language of the Open Axiom computer algebra system...

Smith, Jacob N.

2010-07-14T23:59:59.000Z

38

Generic specificity of Aeromonas extracellular antigens  

E-Print Network (OSTI)

in rabbits. Ouchterlony double diffusion analysis of each antigen in both the homo- logous and heterologous immune systems indicates suffi- cient antigenic relatedness between the species to allow generic identification with the possible exception fh mo... in rabbits. Ouchterlony double diffusion analysis of each antigen in both the homo- logous and heterologous immune systems indicates suffi- cient antigenic relatedness between the species to allow generic identification with the possible exception fh mo...

Brinkley, Allen Wayne

2012-06-07T23:59:59.000Z

39

Mass tracking and material accounting in the Integral Fast Reactor (IFR)  

SciTech Connect

The Integral Fast Reactor (IFR) is a generic advanced liquid metal cooled reactor concept being developed at Argonne National Laboratory (ANL). There are a number of technical features of the IFR which contribute to its potential as a next-generation reactor. These are associated with large safety margins with regard to off-normal events involving the heat transport system, and the use of metallic fuel which makes possible the utilization of innovative fuel cycle processes. The latter feature permits fuel cycle closure the compact, low-cost reprocessing facilities, collocated with the reactor plant. These primary features are being demonstrated in the facilities at ANL-West, utilizing Experimental Breeder Reactor 2 and the associated Fuel Cycle Facility (FCF) as an IFR prototype. The demonstration of this IFR prototype includes the design and implementation of the Mass-Tracking System (MTG). In this system, data from the operations of the FCF, including weights and batch-process parameters, are collected and maintained by the MTG running on distributed workstations. The components of the MTG System include: (1) an Oracle database manager with a Fortran interface, (2) a set of MTG Tasks'' which collect, manipulate and report data, (3) a set of MTG Terminal Sessions'' which provide some interactive control of the Tasks, and (4) a set of servers which manage the Tasks and which provide the communications link between the MTG System and Operator Control Stations, which control process equipment and monitoring devices within the FCF.

Orechwa, Y.; Adams, C.H.; White, A.M.

1991-01-01T23:59:59.000Z

40

Mass tracking and material accounting in the integral fast reactor (IFR)  

SciTech Connect

This paper reports on the Integral Fast Reactor (IFR) which is a generic advanced liquid metal cooled reactor concept being developed at Argonne National Laboratory. There are a number of technical features of the IFR which contribute to its potential as a next-generation reactor. These are associated with large safety margins with regard to off-normal events involving the heat transport system, and the use of metallic fuel which makes possible the utilization of innovative fuel cycle processes. The latter feature permits fuel cycle closure with compact, low-cost reprocessing facilities, collocated with the reactor plant. These primary features are being demonstrated in the facilities at ANL-West, utilizing Experimental Breeder Reactor II and the associated Fuel Cycle Facility (FCF) as an IFR prototype. The demonstration of this IFR prototype includes the design and implementation of the Mass-tracking System (MTG). In this system, data from the operations of the FCF, including weights and batch-process parameters, are collected and maintained by the MTG running on distributed workstations.

Orechwa, Y.; Adams, C.H.; White, A.M. (Argonne National Lab., IL (United States). Reactor Analysis and Safety Div.)

1991-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor generic distributed" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

Invisible Genericity and 0# M.C. Stanley  

E-Print Network (OSTI)

Invisible Genericity and 0# M.C. Stanley February 1995 Abstract. 0# can be invisibly class generic be invisibly generic . . . . . . . . . . . . . . 5 3. Two remarks of the predicates P and G in the conclusion of the theorem. It is in this sense that 0# can be "invisibly generic

Stanley, Maurice M.C. "Mack"

42

University Reactor Conversion Lessons Learned Workshop for Texas A&M University Nuclear Science Center Reactor  

SciTech Connect

The objectives of this meeting were to capture the observations, insights, issues, concerns, and ideas of those involved in the Texas A&M University Nuclear Science Center (TAMU NSC) TRIGA Reactor Conversion so that future efforts can be conducted with greater effectiveness, efficiency, and with fewer challenges. This workshop was held in conjunction with a similar workshop for the University of Florida Reactor Conversion. Some of the generic lessons from that workshop are included in this report for completeness.

Eric C. Woolstenhulme; Dana M. Meyer

2007-04-01T23:59:59.000Z

43

Distribution:  

Office of Legacy Management (LM)

JAN26 19% JAN26 19% Distribution: OR00 Attn: h.H.M.Roth DFMusser ITMM MMMann INS JCRyan FIw(2) Hsixele SRGustavson, Document rocm Formal file i+a@mmm bav@ ~@esiaw*cp Suppl. file 'Br & Div rf's s/health (lic.only) UNITED STATES ATOMIC ENERGY COMMISSION SPECIAL NUCLEAB MATERIAL LICENSE pursuant to the Atomic Energy Act of 1954 and Title 10, Code of Federal Regulations, Chapter 1, P&t 70, "Special Nuclear Material Reg)llatiqm," a license is hereby issued a$hortztng the licensee to rekeive and possess the special nuclear material designated below; to use such special nuclear mat&ial for the purpose(s) and at the place(s) designated below; and to transfer such material to per&s authorized to receive it in accordance with the regula,tions in said Part.

44

Generic clients supported by Web Services  

E-Print Network (OSTI)

Generic clients supported by Web Services Xuesong Liu Kongens Lyngby 10th April 2007 #12;Technical With an established Web Services Framework to expose the functionalities of Maconomy ERP system as Web Services (WS is to investigate two specific WS pro- tocols for SOA - "Microsoft Information Bridge Framework (IBF)" and "Web

45

Distribution of Correspondence  

Directives, Delegations, and Requirements

Defines correct procedures for distribution of correspondence to the Naval Reactors laboratories. Does not cancel another directive. Expired 8-30-97.

1996-08-30T23:59:59.000Z

46

Distribution ICategory: General Reactor Technology  

E-Print Network (OSTI)

, and use it. It is a natural background, a necessity of life, a pollutant when in excess, a cure excessive! amounts of ultraviolet light from the sun to penetrate the earth's atmosphere and reach its a minimum of new hazards. X-rays have been with us since the 1890's and radioactivity was discovered soon

Shlyakhter, Ilya

47

Structure Optimization of CFB Reactor for Moderate Temperature FGD  

Science Journals Connector (OSTI)

The gas velocity distribution, sorbent particle concentration distribution and particle residence time in circulating fluidized bed (CFB) reactors for moderate temperature flue gas desulfurization ... when the in...

Yuan Li; Jie Zhang; Kai Zheng; Changfu You

2013-01-01T23:59:59.000Z

48

NUCLEAR REACTORS.  

E-Print Network (OSTI)

??Nuclear reactors are devices containing fissionable material in sufficient quantity and so arranged as to be capable of maintaining a controlled, self-sustaining NUCLEAR FISSION chain… (more)

Belachew, Dessalegn

2010-01-01T23:59:59.000Z

49

PROGRESS ON GENERIC PHASE-FIELD METHOD DEVELOPMENT  

SciTech Connect

In this report, we summarize our current collobarative efforts, involving three national laboratories: Idaho National Laboratory (INL), Pacific Northwest National Laboratory (PNNL) and Los Alamos National Laboatory (LANL), to develop a computational framework for homogenous and heterogenous nucleation mechanisms into the generic phase-field model. During the studies, the Fe-Cr system was chosen as a model system due to its simplicity and availability of reliable thermodynamic and kinetic data, as well as the range of applications of low-chromium ferritic steels in nuclear reactors. For homogenous nucleation, the relavant parameters determined from atomistic studies were used directly to determine the energy functional and parameters in the phase-field model. Interfacial energy, critical nucleus size, nucleation rate, and coarsening kinetics were systematically examined in two- and three- dimensional models. For the heteregoneous nucleation mechanism, we studied the nucleation and growth behavior of chromium precipitates due to the presence of dislocations. The results demonstrate that both nucleation schemes can be introduced to a phase-field modeling algorithm with the desired accuracy and computational efficiency.

Biner, Bullent; Tonks, Michael; Millett, Paul C.; Li, Yulan; Hu, Shenyang Y.; Gao, Fei; Sun, Xin; Martinez, E.; Anderson, D.

2012-09-26T23:59:59.000Z

50

Transient thermal analysis of a space reactor power system  

E-Print Network (OSTI)

Thermoelectric Power Conversion Module Heat Pipe Radiator Module . Auxiliary Modules . Flow of Calculation . Transient Test Cases Studied Summary . 10 10 CHAPTER II. ENERGY EQUATION FINITE DIFFERENCING . . 12 Energy Equation for a Solid Finite..., but this stud~ uses a generic liquid metal cooled fast reactor concept as the model to test the code. The space power svstem to be modeled consists of a liquid lithium cooled fast reactor, primarv and secondary loops svith a sell-induced thermoelectric...

Gaeta, Michael J.

1988-01-01T23:59:59.000Z

51

In-service Inspection Ultrasonic Testing of Reactor Pressure Vessel Welds for Assessing Flaw Density and Size Distribution per 10 CFR 50.61a, Alternate Fracture Toughness Requirements  

SciTech Connect

Pressurized thermal shock (PTS) events are system transients in a pressurized water reactor (PWR) in which there is a rapid operating temperature cool-down that results in cold vessel temperatures with or without repressurization of the vessel. The rapid cooling of the inside surface of the reactor pressure vessel (RPV) causes thermal stresses that can combine with stresses caused by high pressure. The aggregate effect of these stresses is an increase in the potential for fracture if a pre-existing flaw is present in a material susceptible to brittle failure. The ferritic, low alloy steel of the reactor vessel beltline adjacent to the core, where neutron radiation gradually embrittles the material over the lifetime of the plant, can be susceptible to brittle fracture. The PTS rule, described in the Code of Federal Regulations, Title 10, Section 50.61 (§50.61), “Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events,” adopted on July 23, 1985, establishes screening criteria to ensure that the potential for a reactor vessel to fail due to a PTS event is deemed to be acceptably low. The U.S. Nuclear Regulatory Commission (NRC) completed a research program that concluded that the risk of through-wall cracking due to a PTS event is much lower than previously estimated. The NRC subsequently developed a rule, §50.61a, published on January 4, 2010, entitled “Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events” (75 FR 13). Use of the new rule by licensees is optional. The §50.61a rule differs from §50.61 in that it requires licensees who choose to follow this alternate method to analyze the results from periodic volumetric examinations required by the ASME Code, Section XI, Rules for Inservice Inspection (ISI) of Nuclear Power Plants. These analyses are intended to determine if the actual flaw density and size distribution in the licensee’s reactor vessel beltline welds are bounded by the flaw density and size distribution values used in the PTS technical basis. Under a contract with the NRC, Pacific Northwest National Laboratory (PNNL) has been working on a program to assess the ability of current inservice inspection (ISI)-ultrasonic testing (UT) techniques, as qualified through ASME Code, Appendix VIII, Supplements 4 and 6, to detect small fabrication or inservice-induced flaws located in RPV welds and adjacent base materials. As part of this effort, the investigators have pursued an evaluation, based on the available information, of the capability of UT to provide flaw density/distribution inputs for making RPV weld assessments in accordance with §50.61a. This paper presents the results of an evaluation of data from the 1993 Browns Ferry Nuclear Plant, Unit 3, Spirit of Appendix VIII reactor vessel examination, a comparison of the flaw density/distribution from this data with the distribution in §50.61a, possible reasons for differences, and plans and recommendations for further work in this area.

Sullivan, Edmund J.; Anderson, Michael T.; Norris, Wallace

2012-09-17T23:59:59.000Z

52

Generic turbine design study. Final report  

SciTech Connect

The purpose of Task 12, Generic Turbine Design Study was to develop a conceptual design of a combustion turbine system that would perform in a pressurized fluidized bed combustor (PFBC) application. A single inlet/outlet casing design that modifies the W251B12 combustion turbine to provide compressed air to the PFBC and accept clean hot air from the PFBC was developed. Performance calculations show that the net power output expected, at an inlet temperature of 59{degrees}F, is 20,250 kW.

Not Available

1993-06-01T23:59:59.000Z

53

naval reactors  

National Nuclear Security Administration (NNSA)

After operating for 34 years and training over 14,000 sailors, the Department of Energy S1C Prototype Reactor Site in Windsor, Connecticut, was returned to "green field"...

54

Alternate-fuel reactor studies  

SciTech Connect

A number of studies related to improvements and/or greater understanding of alternate-fueled reactors is presented. These studies cover the areas of non-Maxwellian distributions, materials and lifetime analysis, a /sup 3/He-breeding blanket, tritium-rich startup effects, high field magnet support, and reactor operation spanning the range from full D-T operation to operation with no tritium breeding.

Evans, K. Jr.; Ehst, D.A.; Gohar, Y.; Jung, J.; Mattas, R.F.; Turner, L.R.

1983-02-01T23:59:59.000Z

55

High-Cost Generic Drugs — Implications for Patients and Policymakers  

Science Journals Connector (OSTI)

It is well known that new brand-name drugs are often expensive, but U.S. health care is also witnessing a lesser-known but growing and seemingly paradoxical phenomenon: certain older drugs, many of which are generic and not protected by patents or market exclusivity, are now also extremely expensive... Some older generic drugs have become very expensive, owing to factors including drug shortages, supply disruptions, and consolidations in the generic-drug industry. But generics manufacturers that legally obtain a market monopoly can also unilaterally raise prices.

Alpern J.D.Stauffer W.M.Kesselheim A.S.

2014-11-13T23:59:59.000Z

56

Generic copy of DOE’s IDIQ ESPC contract  

Energy.gov (U.S. Department of Energy (DOE))

Generic copy of the U.S. Department of Energy’s Indefinite Delivery Indefinite Quantity (IDIQ) Energy Savings Performance Contracts (ESPCs) contract.

57

Summary Notes from 28 May 2008 Generic Technical Issue Discussion...  

Office of Environmental Management (EM)

1 of 8 Summary Notes from 28 May 2008 Generic Technical Issue Discussion on Estimating Waste Inventory and Waste Tank Characterization Attendees: Representatives from Department...

58

Geothermal: Sponsored by OSTI -- Generic Natural Systems Evaluation...  

Office of Scientific and Technical Information (OSTI)

Generic Natural Systems Evaluation - Thermodynamic Database Development and Data Management Geothermal Technologies Legacy Collection HelpFAQ | Site Map | Contact Us | Admin Log...

59

Research reactors - an overview  

SciTech Connect

A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs.

West, C.D.

1997-03-01T23:59:59.000Z

60

Light Water Reactor Sustainability  

NLE Websites -- All DOE Office Websites (Extended Search)

4 Light Water Reactor Sustainability ACCOMPLISHMENTS REPORT 2014 Accomplishments Report | Light Water Reactor Sustainability 2 T he mission of the Light Water Reactor...

Note: This page contains sample records for the topic "reactor generic distributed" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

Catalytic reactor  

DOE Patents (OSTI)

A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

Aaron, Timothy Mark (East Amherst, NY); Shah, Minish Mahendra (East Amherst, NY); Jibb, Richard John (Amherst, NY)

2009-03-10T23:59:59.000Z

62

Transport reactor development status  

SciTech Connect

This project is part of METC`s Power Systems Development Facility (PSDF) located at Wilsonville, Alabama. The primary objective of the Advanced Gasifier module is to produce vitiated gases for intermediate-term testing of Particulate Control Devices (PCDs). The Transport reactor potentially allows particle size distribution, solids loading, and particulate characteristics in the off-gas stream to be varied in a number of ways. Particulates in the hot gases from the Transport reactor will be removed in the PCDs. Two PCDs will be initially installed in the module; one a ceramic candle filter, the other a granular bed filter. After testing of the initial PCDs they will be removed and replaced with PCDs supplied by other vendors. A secondary objective is to verify the performance of a Transport reactor for use in advanced Integrated Gasification Combined Cycle (IGCC), Integrated Gasification Fuel Cell (IG-FC), and Pressurized Combustion Combined Cycle (PCCC) power generation units. This paper discusses the development of the Transport reactor design from bench-scale testing through pilot-scale testing to design of the Process Development Unit (PDU-scale) facility at Wilsonville.

Rush, R.E.; Fankhanel, M.O.; Campbell, W.M.

1994-10-01T23:59:59.000Z

63

Generic measures for geodesic flows on nonpositively curved manifolds  

E-Print Network (OSTI)

Generic measures for geodesic flows on nonpositively curved manifolds Yves Coud`ene, Barbara the generic invariant probability measures for the geodesic flow on connected complete nonpositively curved subset of the set of all probability measures invariant by the geodesic flow. The proof of K. Sigmund

Paris-Sud XI, Université de

64

A technique for generic iteration and its optimization  

Science Journals Connector (OSTI)

Software libraries rely increasingly on iterators to provide generic traversal of data structures. These iterators can be represented either as objects that maintain state or as programs that suspend and resume control. This paper addresses two problems ... Keywords: generic program, iterators

Stephen M. Watt

2006-09-01T23:59:59.000Z

65

Generic planning and control of automated material handling systems  

Science Journals Connector (OSTI)

This paper discusses the problem to design a generic planning and control architecture for automated material handling systems (AMHSs). We illustrate the relevance of this research direction, and then address three different market sectors where AMHSs ... Keywords: Automated material handling systems, Generic control architecture, Real-time scheduling

S. W. A. Haneyah; J. M. J. Schutten; P. C. Schuur; W. H. M. Zijm

2013-04-01T23:59:59.000Z

66

Evaluation on the Feasibility of Using Ultrasonic Testing of Reactor Pressure Vessel Welds for Assessing Flaw Density/Distribution per 10 CFR 50.61a, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock  

SciTech Connect

This technical letter report provides the status of an assessment undertaken by PNNL at the request of the NRC to verify the capability of periodic ASME-required volumetric examinations of reactor vessels to characterize the density and distribution of flaws of interest for applying §50.61a on a plant-by-plant basis. The PTS rule, described in the Code of Federal Regulations, Title 10, Section 50.61 (§50.61), "Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events," establishes screening criteria to ensure that the potential for a reactor vessel to fail due to a PTS event is deemed to be acceptably low. Recently, the NRC completed a research program that concluded that the risk of through-wall cracking due to a PTS event is much lower than previously estimated. The NRC subsequently developed and promulgated an alternate PTS rule, §50.61a, that can be implemented by PWR licensees. The §50.61a rule differs from §50.61 in that it requires licensees who choose to follow this alternate method to analyze the results from periodic volumetric examinations required by the ASME Code, Section XI, Rules for Inservice Inspection (ISI) of Nuclear Power Plants.

Sullivan, Edmund J.; Anderson, Michael T.

2014-06-10T23:59:59.000Z

67

Generic Disposal System Modeling, Fiscal Year 2011 Progress Report |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Disposal System Modeling, Fiscal Year 2011 Progress Report Disposal System Modeling, Fiscal Year 2011 Progress Report Generic Disposal System Modeling, Fiscal Year 2011 Progress Report The UFD Campaign is developing generic disposal system models (GDSM) of different disposal environments and waste form options. Currently, the GDSM team is investigating four main disposal environment options: mined repositories in three geologic media (salt, clay, and granite) and the deep borehole concept in crystalline rock (DOE 2010d). Further developed the individual generic disposal system (GDS) models for salt, granite, clay, and deep borehole disposal environments. GenericDisposalSystModelFY11.pdf More Documents & Publications Integration of EBS Models with Generic Disposal System Models TSPA Model Development and Sensitivity Analysis of Processes Affecting

68

Descriptive Model of a Generic WAMS | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Descriptive Model of a Generic WAMS Descriptive Model of a Generic WAMS Descriptive Model of a Generic WAMS The Department of Energy's (DOE) Transmission Reliability Program is supporting the research, deployment, and demonstration of various wide area measurement system (WAMS) technologies to enhance the reliability of the Nation's electrical power grid. Pacific Northwest National Laboratory (PNNL) was tasked by the DOE National SCADA Test Bed Program to conduct a study of WAMS security. This report represents achievement of the milestone to develop a generic WAMS model description that will provide a basis for the security analysis planned in the next phase of this study. Descriptive Model of a Generic WAMS More Documents & Publications Securing Wide Area Measurement Systems 2012 Advanced Applications Research & Development Peer Review - Day 1

69

Generic Argillite/Shale Disposal Reference Case  

SciTech Connect

Radioactive waste disposal in a deep subsurface repository hosted in clay/shale/argillite is a subject of widespread interest given the desirable isolation properties, geochemically reduced conditions, and widespread geologic occurrence of this rock type (Hansen 2010; Bianchi et al. 2013). Bianchi et al. (2013) provides a description of diffusion in a clay-hosted repository based on single-phase flow and full saturation using parametric data from documented studies in Europe (e.g., ANDRA 2005). The predominance of diffusive transport and sorption phenomena in this clay media are key attributes to impede radionuclide mobility making clay rock formations target sites for disposal of high-level radioactive waste. The reports by Hansen et al. (2010) and those from numerous studies in clay-hosted underground research laboratories (URLs) in Belgium, France and Switzerland outline the extensive scientific knowledge obtained to assess long-term clay/shale/argillite repository isolation performance of nuclear waste. In the past several years under the UFDC, various kinds of models have been developed for argillite repository to demonstrate the model capability, understand the spatial and temporal alteration of the repository, and evaluate different scenarios. These models include the coupled Thermal-Hydrological-Mechanical (THM) and Thermal-Hydrological-Mechanical-Chemical (THMC) models (e.g. Liu et al. 2013; Rutqvist et al. 2014a, Zheng et al. 2014a) that focus on THMC processes in the Engineered Barrier System (EBS) bentonite and argillite host hock, the large scale hydrogeologic model (Bianchi et al. 2014) that investigates the hydraulic connection between an emplacement drift and surrounding hydrogeological units, and Disposal Systems Evaluation Framework (DSEF) models (Greenberg et al. 2013) that evaluate thermal evolution in the host rock approximated as a thermal conduction process to facilitate the analysis of design options. However, the assumptions and the properties (parameters) used in these models are different, which not only make inter-model comparisons difficult, but also compromise the applicability of the lessons learned from one model to another model. The establishment of a reference case would therefore be helpful to set up a baseline for model development. A generic salt repository reference case was developed in Freeze et al. (2013) and the generic argillite repository reference case is presented in this report. The definition of a reference case requires the characterization of the waste inventory, waste form, waste package, repository layout, EBS backfill, host rock, and biosphere. This report mainly documents the processes in EBS bentonite and host rock that are potentially important for performance assessment and properties that are needed to describe these processes, with brief description other components such as waste inventory, waste form, waste package, repository layout, aquifer, and biosphere. A thorough description of the generic argillite repository reference case will be given in Jové Colon et al. (2014).

Zheng, Liange; Jov& #233; Colon, Carlos; Bianchi, Marco; Birkholzer, Jens

2014-08-08T23:59:59.000Z

70

Evaluation of in-vessel corium retention through external reactor vessel cooling for integral reactor  

SciTech Connect

In-vessel corium retention through external reactor vessel cooling (IVR-ERVC) for a small integral reactor has been evaluated to determine the thermal margin for the prevention of a reactor vessel failure. A thermal load analysis from the corium pool to the outer reactor vessel wall in the lower plenum of the reactor vessel has been performed to determine the heat flux distribution. The critical heat flux (CHF) on the outer reactor vessel wall has been determined to fix the maximum heat removal rate through the external coolant between the outer reactor vessel and the insulation of the reactor vessel. Finally, the thermal margin has been evaluated by comparison of the thermal load with the maximum heat removal rate of the CHF on the outer reactor vessel wall. The maximum heat flux from the corium pool to the outer reactor vessel is estimated at approximately 0.25 MW/m{sup 2} in the metallic layer because of the focusing effect. The CHF of the outer reactor vessel is approximately 1.1 MW/m{sup 2} because of a two phase natural circulation mass flow. Since the thermal margin for the IVR-ERVC is sufficient, the reactor vessel integrity is maintained during a severe accident of a small integral reactor. (authors)

Park, R. J.; Lee, J. R.; Kim, S. B.; Jin, Y.; Kim, H. Y. [Korea Atomic Energy Research Inst., 1045 Daedeok-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

2012-07-01T23:59:59.000Z

71

Handling Genericity in Chemical Structures Using the Markush Darc Software  

Science Journals Connector (OSTI)

Since an exact search against the entire database would be computationally inconsistent with an online service, screening steps are necessary for reducing the number of candidates to be searched during the atom-by-atom step. ... Markush Darc expresses generic terms as “Superatoms”, entered by two or three characters code, either in databases or in queries structures. ... fragments from generic structures for full structure and substructure searching is described; these include fragments from components described either in specific or in generic terms, and those which overlap them. ...

Pierre Benichou; Christine Klimczak; Philippe Borne

1997-01-27T23:59:59.000Z

72

A Parametric Study of the DUPIC Fuel Cycle to Reflect Pressurized Water Reactor Fuel Management Strategy  

SciTech Connect

For both pressurized water reactor (PWR) and Canada deuterium uranium (CANDU) tandem analysis, the Direct Use of spent PWR fuel In CANDU reactor (DUPIC) fuel cycle in a CANDU 6 reactor is studied using the DRAGON/DONJON chain of codes with the ENDF/B-V and ENDF/B-VI libraries. The reference feed material is a 17 x 17 French standard 900-MW(electric) PWR fuel. The PWR spent-fuel composition is obtained from two-dimensional DRAGON assembly transport and depletion calculations. After a number of years of cooling, this defines the initial fuel nuclide field in the CANDU unit cell calculations in DRAGON, where it is further depleted with the same neutron group structure. The resulting macroscopic cross sections are condensed and tabulated to be used in a full-core model of a CANDU 6 reactor to find an optimized channel fueling rate distribution on a time-average basis. Assuming equilibrium refueling conditions and a particular refueling sequence, instantaneous full-core diffusion calculations are finally performed with the DONJON code, from which both the channel power peaking factors and local parameter effects are estimated. A generic study of the DUPIC fuel cycle is carried out using the linear reactivity model for initial enrichments ranging from 3.2 to 4.5 wt% in a PWR. Because of the uneven power histories of the spent PWR assemblies, the spent PWR fuel composition is expected to differ from one assembly to the next. Uneven mixing of the powder during DUPIC fuel fabrication may lead to uncertainties in the composition of the fuel bundle and larger peaking factors in CANDU. A mixing method for reducing composition uncertainties is discussed.

Rozon, Daniel; Shen Wei [Institut de Genie Nucleaire (Canada)

2001-05-15T23:59:59.000Z

73

Light Water Reactors Technology Development - Nuclear Reactors  

NLE Websites -- All DOE Office Websites (Extended Search)

Light Water Reactors Light Water Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

74

Scaling behavior of optimally structured catalytic microfluidic reactors  

E-Print Network (OSTI)

In this study of catalytic microfluidic reactors we show that, when optimally structured, these reactors share underlying scaling properties. The scaling is predicted theoretically and verified numerically. Furthermore, we show how to increase the reaction rate significantly by distributing the active porous material within the reactor using a high-level implementation of topology optimization.

Okkels, F; Bruus, Henrik; Okkels, Fridolin

2006-01-01T23:59:59.000Z

75

Generic CTS Data Upload Templates | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Generic CTS Data Upload Templates Generic CTS Data Upload Templates Generic CTS Data Upload Templates October 8, 2013 - 2:03pm Addthis These generic Excel templates are being made available to the public so that Federal contractors and service providers can provide their clients with reports and findings consistent with the formats required for agencies to easily upload the data into CTS. Agencies may refer their energy/water audit contractors and project developers/evaluators to these templates as a reference for required data elements and reporting. Data may be batch imported by the Federal agencies into CTS using the following Excel spreadsheet templates. The "Agency Acronym" and facility identifying data contained in these templates must correspond to the existing IDs used in CTS. Contractors populating these templates for agency

76

3116 WASTE DETERMINATIONS PUBLIC MEETINGS AND GENERIC TECHNICAL ISSUES  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

3116 WASTE DETERMINATIONS PUBLIC MEETINGS AND GENERIC TECHNICAL 3116 WASTE DETERMINATIONS PUBLIC MEETINGS AND GENERIC TECHNICAL ISSUES SUMMARIES 3116 WASTE DETERMINATIONS PUBLIC MEETINGS AND GENERIC TECHNICAL ISSUES SUMMARIES Below are public meeting summaries and general technical issue summaries relating to 3116 waste determinations. The 3116 Public Meeting Summaries cover public meetings that the Department of Energy (DOE) and Nuclear Regulatory Commission (NRC) periodically host to provide the status of activities associated with waste determinations under Section 3116 (a) of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005. 3116 Public Meeting Summaries - July 2007 3116 Public Meeting Summaries - November 2006 The Generic Technical Issues Summaries cover the informal technical discussions between representatives of the Department of Energy (DOE),

77

Evaluation of Generic EBS Design Concepts and Process Models Implications  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Generic EBS Design Concepts and Process Models Generic EBS Design Concepts and Process Models Implications to EBS Design Optimization Evaluation of Generic EBS Design Concepts and Process Models Implications to EBS Design Optimization The assessment of generic Engineered Barrier System (EBS) concepts and design optimization to harbor various disposal configurations and waste types needs advanced approaches and methods to analyze barrier performance. The report addresses: 1) Overview of the importance of Thermal-Hydrological-Mechanical-Chemical (THMC) processes to barrier performance, and international collaborations; 2) THMC processes in clay barriers; 3) experimental studies of clay stability and clay-metal interactions at high temperatures and pressures; 4) thermodynamic modeling and database development; 5) Molecular Dynamics (MD) study of clay

78

Integration of EBS Models with Generic Disposal System Models | Department  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Integration of EBS Models with Generic Disposal System Models Integration of EBS Models with Generic Disposal System Models Integration of EBS Models with Generic Disposal System Models This report summarizes research activities on engineered barrier system (EBS) model integration with the generic disposal system model (GDSM), and used fuel degradation and radionuclide mobilization (RM) in support of the EBS evaluation and tool development within the Used Fuel Disposition campaign. This report addresses: predictive model capability for used nuclear fuel degradation based on electrochemical and thermodynamic principles, radiolysis model to evaluate the U(VI)-H2O-CO2 system, steps towards the evaluation of uranium alteration products, discussion of instant release fraction (IRF) of radionuclides from the nuclear fuel, and

79

Integration of EBS Models with Generic Disposal System Models | Department  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Integration of EBS Models with Generic Disposal System Models Integration of EBS Models with Generic Disposal System Models Integration of EBS Models with Generic Disposal System Models This report summarizes research activities on engineered barrier system (EBS) model integration with the generic disposal system model (GDSM), and used fuel degradation and radionuclide mobilization (RM) in support of the EBS evaluation and tool development within the Used Fuel Disposition campaign. This report addresses: predictive model capability for used nuclear fuel degradation based on electrochemical and thermodynamic principles, radiolysis model to evaluate the U(VI)-H2O-CO2 system, steps towards the evaluation of uranium alteration products, discussion of instant release fraction (IRF) of radionuclides from the nuclear fuel, and

80

THE CHARACTERISTIC VARIETY OF A GENERIC FOLIATION JORGE VITORIO PEREIRA  

E-Print Network (OSTI)

THE CHARACTERISTIC VARIETY OF A GENERIC FOLIATION JORGE VIT´ORIO PEREIRA Abstract. We confirm. characteristic foliation, invariant variety, D-modules. 1 #12;2 JORGE VIT ´ORIO PEREIRA F is non

Pereira, Jorge Vitório

Note: This page contains sample records for the topic "reactor generic distributed" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


81

A Generic Approach to Coat Carbon Nanotubes With Nanoparticles  

E-Print Network (OSTI)

A Generic Approach to Coat Carbon Nanotubes With Nanoparticles for Potential Energy Applications vari- ous nanoparticles onto multiwalled carbon nanotubes (CNTs). Charged and nonagglomerated aerosol unique hybrid nanostructures at- tractive for various energy applications. DOI: 10

Chen, Junhong

82

Generic implementations of parallel prefix sums and its applications  

E-Print Network (OSTI)

Parallel prefix sums algorithms are one of the simplest and most useful building blocks for constructing parallel algorithms. A generic implementation is valuable because of the wide range of applications for this method. This thesis presents a...

Huang, Tao

2009-05-15T23:59:59.000Z

83

Generic Sorting in RESOLVE Yu-Shan Sun  

E-Print Network (OSTI)

1 Generic Sorting in RESOLVE Yu-Shan Sun Dept. of Mathematics and Computer Science Denison University Granville OH, 43023, USA Email: sun s@denison.edu Joan Krone Dept. of Mathematics and Computer

84

Generic Models and Their Support in Modeling Problem Solving Behavior  

Science Journals Connector (OSTI)

Generic models have received widespread attention in knowledge based systems research (KBS) as an important aid in the process of modeling “problem solving behavior”. However, little empirical evidence has bee...

Philip Rademakers; Johan Vanwelkenhuysen

1993-01-01T23:59:59.000Z

85

Photocatalytic reactor  

DOE Patents (OSTI)

A photocatalytic reactor is described for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane. 4 figs.

Bischoff, B.L.; Fain, D.E.; Stockdale, J.A.D.

1999-01-19T23:59:59.000Z

86

Development of Hydrogen Selective Membranes/Modules as Reactors/Separators for Distributed Hydrogen Production - DOE Hydrogen and Fuel Cells Program FY 2012 Annual Progress Report  

NLE Websites -- All DOE Office Websites (Extended Search)

3 3 FY 2012 Annual Progress Report DOE Hydrogen and Fuel Cells Program Paul KT Liu Media and Process Technology Inc. (M&P) 1155 William Pitt Way Pittsburgh, PA 15238 Phone: (412) 826-3711 Email: pliu@mediaandprocess.com DOE Managers HQ: Sara Dillich Phone: (202) 586-7925 Email: Sara.Dillich@ee.doe.gov GO: Katie Randolph Phone: (720) 356-1759 Email: Katie.Randolph@go.doe.gov Contract Number: DE-FG36-05GO15092 Subcontractor: University of Southern California Project Start Date: July 1, 2005 Projected End Date: December 31, 2012 Fiscal Year (FY) 2012 Objectives The water-gas shift (WGS) reaction becomes less efficient when high CO conversion is required, such as for distributed hydrogen production applications. Our project

87

Hybrid adsorptive membrane reactor  

DOE Patents (OSTI)

A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

Tsotsis, Theodore T. (Huntington Beach, CA); Sahimi, Muhammad (Altadena, CA); Fayyaz-Najafi, Babak (Richmond, CA); Harale, Aadesh (Los Angeles, CA); Park, Byoung-Gi (Yeosu, KR); Liu, Paul K. T. (Lafayette Hill, PA)

2011-03-01T23:59:59.000Z

88

GEN-IV Reactors  

Science Journals Connector (OSTI)

Generation-IV reactors are a set of nuclear reactors currently being developed under international collaborations targeting ... economics, proliferation resistance, and physical protection of nuclear energy. Nuclear

Taek K. Kim

2013-01-01T23:59:59.000Z

89

The Netherlands Reactor Centre  

Science Journals Connector (OSTI)

... Two illustrated brochures in English have recently J. been issued by the Netherlands Reactor Centre ( ... Centre (Reactor Centrum Nederland). The first* gives a general survey of the ...

S. WEINTROUB

1964-04-04T23:59:59.000Z

90

Macroalgal distribution at Lee Stocking Island, Bahamas  

E-Print Network (OSTI)

cover) H' ln S H' = generic diversity S = ? of genera ln S = maximum possible generic diversity Kruskal-Wallis tests, which are distribution free analysis of variance techniques, were conducted to determine if significant differences occurred... comparisons between study sites (sampling periods combined). Based on the Kruskal-Wallis Test, ns indicates there was no significant difference. Sites Percent Cover Standard Deviation Statistical Significance Differences Iguana Norman's Pond North...

Roberts, Jill Christie

2012-06-07T23:59:59.000Z

91

T-SA-00582-2004.R1 A Generic Audio Classification and Segmentation Approach for Multimedia Indexing and Retrieval 1 A Generic Audio Classification and  

E-Print Network (OSTI)

T-SA-00582-2004.R1 A Generic Audio Classification and Segmentation Approach for Multimedia Indexing and Retrieval 1 A Generic Audio Classification and Segmentation Approach for Multimedia Indexing and Retrieval of generic and automatic audio classification and segmentation for audio-based multimedia indexing

Gabbouj, Moncef

92

Categorizing Threat Building and Using a Generic Threat Matrix | Department  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Categorizing Threat Building and Using a Generic Threat Matrix Categorizing Threat Building and Using a Generic Threat Matrix Categorizing Threat Building and Using a Generic Threat Matrix The key piece of knowledge necessary for building defenses capable of withstanding or surviving cyber and kinetic attacks is an understanding of the capabilities posed by threats to a government, function, or system. With the number of threats continuing to increase, it is no longer feasible to enumerate the capabilities of all known threats and then build defenses based on those threats that are considered, at the time, to be the most relevant. Exacerbating the problem for critical infrastructure entities is the fact that the majority of detailed threat information for higher-level threats is held in classified status and is not available for general

93

Radiation effects on reactor pressure vessel supports  

SciTech Connect

The purpose of this report is to present the findings from the work done in accordance with the Task Action Plan developed to resolve the Nuclear Regulatory Commission (NRC) Generic Safety Issue No. 15, (GSI-15). GSI-15 was established to evaluate the potential for low-temperature, low-flux-level neutron irradiation to embrittle reactor pressure vessel (RPV) supports to the point of compromising plant safety. An evaluation of surveillance samples from the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) had suggested that some materials used for RPV supports in pressurized-water reactors could exhibit higher than expected embrittlement rates. However, further tests designed to evaluate the applicability of the HFIR data to reactor RPV supports under operating conditions led to the conclusion that RPV supports could be evaluated using traditional method. It was found that the unique HFIR radiation environment allowed the gamma radiation to contribute significantly to the embrittlement. The shielding provided by the thick steel RPV shell ensures that degradation of RPV supports from gamma irradiation is improbable or minimal. The findings reported herein were used, in part, as the basis for technical resolution of the issue.

Johnson, R.E. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering Technology; Lipinski, R.E. [Idaho National Engineering Lab., Rockville, MD (United States)

1996-05-01T23:59:59.000Z

94

SRS Small Modular Reactors  

SciTech Connect

The small modular reactor program at the Savannah River Site and the Savannah River National Laboratory.

None

2012-04-27T23:59:59.000Z

95

SRS Small Modular Reactors  

ScienceCinema (OSTI)

The small modular reactor program at the Savannah River Site and the Savannah River National Laboratory.

None

2014-05-21T23:59:59.000Z

96

Generic repository design concepts and thermal analysis (FY11).  

SciTech Connect

Reference concepts for geologic disposal of used nuclear fuel and high-level radioactive waste in the U.S. are developed, including geologic settings and engineered barriers. Repository thermal analysis is demonstrated for a range of waste types from projected future, advanced nuclear fuel cycles. The results show significant differences among geologic media considered (clay/shale, crystalline rock, salt), and also that waste package size and waste loading must be limited to meet targeted maximum temperature values. In this study, the UFD R&D Campaign has developed a set of reference geologic disposal concepts for a range of waste types that could potentially be generated in advanced nuclear FCs. A disposal concept consists of three components: waste inventory, geologic setting, and concept of operations. Mature repository concepts have been developed in other countries for disposal of spent LWR fuel and HLW from reprocessing UNF, and these serve as starting points for developing this set. Additional design details and EBS concepts will be considered as the reference disposal concepts evolve. The waste inventory considered in this study includes: (1) direct disposal of SNF from the LWR fleet, including Gen III+ advanced LWRs being developed through the Nuclear Power 2010 Program, operating in a once-through cycle; (2) waste generated from reprocessing of LWR UOX UNF to recover U and Pu, and subsequent direct disposal of used Pu-MOX fuel (also used in LWRs) in a modified-open cycle; and (3) waste generated by continuous recycling of metal fuel from fast reactors operating in a TRU burner configuration, with additional TRU material input supplied from reprocessing of LWR UOX fuel. The geologic setting provides the natural barriers, and establishes the boundary conditions for performance of engineered barriers. The composition and physical properties of the host medium dictate design and construction approaches, and determine hydrologic and thermal responses of the disposal system. Clay/shale, salt, and crystalline rock media are selected as the basis for reference mined geologic disposal concepts in this study, consistent with advanced international repository programs, and previous investigations in the U.S. The U.S. pursued deep geologic disposal programs in crystalline rock, shale, salt, and volcanic rock in the years leading up to the Nuclear Waste Policy Act, or NWPA (Rechard et al. 2011). The 1987 NWPA amendment act focused the U.S. program on unsaturated, volcanic rock at the Yucca Mountain site, culminating in the 2008 license application. Additional work on unsaturated, crystalline rock settings (e.g., volcanic tuff) is not required to support this generic study. Reference disposal concepts are selected for the media listed above and for deep borehole disposal, drawing from recent work in the U.S. and internationally. The main features of the repository concepts are discussed in Section 4.5 and summarized in Table ES-1. Temperature histories at the waste package surface and a specified distance into the host rock are calculated for combinations of waste types and reference disposal concepts, specifying waste package emplacement modes. Target maximum waste package surface temperatures are identified, enabling a sensitivity study to inform the tradeoff between the quantity of waste per disposal package, and decay storage duration, with respect to peak temperature at the waste package surface. For surface storage duration on the order of 100 years or less, waste package sizes for direct disposal of SNF are effectively limited to 4-PWR configurations (or equivalent size and output). Thermal results are summarized, along with recommendations for follow-on work including adding additional reference concepts, verification and uncertainty analysis for thermal calculations, developing descriptions of surface facilities and other system details, and cost estimation to support system-level evaluations.

Howard, Robert (Oak Ridge National Laboratory, Oak Ridge, TN); Dupont, Mark (Savannah River Nuclear Solutions, Aiken, SC); Blink, James A. (Lawrence Livermore National Laboratory, Livermore, CA); Fratoni, Massimiliano (Lawrence Livermore National Laboratory, Livermore, CA); Greenberg, Harris (Lawrence Livermore National Laboratory, Livermore, CA); Carter, Joe (Savannah River Nuclear Solutions, Aiken, SC); Hardin, Ernest L.; Sutton, Mark A. (Lawrence Livermore National Laboratory, Livermore, CA)

2011-08-01T23:59:59.000Z

97

Nuclear reactor  

DOE Patents (OSTI)

A nuclear reactor comprising a cylindrical pressure vessel, an elongated annular core centrally disposed within and spaced from the pressure vessel, and a plurality of ducts disposed longitudinally of the pressure vessel about the periphery thereof, said core comprising an annular active portion, an annular reflector just inside the active portion, and an annular reflector just outside the active a portion, said annular active portion comprising rectangular slab, porous fuel elements radially disposed around the inner reflector and extending the length of the active portion, wedge-shaped, porous moderator elements disposed adjacent one face of each fuel element and extending the length of the fuel element, the fuel and moderator elements being oriented so that the fuel elements face each other and the moderator elements do likewise, adjacent moderator elements being spaced to provide air inlet channels, and adjacent fuel elements being spaced to provide air outlet channels which communicate with the interior of the peripheral ducts, and means for introducing air into the air inlet channels which passes through the porous moderator elements and porous fuel elements to the outlet channel.

Thomson, Wallace B. (Severna Park, MD)

2004-03-16T23:59:59.000Z

98

Parallel Monte Carlo reactor neutronics  

SciTech Connect

The issues affecting implementation of parallel algorithms for large-scale engineering Monte Carlo neutron transport simulations are discussed. For nuclear reactor calculations, these include load balancing, recoding effort, reproducibility, domain decomposition techniques, I/O minimization, and strategies for different parallel architectures. Two codes were parallelized and tested for performance. The architectures employed include SIMD, MIMD-distributed memory, and workstation network with uneven interactive load. Speedups linear with the number of nodes were achieved.

Blomquist, R.N.; Brown, F.B.

1994-03-01T23:59:59.000Z

99

Proceedings CHI'95, Denver, May 1995 A Generic Platform  

E-Print Network (OSTI)

, multimodal interaction requires [3]: · the fusion of different types of data originating from distinct is concerned with the fusion of information produced through distinct interaction techniques. In this article, we present a generic fusion engine that can be embedded in a multi-agent architecture modelling

Nigay, Laurence

100

Proceedings CHI'95, Denver, May 1995 A Generic Platform  

E-Print Network (OSTI)

. In particular, multimodal interaction requires [3]: . the fusion of different types of data originating from is concerned with the fusion of information produced through distinct interaction techniques. In this article, we present a generic fusion engine that can be embedded in a multi­agent architecture modelling

Nigay, Laurence

Note: This page contains sample records for the topic "reactor generic distributed" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


101

Towards a Generic Architecture for Autonomous Landing Systems1  

E-Print Network (OSTI)

architecture, autonomous unmanned vehicle, small body (comet/asteroid) landing, real-time control 1 interest in small planetary body (comet/asteroid) landing, not least because of the special challengesTowards a Generic Architecture for Autonomous Landing Systems1 James Goodwin, Alan Winfield, Quan

Winfield, Alan FT

102

Generic Multiversion STM Li Lu and Michael L. Scott  

E-Print Network (OSTI)

Generic Multiversion STM Li Lu and Michael L. Scott Computer Science Department, University of Rochester Rochester, NY 14627-0226 USA {llu,scott}@cs.rochester.edu Abstract. Multiversion software by the National Science Foundation under grants CCR- 0963759, CCF-1116055, and CNS-1116109. Y. Afek (Ed.): DISC

Scott, Michael L.

103

Nonlinear Adaptive Dynamic Inversion Applied to a Generic Hypersonic Vehicle  

E-Print Network (OSTI)

Nonlinear Adaptive Dynamic Inversion Applied to a Generic Hypersonic Vehicle Elizabeth Rollins Conclusions Extensions 3 / 50 #12;Motivation Control of Hypersonic Vehicles · Wide range of flight conditions in hypersonic flight · Three main causes of inlet unstarts: 1 A flow to the inlet that is slower than

Valasek, John

104

Nonlinear Adaptive Dynamic Inversion Applied to a Generic Hypersonic Vehicle  

E-Print Network (OSTI)

Nonlinear Adaptive Dynamic Inversion Applied to a Generic Hypersonic Vehicle Elizabeth Rollins of hypersonic vehicles is challenging because of the wide range of oper- ating conditions encountered and certain aspects unique to high speed flight. A particular safety concern in hypersonic flight is the risk

Valasek, John

105

Type-Based Analysis of Generic Key Management APIs  

E-Print Network (OSTI)

Type-Based Analysis of Generic Key Management APIs Pedro Ad~ao1,2 , Riccardo Focardi3, Universit`a Ca' Foscari, Venezia, Italy Abstract In the past few years, cryptographic key management APIs. In fact, real APIs provide mechanisms to declare the intended use of keys but they are not strong enough

106

ERP data sharing framework using the Generic Product Model (GPM)  

Science Journals Connector (OSTI)

Nowadays, all product life cycle processes are investigated deeply in order to get an advantage over competitors. To support these processes, several software applications are available. However, this wide range of heterogeneous applications leads to ... Keywords: Enterprise Resource Planning, Generic Product Model, Information sharing, Management data

Souleiman Naciri; Naoufel Cheikhrouhou; Michel Pouly; Jean-Charles Binggeli; Rémy Glardon

2011-02-01T23:59:59.000Z

107

A Generic FMU Interface for Modelica Wuzhu Chen1  

E-Print Network (OSTI)

A Generic FMU Interface for Modelica Wuzhu Chen1 Michaela Huhn1 Peter Fritzson2 1 Department-up Unit (FMU) into Modelica simulators, specifically the Open- Modelica environment. Whereas other.0 Specification for model exchange from MOD- ELISAR can be imported into any Modelica simulator. When importing

Zhao, Yuxiao

108

Reasoning with Generic Cases in the Arithmetic of Abstract Matrices  

E-Print Network (OSTI)

Department of Computer Science, University of Western Ontario www.csd.uwo.ca/watt Abstract. In previous work specifying n. We call this working with "symbolic" or "abstract" values. In previous work we have givenReasoning with Generic Cases in the Arithmetic of Abstract Matrices Alan P. Sexton1 , Volker Sorge1

Sorge, Volker

109

Reasoning with Generic Cases in the Arithmetic of Abstract Matrices  

E-Print Network (OSTI)

Department of Computer Science, University of Western Ontario www.csd.uwo.ca/~watt Abstract. In previous work specifying n. We call this working with "symbolic" or "abstract" values. In previous work we have givenReasoning with Generic Cases in the Arithmetic of Abstract Matrices Alan P. Sexton1 , Volker Sorge1

Watt, Stephen M.

110

A Generic Ontology for Spatial Frans Coenen and Pepijn Visser  

E-Print Network (OSTI)

, environmental impact assessment, shape fitting, timetabling and scheduling, and AI problems such as the N 1 #12; statement expressed using the generic ontology described here, and then to ``run be associated with such entities using the ontology. Finally in section 8 we present some conclusions

Coenen, Frans

111

Representative Source Terms and the Influence of Reactor Attributes on Functional Containment in Modular High-Temperature Gas-Cooled Reactors  

SciTech Connect

Modular high-temperature gas-cooled reactors (MHTGRs) offer a high degree of passive safety. The low power density of the reactor and the high heat capacity of the graphite core result in slow transients that do not challenge the integrity of the robust TRISO fuel. Another benefit of this fuel form and the surrounding graphite is their superior ability to retain fission products under all anticipated normal and off-normal conditions, which limits reactor accident source terms to very low values. In this paper, we develop estimates of the source term for a generic MHTGR to illustrate the performance of the radionuclide barriers that comprise the MHTGR functional containment. We also examine the influence of initial fuel quality, fuel performance/failure, reactor outlet temperature, and retention outside of the reactor core on the resultant source term to the environment.

D. A. Petti; Hans Gougar; Dick Hobbins; Pete Lowry

2013-11-01T23:59:59.000Z

112

Assessment of industrial attitudes toward generic research needs in tribology  

SciTech Connect

Based on extended discussions during visits with 27 companies representing 13 different parts of the tribology industry (such as bearings, lubricants, coatings, powerplants), it is apparent that only a tiny fraction of the large sums publicly reported as R and D expenditures by industry are used to fund generic tribology research. For example, of the greater than $2 B expenditures reported for R and D in the lubricants sector for 1982, the estimated total for generic tribology research was $12 M. This was the largest expenditure in any sector of the tribology industry and one-third of the total of $36 M. In the automotive industry out of a reported expenditure of $4 B, the estimated generic tribology research was $3 M. In some segments of the tribology industry, for example coatings and filters, there were no expenditures on generic research. There was little tendency to improve the state of the art of the tribology industry through long-term investment in generic R and D in ways that would foster innovation and productivity of energy conservation technology. Expenditures were oriented to development of specific commercial and military products, or to basic research focused on unspecified far term results, although useful spin-off of military developments into commercial fields sometimes occurs. There was a broad consensus in the companies visited that existing research results were not always made easily accessible to potential users in industry. The implication was that industry might benefit more if a larger fraction of the funds were devoted to putting the research results into a form design and development engineers could more readily apply. The need for a more effective presentation of research results was expressed with greater urgency at the smaller companies, but there seemed to be a broad consensus on the need for improvement. Recommendations are given.

Sibley, L.B.; Zlotnick, M.; Levinson, T.M.

1985-09-01T23:59:59.000Z

113

Generic TriBITS PRoject, Build, Test, and Install Quick Reference...  

NLE Websites -- All DOE Office Websites (Extended Search)

Generic TriBITS PRoject, Build, Test, and Install Quick Reference Guide Ross Bartlett Oak Ridge National Laboratory CASL-U-2014-0075-000-a CASL-U-2014-0075-000-a Generic TriBITS...

114

Improving support for generic programming in C# with associated types and constraint propagation  

E-Print Network (OSTI)

Generics has recently been adopted to many mainstream object oriented languages, such as C# and Java. As a particular design choice, generics in C# and Java use a sub-typing relation to constraint type parameters. Failing to encapsulate type...

Srinivasa Raghavan, Aravind

2009-05-15T23:59:59.000Z

115

A GENERIC AUDIO CLASSIFICATION AND SEGMENTATION APPROACH FOR MULTIMEDIA INDEXING AND RETRIEVAL  

E-Print Network (OSTI)

A GENERIC AUDIO CLASSIFICATION AND SEGMENTATION APPROACH FOR MULTIMEDIA INDEXING AND RETRIEVAL the attention on the area of generic and automatic audio classification and segmentation for audio audio classification and global segmentation framework based on automatic audio analysis providing

Gabbouj, Moncef

116

Attrition reactor system  

DOE Patents (OSTI)

A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.

Scott, C.D.; Davison, B.H.

1993-09-28T23:59:59.000Z

117

Elementary Reactor Physics  

Science Journals Connector (OSTI)

... THERE are few subjects which have developed at the rate at which reactor physics and ... physics and reactor theory have done. This, of course, is largely due to the circumstances in ...

J. F. HILL

1962-02-10T23:59:59.000Z

118

Colliding Beam Fusion Reactors  

Science Journals Connector (OSTI)

The recirculating power for virtually all types of fusion reactors has previously been calculated [1] with the Fokker–Planck equation. The reactors involve non-Maxwellian plasmas. The calculations are ... the rec...

Norman Rostoker; Artan Qerushi; Michl Binderbauer

2003-06-01T23:59:59.000Z

119

Prospects for spheromak fusion reactors  

Science Journals Connector (OSTI)

The reactor study of Hagenson and Krakowski demonstrated the attractiveness of the spheromak as a compact fusion reactor, based on...

T. K. Fowler; D. D. Hua

1995-06-01T23:59:59.000Z

120

Advanced Test Reactor Tour  

SciTech Connect

The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

Miley, Don

2011-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor generic distributed" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


121

Improved vortex reactor system  

DOE Patents (OSTI)

An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

Diebold, James P. (Lakewood, CO); Scahill, John W. (Evergreen, CO)

1995-01-01T23:59:59.000Z

122

Advanced Test Reactor Tour  

ScienceCinema (OSTI)

The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

Miley, Don

2013-05-28T23:59:59.000Z

123

Developing Generic Dynamic Models for the 2030 Eastern Interconnection Grid  

SciTech Connect

The Eastern Interconnection Planning Collaborative (EIPC) has built three major power flow cases for the 2030 Eastern Interconnection (EI) based on various levels of energy/environmental policy conditions, technology advances, and load growth. Using the power flow cases, this report documents the process of developing the generic 2030 dynamic models using typical dynamic parameters. The constructed model was validated indirectly using the synchronized phasor measurements by removing the wind generation temporarily.

Kou, Gefei [ORNL; Hadley, Stanton W [ORNL; Markham, Penn N [ORNL; Liu, Yilu [ORNL

2013-12-01T23:59:59.000Z

124

A simplified energy audit technique for generic buildings  

SciTech Connect

A simplified energy audit technique for generic buildings is developed. The premise is that buildings in similar climates and with similar patterns of use and thermal characteristics (structural mass and R-value) should have similar total energy usages if operated in an energy efficient manner. The deviation in the actual total energy usage from that expected for energy-efficient operation can then be examined, in conjunction with the findings of a building walk-through, in order to identify opportunities for energy conservation. The index used for comparison is the Energy Utilization Index (EUI), the total site energy used for per square foot per year. The procedure is illustrated by application to the generic class of Mississippi public school buildings. Validation is achieved by examining the sensitivity of building parameters and energy usage and by observations made in using the simplified procedure. The results show that the simplified technique gives results with as much fidelity as more labor intensive energy-auditing procedures and that the the technique is viable for generic classes of buildings.

Hodge, B.K.; Steele, W.G. Jr.; King, R.

1986-01-01T23:59:59.000Z

125

The Argonaut Reactor - Reactors designed/built by Argonne National  

NLE Websites -- All DOE Office Websites (Extended Search)

Achievements > Achievements > Argonne Reactors > Training Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy

126

Distributed Medium Access Control for Next Generation CDMA Wireless Networks  

E-Print Network (OSTI)

Distributed Medium Access Control for Next Generation CDMA Wireless Networks Hai Jiang, Princeton wireless networks are expected to have a simple infrastructure with distributed control. In this article, we consider a generic distributed network model for future wireless multi- media communications

Zhuang, Weihua

127

Simulating the generic job shop as a multi-agent system  

Science Journals Connector (OSTI)

A model combining Discrete Event System (DES) and Multi-Agent System (MAS) is proposed to simulate a real-time job shop so that it can work as a test bed to systematically study the performance of control rules and algorithms in dynamic job shop scheduling. A definition of a generic job shop is given considering dynamic events, shop floor layout and the Material Handling System (MHS). It is first simulated as a DES and then implemented as an MAS so that data recording and analysis can be naturally distributed to the most related entities and events can be executed simultaneously at different locations. The state changes of agents and the communication among them are illustrated with state charts and sequential diagrams in Unified Modelling Language (UML). The proposed model is validated with a case study through statistical analysis and comparison with reported works.

R. Zhou; H.P. Lee; A.Y.C. Nee

2008-01-01T23:59:59.000Z

128

Reactor vessel support system  

DOE Patents (OSTI)

A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

Golden, Martin P. (Trafford, PA); Holley, John C. (McKeesport, PA)

1982-01-01T23:59:59.000Z

129

Generic Signatures of the Time Profiles of BATSE Cosmic Gamma-Ray Bursts  

Science Journals Connector (OSTI)

A new method is proposed, which allows the study of generic signatures of cosmic gamma-ray burst time histories. We average the 64 ms resolution time profiles of 275 bright bursts detected by BATSE. The profile of each burst is normalized by the maximum number of counts at the peak of the primary pulse, and individual pulses and interpulse valleys are selected from the normalized profiles by identical selection criteria. New generic temporal parameters are introduced, which characterize the duration and equivalent width of each pulse and the duration of each valley. The histograms of the total equivalent pulse width and summed pulse duration are bimodal. Bimodality is also seen in the histogram of the mean duration of individual pulses. Bursts from the short and long peaks of these distributions correspond to the two modes of the Third BATSE Burst Catalog T50 and T90 distributions. Therefore, these new burst parameters demonstrate that the observed bimodal temporal behavior results from properties of the pulsed emission of gamma-ray bursts. The long mode of the T90 histogram includes bursts with from one to ~20 pulses; the logarithmic mean pulse duration is 1.17 ± 0.09 s; for the long events with more than one pulse, the logarithmic mean valley duration is 1.28 ± 0.15 s. Bursts of the short mode of T90 are mainly single-pulse events, and the logarithmic mean pulse duration is much smaller, 0.20 ± 0.01 s. For multipulse bursts of the T90 long mode, marginal correlations were found between the parameters of the pulses and valleys and the number of pulses. The basic signatures of the evolution of pulses and valleys along the time course of bursts are examined. Conclusions are drawn concerning the physics of gamma-ray emission by taking into account these signatures.

Igor; Alexei; Michael S. Briggs; William S. Paciesas; Robert D. Preece; Geoffrey; Charles

1998-01-01T23:59:59.000Z

130

Reactor water cleanup system  

DOE Patents (OSTI)

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

Gluntz, Douglas M. (San Jose, CA); Taft, William E. (Los Gatos, CA)

1994-01-01T23:59:59.000Z

131

Reactor water cleanup system  

DOE Patents (OSTI)

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

Gluntz, D.M.; Taft, W.E.

1994-12-20T23:59:59.000Z

132

Influence of gas flow rate on liquid distribution in trickle-beds using perforated plates as liquid distributors  

E-Print Network (OSTI)

" the distribution imposed at the top of the reactor. Finally, a comparison between the two measuring techniques-beds reactors, the second will directly affect its performances. Indeed, a bad liquid distribution will not only distribution when fluids distribution on top of the reactor is ensured by a perforated plate. In opposition

Paris-Sud XI, Université de

133

Regulatory analysis for the resolution of Generic Safety Issue 29: Bolting degradation or failure in nuclear power plants  

SciTech Connect

Generic Safety Issue (GSI)-29 deals with staff concerns about public risk due to degradation or failure of safety-related bolting in nuclear power plants. The issue was initiated in November 1982. Value-impact studies of a mandatory program on safety-related bolting for operating plants were inconclusive: therefore, additional regulatory requirements for operating plants could not be justified in accordance with provisions of 10 CFR 50.109. In addition, based on operating experience with bolting in both nuclear and conventional power plants, the actions already taken through bulletins, generic letters, and information notices, and the industry-proposed actions, the staff concluded that a sufficient technical basis exists for the resolution of GSI-29. The staff further concluded that leakage of bolted pressure joints is possible but catastrophic failure of a reactor coolant pressure boundary joint that will lead to significant accident sequences is highly unlikely. For future plants, it was concluded that a new Standard Review Plant section should be developed to codify existing bolting requirements and industry-developed initiatives. 9 refs., 1 tab.

Chang, T.Y.

1991-09-01T23:59:59.000Z

134

Flow Simulation and Optimization of Plasma Reactors for Coal Gasification  

Science Journals Connector (OSTI)

This paper reports a 3-d numerical simulation system to analyze the complicated flow in plasma reactors for coal gasification, which involve complex chemical reaction, two-phase flow and plasma effect. On the basis of analytic results, the distribution of the density, temperature and components' concentration are obtained and a different plasma reactor configuration is proposed to optimize the flow parameters. The numerical simulation results show an improved conversion ratio of the coal gasification. Different kinds of chemical reaction models are used to simulate the complex flow inside the reactor. It can be concluded that the numerical simulation system can be very useful for the design and optimization of the plasma reactor.

Ji Chunjun; Zhang Yingzi; Ma Tengcai

2003-01-01T23:59:59.000Z

135

A GENERIC MODEL OF A BASE-ISOLATED BUILDING This chapter draws together the work of Chapters 3 and 4 and describes the assembly of a generic  

E-Print Network (OSTI)

145 Chapter 5 A GENERIC MODEL OF A BASE-ISOLATED BUILDING This chapter draws together the work of Chapters 3 and 4 and describes the assembly of a generic model of a base-isolated building. The first section describes an existing two-dimensional model of a building, which is based on the dynamic

Talbot, James P.

136

Experiments and analysis for an axially heterogeneous liquid-metal reactor assembly at the zero-power physics reactor  

SciTech Connect

Experiments in zero-power physics reactor 17 provided physics data for a full-scale axially heterogeneous 650-MW(electric) liquid-metal reactor. Measurements and analysis are reported for control rod worths, reaction rate distributions, gamma dose distributions, sodium void worths, and criticality. Agreement between measurement and calculation is generally satisfactory, but the axial heterogeneity did introduce analytical complications. Design-level calculation methods gave somewhat worse agreement with measurement than in previous homogeneous or radially heterogeneous assemblies.

Brumbach, S.B.; Collins, P.J. (Argonne National Lab., Idaho Falls, ID (USA))

1989-11-01T23:59:59.000Z

137

Solar solids reactor  

DOE Patents (OSTI)

A solar powered kiln is provided, that is of relatively simple design and which efficiently uses solar energy. The kiln or solids reactor includes a stationary chamber with a rearward end which receives solid material to be reacted and a forward end through which reacted material is disposed of, and a screw conveyor extending along the bottom of the chamber for slowly advancing the material between the chamber ends. Concentrated solar energy is directed to an aperture at the forward end of the chamber to heat the solid material moving along the bottom of the chamber. The solar energy can be reflected from a mirror facing at an upward incline, through the aperture and against a heat-absorbing material near the top of the chamber, which moves towards the rear of the chamber to distribute heat throughout the chamber. Pumps at the forward and rearward ends of the chamber pump heated sweep gas through the length of the chamber, while minimizing the flow of gas through an open aperture through which concentrated sunlight is received.

Yudow, Bernard D. (Chicago, IL)

1987-01-01T23:59:59.000Z

138

Solar solids reactor  

DOE Patents (OSTI)

A solar powered kiln is provided, that is of relatively simple design and which efficiently uses solar energy. The kiln or solids reactor includes a stationary chamber with a rearward end which receives solid material to be reacted and a forward end through which reacted material is disposed of, and a screw conveyor extending along the bottom of the chamber for slowly advancing the material between the chamber ends. Concentrated solar energy is directed to an aperture at the forward end of the chamber to heat the solid material moving along the bottom of the chamber. The solar energy can be reflected from a mirror facing at an upward incline, through the aperture and against a heat-absorbing material near the top of the chamber, which moves towards the rear of the chamber to distribute heat throughout the chamber. Pumps at the forward and rearward ends of the chamber pump heated sweep gas through the length of the chamber, while minimizing the flow of gas through an open aperture through which concentrated sunlight is received.

Yudow, B.D.

1986-02-24T23:59:59.000Z

139

Power Reactor Progress  

Science Journals Connector (OSTI)

Argonne kicks off EBWR; Allis-Chalmers plans power reactor using both nuclear and conventional fuels ... NUCLEAR POWER took two giant steps last week. ... Just as the first nuclear power system in the U. S. designed and built solely for the generation of electric power went into full operation at Argonne, Allis-Chalmers came up with a new twist in power reactors—a controlled recirculation boiling reactor (CRBR) using both nuclear and conventional fuels (C&EN, Feb. 18, page 7). ...

1957-02-25T23:59:59.000Z

140

Improved vortex reactor system  

DOE Patents (OSTI)

An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

Diebold, J.P.; Scahill, J.W.

1995-05-09T23:59:59.000Z

Note: This page contains sample records for the topic "reactor generic distributed" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


141

AEC Pushes Fusion Reactors  

Science Journals Connector (OSTI)

AEC Pushes Fusion Reactors ... Project Sherwood, as the study program is called, began in 1951-52 soon after the first successful thermonuclear explosion in the Pacific. ...

1955-10-10T23:59:59.000Z

142

Tokamak reactor first wall  

DOE Patents (OSTI)

This invention relates to an improved first wall construction for a tokamak fusion reactor vessel, or other vessels subjected to similar pressure and thermal stresses.

Creedon, R.L.; Levine, H.E.; Wong, C.; Battaglia, J.

1984-11-20T23:59:59.000Z

143

Advanced Nuclear Research Reactor  

SciTech Connect

This report describes technical modifications implemented by INVAP to improve the safety of the Research Reactors the company designs and builds.

Lolich, J.V.

2004-10-06T23:59:59.000Z

144

Relap5-3d model validation and benchmark exercises for advanced gas cooled reactor application  

E-Print Network (OSTI)

HTTR High Temperature engineering Test Reactor INET Institute of Nuclear Energy Technology LWR Light Water Reactor OKBM Test Design Bureau for Machine Building ORNL Oak Ridge National Laboratory RCCS Reactor Cavity Cooling System... to be at right angles to each other, ignoring an angular distribution of radiant heat.7 MORECA, used by ORNL, simulates accident scenarios for certain gas-cooled reactor types.7 INET conducts their analysis using Thermix, which performs two...

Moore, Eugene James Thomas

2006-08-16T23:59:59.000Z

145

The siting of UK nuclear reactors  

Science Journals Connector (OSTI)

Choosing a suitable site for a nuclear power station requires the consideration and balancing of several factors. Some 'physical' site characteristics, such as the local climate and the potential for seismic activity, will be generic to all reactors designs, while others, such as the availability of cooling water, the area of land required and geological conditions capable of sustaining the weight of the reactor and other buildings will to an extent be dependent on the particular design of reactor chosen (or alternatively the reactor design chosen may to an extent be dependent on the characteristics of an available site). However, one particularly interesting tension is a human and demographic one. On the one hand it is beneficial to place nuclear stations close to centres of population, to reduce transmission losses and other costs (including to the local environment) of transporting electricity over large distances from generator to consumer. On the other it is advantageous to place nuclear stations some distance away from such population centres in order to minimise the potential human consequences of a major release of radioactive materials in the (extremely unlikely) event of a major nuclear accident, not only in terms of direct exposure but also concerning the management of emergency planning, notably evacuation.This paper considers the emergence of policies aimed at managing this tension in the UK. In the first phase of nuclear development (roughly speaking 1945–1965) there was a highly cautious attitude, with installations being placed in remote rural locations with very low population density. The second phase (1965–1985) saw a more relaxed approach, allowing the development of AGR nuclear power stations (which with concrete pressure vessels were regarded as significantly safer) closer to population centres (in 'semi-urban' locations, notably at Hartlepool and Heysham). In the third phase (1985–2005) there was very little new nuclear development, Sizewell B (the first and so far only PWR power reactor in the UK) being colocated with an early Magnox station on the rural Suffolk coast. Renewed interest in nuclear new build from 2005 onward led to a number of sites being identified for new reactors before 2025, all having previously hosted nuclear stations and including the semi-urban locations of the 1960s and 1970s. Finally, some speculative comments are made as to what a 'fifth phase' starting in 2025 might look like.

Malcolm Grimston; William J Nuttall; Geoff Vaughan

2014-01-01T23:59:59.000Z

146

A TEN MEGAWATT BOILING HETEROGENEOUS PACKAGE POWER REACTOR. Reactor...  

Office of Scientific and Technical Information (OSTI)

A TEN MEGAWATT BOILING HETEROGENEOUS PACKAGE POWER REACTOR. Reactor Design and Feasibility Problem Re-direct Destination: Temp Data Fields Rosen, M. A.; Coburn, D. B.; Flynn, T....

147

Portfolio for fast reactor collaboration  

SciTech Connect

The development of the LMFBR type reactor in the United Kingdom is reviewed. Design characteristics of a commercial demonstration fast reactor are presented and compared with the Super Phenix reactor.

Rippon, S.

1981-12-01T23:59:59.000Z

148

Handbook of Reactor Physics  

Science Journals Connector (OSTI)

... THIS handbook is one volume in a series sponsored by the United States Atomic Energy Commission with ... data and reference information in the field of reactors. The volume is devoted to reactor physics and radiation shielding, the latter subject occupying approximately a quarter of the book.

PETER W. MUMMERY

1956-08-25T23:59:59.000Z

149

Fast reactor safety  

Science Journals Connector (OSTI)

... SIR, - In his article on fast reactor safety (26 July, page 270) Norman Dombey claims to introduce to non-specialists ... , page 270) Norman Dombey claims to introduce to non-specialists some features of fast reactors that are not available outside the technical literature. The non-specialist would do well ...

R.D. SMITH

1979-08-23T23:59:59.000Z

150

Instrumentation of Nuclear Reactors  

Science Journals Connector (OSTI)

... s Lecture Theatre on January 8, a symposium of papers on the instrumentation of nuclear reactors was organized, at which about five hundred members and visitors attended, including guests from ... the Institution, took the chair and introduced Sir John Cockcroft, whose lecture on "Nuclear Reactors and their Applications" provided a general background for the three specialized papers which followed. ...

1953-03-07T23:59:59.000Z

151

Nuclear Research Reactors  

Science Journals Connector (OSTI)

... their countries for the advent of nuclear power. A few countries had built large research reactors for the production of isotopes and to study the behaviour of nuclear fuel, but ... production of isotopes and to study the behaviour of nuclear fuel, but the small training reactor had not been developed. Since then, research ...

T. E. ALLIBONE

1963-07-20T23:59:59.000Z

152

Canadian university research reactors  

SciTech Connect

In Canada there are seven university research reactors: one medium-power (2-MW) swimming pool reactor at McMaster University and six low-power (20-kW) SLOWPOKE reactors at Dalhousie University, Ecole Polytechnique, the Royal Military College, the University of Toronto, the University of Saskatchewan, and the University of Alberta. This paper describes primarily the McMaster Nuclear Reactor (MNR), which operates on a wider scale than the SLOWPOKE reactors. The MNR has over a hundred user groups and is a very broad-based tool. The main applications are in the following areas: (1) neutron activation analysis (NAA); (2) isotope production; (3) neutron beam research; (4) nuclear engineering; (5) neutron radiography; and (6) nuclear physics.

Ernst, P.C.; Collins, M.F.

1989-11-01T23:59:59.000Z

153

Impact of Generic Advertising on Brand Advertising in Food and Agricultural Industries.  

E-Print Network (OSTI)

??Generic advertising programs managed by several agricultural commodity groups are collective economic actions to promote agricultural products. Several previous studies have shown the effectiveness of… (more)

Suh, Daeseok

2010-01-01T23:59:59.000Z

154

Reactor & Nuclear Systems Publications | ORNL  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Science Home | Science & Discovery | Nuclear Science | Publications and Reports | Reactor and Nuclear Systems Publications SHARE Reactor and Nuclear Systems Publications...

155

An effectual template bank for the detection of gravitational waves from inspiralling compact binaries with generic spins  

E-Print Network (OSTI)

We report the construction of a three-dimensional template bank for the search for gravitational waves from inspiralling binaries consisting of spinning compact objects. The parameter space consists of two dimensions describing the mass parameters and one "reduced-spin" parameter, which describes the secular (non-precessing) spin effects in the waveform. The template placement is based on an efficient stochastic algorithm and makes use of the semi-analytical computation of a metric in the parameter space. We demonstrate that for "low-mass" ($m_1 + m_2 \\lesssim 12\\,M_\\odot$) binaries, this template bank achieves effective fitting factors $\\sim0.92$--$0.99$ towards signals from generic spinning binaries in the advanced detector era over the entire parameter space of interest (including binary neutron stars, binary black holes, and black hole-neutron star binaries). This provides a powerful and viable method for searching for gravitational waves from generic spinning low-mass compact binaries. Under the assumption that spin magnitudes of black-holes [neutron-stars] are uniformly distributed between 0--0.98 [0 -- 0.4] and spin angles are isotropically distributed, the expected improvement in the average detection volume (at a fixed signal-to-noise-ratio threshold) of a search using this reduced-spin bank is $\\sim20-52\\%$, as compared to a search using a non-spinning bank.

P. Ajith; N. Fotopoulos; S. Privitera; A. Neunzert; N. Mazumder; A. J. Weinstein

2014-05-21T23:59:59.000Z

156

Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1  

SciTech Connect

The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner`s Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized into six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section.

NONE

1995-08-01T23:59:59.000Z

157

DOE-STD-0100T; DOE Standard Licensed Reactor Nuclear Safety Criteria Applicable to DOE Reactors  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

00T 00T November 1993 Superseding DOE/NE-0100T April 1991 DOE STANDARD LICENSED REACTOR NUCLEAR SAFETY CRITERIA APPLICABLE TO DOE REACTORS U.S. Department of Energy Washington, D.C. 20585 AREA SAFT DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. This document has been reproduced directly frorn the best available copy. Available to DOE and DOE contractors from the Office of Scientific and Technical Information, P.O. Box 62, Oak Ridge, TN 37831; (615) 576-8401. Available to the public from the U.S. Department of Commerce, Technology Administration, National Technical Information Service, Springfield, VA 22161; (703) 487-4650. Order No. DE94005221 CONTENTS

158

A safety and regulatory assessment of generic BWR and PWR permanently shutdown nuclear power plants  

SciTech Connect

The long-term availability of less expensive power and the increasing plant modification and maintenance costs have caused some utilities to re-examine the economics of nuclear power. As a result, several utilities have opted to permanently shutdown their plants. Each licensee of these permanently shutdown (PSD) plants has submitted plant-specific exemption requests for those regulations that they believe are no longer applicable to their facility. This report presents a regulatory assessment for generic BWR and PWR plants that have permanently ceased operation in support of NRC rulemaking activities in this area. After the reactor vessel is defueled, the traditional accident sequences that dominate the operating plant risk are no longer applicable. The remaining source of public risk is associated with the accidents that involve the spent fuel. Previous studies have indicated that complete spent fuel pool drainage is an accident of potential concern. Certain combinations of spent fuel storage configurations and decay times, could cause freshly discharged fuel assemblies to self heat to a temperature where the self sustained oxidation of the zircaloy fuel cladding may cause cladding failure. This study has defined four spent fuel configurations which encompass all of the anticipated spent fuel characteristics and storage modes following permanent shutdown. A representative accident sequence was chosen for each configuration. Consequence analyses were performed using these sequences to estimate onsite and boundary doses, population doses and economic costs. A list of candidate regulations was identified from a screening of 10 CFR Parts 0 to 199. The continued applicability of each regulation was assessed within the context of each spent fuel storage configuration and the results of the consequence analyses.

Travis, R.J.; Davis, R.E.; Grove, E.J.; Azarm, M.A. [Brookhaven National Lab., Upton, NY (United States)

1997-08-01T23:59:59.000Z

159

The missing cavities in the SEEDS polarized scattered light images of transitional protoplanetary disks: a generic disk model  

E-Print Network (OSTI)

Transitional circumstellar disks around young stellar objects have a distinctive infrared deficit around 10 microns in their Spectral Energy Distributions (SED), recently measured by the Spitzer Infrared Spectrograph (IRS), suggesting dust depletion in the inner regions. These disks have been confirmed to have giant central cavities by imaging of the submillimeter (sub-mm) continuum emission using the Submillimeter Array (SMA). However, the polarized near-infrared scattered light images for most objects in a systematic IRS/SMA cross sample, obtained by HiCIAO on the Subaru telescope, show no evidence for the cavity, in clear contrast with SMA and Spitzer observations. Radiative transfer modeling indicates that many of these scattered light images are consistent with a smooth spatial distribution for micron-sized grains, with little discontinuity in the surface density of the micron-sized grains at the cavity edge. Here we present a generic disk model that can simultaneously account for the general features in...

Dong, R; Zhu, Z; Hartmann, L; Whitney, B; Brandt, T; Muto, T; Hashimoto, J; Grady, C; Follette, K; Kuzuhara, M; Tanii, R; Itoh, Y; Thalmann, C; Wisniewski, J; Mayama, S; Janson, M; Abe, L; Brandner, W; Carson, J; Egner, S; Feldt, M; Goto, M; Guyon, O; Hayano, Y; Hayashi, M; Hayashi, S; Henning, T; Hodapp, K W; Honda, M; Inutsuka, S; Ishii, M; Iye, M; Kandori, R; Knapp, G R; Kudo, T; Kusakabe, N; Matsuo, T; McElwain, M W; Miyama, S; Morino, J -I; Moro-Martin, A; Nishimura, T; Pyo, T -S; Suto, H; Suzuki, R; Takami, M; Takato, N; Terada, H; Tomono, D; Turner, E L; Watanabe, M; Yamada, T; Takami, H; Usuda, T; Tamura, M

2012-01-01T23:59:59.000Z

160

Nuclear reactor control column  

DOE Patents (OSTI)

The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

Bachovchin, Dennis M. (Plum Borough, PA)

1982-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor generic distributed" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


161

Reactor Safety Research Programs  

SciTech Connect

This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

Edler, S. K.

1981-07-01T23:59:59.000Z

162

Five Years of Building the Next Generation of Reactors | Department of  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Five Years of Building the Next Generation of Reactors Five Years of Building the Next Generation of Reactors Five Years of Building the Next Generation of Reactors August 15, 2012 - 5:17pm Addthis Simulated three-dimensional fission power distribution of a single 17x17 rod PWR fuel assembly. | Photo courtesy of the Consortium for Advanced Simulation of Light Water Reactors (CASL). Simulated three-dimensional fission power distribution of a single 17x17 rod PWR fuel assembly. | Photo courtesy of the Consortium for Advanced Simulation of Light Water Reactors (CASL). Doug Kothe Director, Consortium for Advanced Simulation of Light Water Reactors What are the key facts? CASL has the virtual capability to look closely at reactor core models. These models operate with 193 fuel assemblies, nearly 51,000 fuel rods, and about 18 million fuel pellets.

163

Nuclear reactor reflector  

DOE Patents (OSTI)

A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled.

Hopkins, Ronald J. (Pensacola, FL); Land, John T. (Pensacola, FL); Misvel, Michael C. (Pensacola, FL)

1994-01-01T23:59:59.000Z

164

Spherical torus fusion reactor  

DOE Patents (OSTI)

The object of this invention is to provide a compact torus fusion reactor with dramatic simplification of plasma confinement design. Another object of this invention is to provide a compact torus fusion reactor with low magnetic field and small aspect ratio stable plasma confinement. In accordance with the principles of this invention there is provided a compact toroidal-type plasma confinement fusion reactor in which only the indispensable components inboard of a tokamak type of plasma confinement region, mainly a current conducting medium which carries electrical current for producing a toroidal magnet confinement field about the toroidal plasma region, are retained.

Martin Peng, Y.K.M.

1985-10-03T23:59:59.000Z

165

The use of personal computers in reactor physics  

SciTech Connect

This paper points out that personal computers are now powerful enough (in terms of core size and speed) to allow them to be used for serious reactor physics applications. In addition the low cost of personal computers means that even small institutes can now have access to a significant amount of computer power. At the present time distribution centers, such as RSIC, are beginning to distribute reactor physics codes for use on personal computers; hopefully in the near future more and more of these codes will become available through distribution centers, such as RSIC.

Cullen, D.E.

1988-01-01T23:59:59.000Z

166

Distributed Algorithms Distributed Transactions  

E-Print Network (OSTI)

Algorithms© Gero Mühl 8 Concurrency Control serial RC (ReCoverable) ACA (Avoiding Cascading Aborts) ST (StricDistributed Algorithms Distributed Transactions PD Dr.-Ing. Gero Mühl Kommunikations- und Betriebssysteme Fakultät für Elektrotechnik u. Informatik Technische Universität Berlin #12;Distributed Algorithms

Wichmann, Felix

167

Incipient Transient Detection in Reactor Systems: Experimental and Theoretical Investigation  

SciTech Connect

The main goal of this research was to develop a method for detecting reactor system transients at the earliest possible time through a comprehensive experimental, testing and benchmarking program. This approach holds strong promise for developing new diagnostic technologies that are non-intrusive, generic and highly portable across different systems. It will help in the design of new generation nuclear power reactors, which utilize passive safety systems with a reliable and non-intrusive multiphase flow diagnostic system to monitor the function of the passive safety systems. The main objective of this research was to develop an improved fuzzy logic based detection method based on a comprehensive experimental testing program to detect reactor transients at the earliest possible time, practically at their birth moment. A fuzzy logic and neural network based transient identification methodology and implemented in a computer code called PROTREN was considered in this research and was compared with SPRT (Sequentially Probability Ratio Testing) decision and Bayesian inference. The project involved experiment, theoretical modeling and a thermal-hydraulic code assessment. It involved graduate and undergraduate students participation providing them with exposure and training in advanced reactor concepts and safety systems. In this final report, main tasks performed during the project period are summarized and the selected results are presented. Detailed descriptions for the tasks and the results are presented in previous yearly reports (Revankar et al 2003 and Revankar et al 2004).

Lefteri H. Tsoukalas; S.T. Revankar; X Wang; R. Sattuluri

2005-09-27T23:59:59.000Z

168

Analysis, Design and Development of a Generic Framework for Power Trading  

E-Print Network (OSTI)

in their producing plans (e.g. wind power plant). The bid model is the best model for clients with units havingAnalysis, Design and Development of a Generic Framework for Power Trading Rasmus Skovmark, s001509 analysis, design and development of a generic framework for automatic real- time trading of power among

169

Towards A Theory-Of-Mind-Inspired Generic Decision-Making Framework Mihai Polceanu, Cedric Buche  

E-Print Network (OSTI)

Towards A Theory-Of-Mind-Inspired Generic Decision-Making Framework Mihai Polceanu, C´edric Buche of simulation as a decision-making technique, we propose a generic framework based on theory of mind, which the framework are discussed. 1 Introduction When attempting to make a suitable decision within an en- vironment

Paris-Sud XI, Université de

170

Short-Term Audio-Visual Atoms for Generic Video Concept Classification  

E-Print Network (OSTI)

Short-Term Audio-Visual Atoms for Generic Video Concept Classification Wei Jiang1 Courtenay Cotton1 the challenging issue of joint audio-visual analysis of generic videos targeting at semantic concept de- tection. We propose to extract a novel representation, the Short-term Audio-Visual Atom (S-AVA), for improved

Ellis, Dan

171

Reactor Thermal-Hydraulics  

NLE Websites -- All DOE Office Websites (Extended Search)

Thermal-Hydraulics Thermal-Hydraulics Dr. Tanju Sofu, Argonne National Laboratory In a power reactor, the energy produced in fission reaction manifests itself as heat to be removed by a coolant and utilized in a thermodynamic energy conversion cycle to produce electricity. A simplified schematic of a typical nuclear power plant is shown in the diagram below. Primary coolant loop Steam Reactor Heat exchanger Primary pump Secondary pump Condenser Turbine Water Although this process is essentially the same as in any other steam plant configuration, the power density in a nuclear reactor core is typically four orders of magnitude higher than a fossil fueled plant and therefore it poses significant heat transfer challenges. Maximum power that can be obtained from a nuclear reactor is often limited by the

172

Reactor hot spot analysis  

SciTech Connect

The principle methods for performing reactor hot spot analysis are reviewed and examined for potential use in the Applied Physics Division. The semistatistical horizontal method is recommended for future work and is now available as an option in the SE2-ANL core thermal hydraulic code. The semistatistical horizontal method is applied to a small LMR to illustrate the calculation of cladding midwall and fuel centerline hot spot temperatures. The example includes a listing of uncertainties, estimates for their magnitudes, computation of hot spot subfactor values and calculation of two sigma temperatures. A review of the uncertainties that affect liquid metal fast reactors is also presented. It was found that hot spot subfactor magnitudes are strongly dependent on the reactor design and therefore reactor specific details must be carefully studied. 13 refs., 1 fig., 5 tabs.

Vilim, R.B.

1985-08-01T23:59:59.000Z

173

Molten metal reactors  

DOE Patents (OSTI)

A molten metal reactor for converting a carbon material and steam into a gas comprising hydrogen, carbon monoxide, and carbon dioxide is disclosed. The reactor includes an interior crucible having a portion contained within an exterior crucible. The interior crucible includes an inlet and an outlet; the outlet leads to the exterior crucible and may comprise a diffuser. The exterior crucible may contain a molten alkaline metal compound. Contained between the exterior crucible and the interior crucible is at least one baffle.

Bingham, Dennis N; Klingler, Kerry M; Turner, Terry D; Wilding, Bruce M

2013-11-05T23:59:59.000Z

174

F Reactor Inspection  

ScienceCinema (OSTI)

Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosure and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2014-11-24T23:59:59.000Z

175

F Reactor Inspection  

SciTech Connect

Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosure and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2014-10-29T23:59:59.000Z

176

Hybrid energy systems (HESs) using small modular reactors (SMRs)  

SciTech Connect

Large-scale nuclear reactors are traditionally operated for a singular purpose: steady-state production of dispatchable baseload electricity that is distributed broadly on the electric grid. While this implementation is key to a sustainable, reliable energy grid, small modular reactors (SMRs) offer new opportunities for increased use of clean nuclear energy for both electric and thermal ap plications in more locations – while still accommodating the desire to support renewable production sources.

S. Bragg-Sitton

2014-10-01T23:59:59.000Z

177

Linear quadratic regulator for time-varying hyperbolic distributed parameter systems  

Science Journals Connector (OSTI)

......concentration along the reactor are shown in Figure 2. To assess the reliability of the above algorithm...Contribution to the analysis and control of distributed...2001) Trajectory analysis of nonisothermal tubular reactor nonlinear models......

Ilyasse Aksikas; J. Fraser Forbes

2010-09-01T23:59:59.000Z

178

A Generic Model for the Resuspension of Multilayer Aerosol Deposits by Turbulent Flow  

SciTech Connect

An idealized lattice structure is considered of multilayer aerosol deposits, where every particle at the deposit surface is associated with a resuspension rate constant depending on a statistically distributed particle parameter and on flow conditions. The response of this generic model is represented by a set of integrodifferential equations. As a first application of the general formalism, the behavior of Fromentin's multilayer model is analyzed, and the model parameters are adapted to experimental data. In addition, improved relations between model parameters and physical input parameters are proposed. As a second application, a method is proposed for building multilayer models by using resuspension rate constants of existing monolayer models. The method is illustrated by a sample of monolayer data resulting from the model of Reeks, Reed, and Hall. Also discussed is the error to be expected if a monolayer resuspension model, which works well for thin aerosol deposits, is applied to thick deposits under the classical monolayer assumption that all deposited particles interact with the fluid at all times.

Friess, H.; Yadigaroglu, G. [Swiss Federal Institute of Technology (Switzerland)

2001-06-15T23:59:59.000Z

179

Solar Thermal Reactor Materials Characterization  

SciTech Connect

Current research into hydrogen production through high temperature metal oxide water splitting cycles has created a need for robust high temperature materials. Such cycles are further enhanced by the use of concentrated solar energy as a power source. However, samples subjected to concentrated solar radiation exhibited lifetimes much shorter than expected. Characterization of the power and flux distributions representative of the High Flux Solar Furnace(HFSF) at the National Renewable Energy Laboratory(NREL) were compared to ray trace modeling of the facility. In addition, samples of candidate reactor materials were thermally cycled at the HFSF and tensile failure testing was performed to quantify material degradation. Thermal cycling tests have been completed on super alloy Haynes 214 samples and results indicate that maximum temperature plays a significant role in reduction of strength. The number of cycles was too small to establish long term failure trends for this material due to the high ductility of the material.

Lichty, P. R.; Scott, A. M.; Perkins, C. M.; Bingham, C.; Weimer, A. W.

2008-03-01T23:59:59.000Z

180

The B & W Owners` Group Generic License Renewal Program  

SciTech Connect

Since the late 1970s, the Babcock & Wilcox (B & W) Owners Group (BWOG) has sponsored significant activities that address technical, economic, and licensing issues to ensure that the B & W nuclear steam supply system (NSSS) power plants operate until the end of their current plant licensed life and to preserve the license renewal option. It should be no surprise that the BWOG decided in late 1992 to aggressively pursue a license renewal effort. This effort, the Generic License Renewal Program (GLRP), has over the past 18 months contributed significantly to the industry`s license renewal initiative. The GLRP was established as a project with a full-time management organization within the BWOG structure. Its primary objective was the development and demonstration of an integrated plant assessment (IPA) process that would meet the requirements of the License Renewal Rule, published by the US Nuclear Regulatory Commission (NRC) in December 1991. The BWOG consists of five utilities with plants of very similar design, operation, and age. The owners, along with technical support from B & W Nuclear Technologies, created a highly capable and effective team to address the elements of the license renewal rule. This paper presents the BWOG strategy from the beginning of the program, the accomplishments to date, and the current role of the BWOG GLRP.

Gill, R.L. Jr.

1994-12-31T23:59:59.000Z

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181

Generic effective source for scalar self-force calculations  

E-Print Network (OSTI)

A leading approach to the modelling of extreme mass ratio inspirals involves the treatment of the smaller mass as a point particle and the computation of a regularized self-force acting on that particle. In turn, this computation requires knowledge of the regularized retarded field generated by the particle. A direct calculation of this regularized field may be achieved by replacing the point particle with an effective source and solving directly a wave equation for the regularized field. This has the advantage that all quantities are finite and require no further regularization. In this work, we present a method for computing an effective source which is finite and continuous everywhere, and which is valid for a scalar point particle in arbitrary geodesic motion in an arbitrary background spacetime. We explain in detail various technical and practical considerations that underlie its use in several numerical self-force calculations. We consider as examples the cases of a particle in a circular orbit about Schwarzschild and Kerr black holes, and also the case of a particle following a generic time-like geodesic about a highly spinning Kerr black hole. We provide numerical C code for computing an effective source for various orbital configurations about Schwarzschild and Kerr black holes.

Barry Wardell; Ian Vega; Jonathan Thornburg; Peter Diener

2011-12-29T23:59:59.000Z

182

Generic Disposal System Modeling--Fiscal Year 2011 Progress Report  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

12/2011 12/2011 Rev. 2 FCRD- T10-201 0-00005 FCRD Technical Integration Office (TIO) DOCUMENT NUMBER REQUEST TRANSMITTAL SHEET 1. Document Information Document Title/Description: Generic Disposal System Modeling--Fiscal Year 2011 Revision: 0 Progress Re~ort Assigned Document Number: FCRD-USED-2011-000184 Effective Start Date: 08/1112011 Document Author/Creator: D. Clayton, G. Freeze, T. Hadgu, E. Hardin, J. Lee, OR .~: J. Prouty, R. Rogers, W.M. Nutt, J. Berkholzer, H.H. Liu, L. Zheng, S. Chu Document Owner: Palmer Vaughn Date Range: Originating Organization: Sandia National Laboratories From: To: Milestone OM1 ~M2 OM3 ~M4 o Not a Milestone Milestone Number:: M21UF034101 and M41UF035102 Work Package WBS Number: FTSN11 UF0341 and FTSN11 UF0351; 1.02.08.03 Controlled Unclassified Infonnation (CUI) Type ~ None OOUO OAT o Other FCRD SYSTEM: Year: o FUEL Fuels 2011 OINTL International

183

REACTOR GROUT THERMAL PROPERTIES  

SciTech Connect

Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.

Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.

2011-01-28T23:59:59.000Z

184

Reactor Safety Planning for Prometheus Project, for Naval Reactors Information  

SciTech Connect

The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

P. Delmolino

2005-05-06T23:59:59.000Z

185

1 MODELING THE PERFORMANCE OF ULTRAVIOLET REACTOR IN EULERIAN AND LAGRANGIAN FRAMEWORKS  

E-Print Network (OSTI)

CFD models for simulating the performance of ultraviolet (UV) reactors for micro-organism inactivation were developed in Eulerian and Lagrangian frameworks, taking into account hydrodynamics, kinetics, and radiation field within UV reactor. In the Lagrangian framework, micro-organisms were treated as discrete particles where the trajectory was predicted by integrating the force balance on the particle. In the Eulerian framework, the conservation equation of species (microorganisms) was solved along with the transport equations. The fluid flow was characterized experimentally using particle image velocimetry (PIV) flow visualization techniques and modeled using CFD for a UV reactor prototype model. The performance of annular UV reactors with an inlet parallel and perpendicular to the reactor axis were investigated. The results indicated that the fluid flow distribution within the reactor volume can significantly affect the reactor performance. Both the Eulerian and Lagrangian models were used to obtain complimentary information on the reactors; while the Lagrangian method provided an estimation of the UV-fluence distribution and the trajectory of species, the Eulerian approach showed the concentration distribution and local photo-reaction rates. The combined information can be used to predict and monitor reactor performance and to improve the reactor design.

Angelo Sozzi; Fariborz Taghipour

2006-01-01T23:59:59.000Z

186

Export possibilities for small nuclear reactors  

SciTech Connect

The worldwide deployment of peaceful nuclear technology is predicated on conformance with the Nuclear Non-Proliferation Treaty of 1972. Under this international treaty, countries have traded away pursuit of nuclear weapons in exchange for access to commercial nuclear technology that could help them grow economically. Realistically, however, most nuclear technology has been beyond the capacity of the NPT developing countries to afford. Even if the capital cost of the plant is managed, the costs of the infrastructure and the operational complexity of most nuclear technology have taken it out of the hands of the nations who need it the most. Now, a new class of small sodium cooled reactors has been specifically designed to meet the electrical power, water, hydrogen and heat needs of small and remote users. These reactors feature small size, long refueling interval, no onsite fuel storage, and simplified operations. Sized in the 10 MW(e) to 50 MW(e) range these reactors are modularized for factory production and for rapid site assembly. The fuel would be <20% U-235 uranium fuel with a 30-year core life. This new reactor type more appropriately fills the needs of countries for lower power distributed systems that can fill the gap between large developed infrastructure and primitive distributed energy systems. Looking at UN Resolution 1540 and the impact of other agreements, there is a need to address the issues of nuclear security, fuel, waste, and economic/legal/political-stakeholder concerns. This paper describes the design features of this new reactor type that specifically address these issues in a manner that increases the availability of commercial nuclear technology to the developing nations of the world. (authors)

Campagna, M.S.; Hess, C.; Moor, P. [Burns and Roe Enterprises, Inc., Oradell, NJ (United States); Sawruk, W. [ABSG Consulting, Inc., Shillington, PA (United States)

2007-07-01T23:59:59.000Z

187

B Reactor | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Operational Management » History » Manhattan Project » Signature Operational Management » History » Manhattan Project » Signature Facilities » B Reactor B Reactor B Reactor Completed in September 1944, the B Reactor was the world's first large-scale plutonium production reactor. As at Oak Ridge, the need for labor turned Hanford into an atomic boomtown, with the population reaching 50,000 by summer 1944. Similar to the X-10 Graphite Reactor at Oak Ridge in terms of loading and unloading fuel, the B Reactor was built on a much larger scale and used water rather than air as a coolant. Whereas the X-10 had an initial design output of 1,000 kilowatts, the B Reactor was designed to operate at 250,000 kilowatts. Consisting of a 28- by 36-foot, 1,200-ton graphite cylinder lying on its side, the reactor was penetrated through its

188

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

Science Journals Connector (OSTI)

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

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

2015-01-01T23:59:59.000Z

189

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges The most life-limiting structural component in light-water reactors (LWR) is the reactor pressure vessel (RPV) because replacement of the RPV is not considered a viable option at this time. LWR licenses are now being extended from 40y to 60y by the U.S. Nuclear Regulatory Commission (NRC) with intentions to extend licenses to 80y and beyond. The RPV materials exhibit varying degrees of sensitivity to irradiation-induced embrittlement (decreased toughness) , as shown in Fig. 1.1, and extending operation from

190

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations, which govern the operation of commercial nuclear power plants, require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including

191

Reactor for exothermic reactions  

DOE Patents (OSTI)

A liquid phase process is described for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

Smith, L.A. Jr.; Hearn, D.; Jones, E.M. Jr.

1993-03-02T23:59:59.000Z

192

Thermionic Reactor Design Studies  

SciTech Connect

Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their (thermionic reactor) performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling.

Schock, Alfred

1994-08-01T23:59:59.000Z

193

System modeling and reactor design studies of the Advanced Thermionic Initiative space nuclear reactor  

SciTech Connect

In-core thermionic space reactor design concepts that operate at a nominal power output range of 20 to 50 kW(electric) are described. Details of the neutronic, thermionic, thermal hydraulics, and shielding performance are presented. Because of the strong absorption of thermal neutrons by natural tungsten and the large amount of natural tungsten within the reactor core, two designs are considered. An overall system design code has been developed at Oregon State University to model advanced in-core thermionic energy conversion-based nuclear reactor systems for space applications. The results show that the driverless single-cell Advanced Thermionic Initiative (ATI) configuration, which does not have driver fuel rods, proved to be more efficient than the driven core, which has driver rods. The results also show that the inclusion of the true axial and radial power distribution decrease the overall conversion efficiency. The flattening of the radial power distribution by three different methods would lead to a higher efficiency. The results show that only one TFE works at the optimum emitter temperature; all other TFEs are off the optimum performance and result in a 40% decrease of the efficiency of the overall system. The true axial profile is significantly different as there is a considerable amount of neutron leakage out of the top and bottom of the reactor. The analysis reveals that the axial power profile actually has a chopped cosine shape. For this axial profile, the reactor core overall efficiency for the driverless ATI reactor version is found to be 5.84% with a total electrical power of 21.92 kW(electric). By considering the true axial power profile instead of the uniform power profile, each TFE loses {approximately}80 W(electric).

Lee, H.H.; Abdul-Hamid, S.; Klein, A.C. [Oregon State Univ., Corvallis, OR (United States). Dept. of Nuclear Engineering Radiation Center] [Oregon State Univ., Corvallis, OR (United States). Dept. of Nuclear Engineering Radiation Center

1996-07-01T23:59:59.000Z

194

Generic effluent monitoring system certification for salt well portable exhauster  

SciTech Connect

Tests were conducted to verify that the Generic Effluent Monitoring System (GEMS), as it is applied to the Salt Well Portable Exhauster, meets all applicable regulatory performance criteria for air sampling systems at nuclear facilities. These performance criteria address both the suitability of the air sampling probe location and the transport of the sample to the collection devices. The criteria covering air sampling probe location ensure that the contaminants in the stack are well mixed with the airflow at the probe location such that the extracted sample represents the whole. The sample transport criteria ensure that the sampled contaminants are quantitatively delivered to the collection device. The specific performance criteria are described in detail in the report. The tests demonstrated that the GEMS/Salt Well Exhauster system meets all applicable performance criteria. Pacific Northwest National Laboratory conducted the testing using a mockup of the Salt Well Portable Exhauster stack at the Numatec Hanford Company`s 305 Building. The stack/sampling system configuration tested was designed to provide airborne effluent control for the Salt Well pumping operation at some U.S. Department of Energy (DOE) radioactive waste storage tanks at the Hanford Site, Washington. The portable design of the exhauster allows it to be used in other applications and over a range of exhaust air flowrates (approximately 200 - 1100 cubic feet per minute). The unit includes a stack section containing the sampling probe and another stack section containing the airflow, temperature and humidity sensors. The GEMS design features a probe with a single shrouded sampling nozzle, a sample delivery line, and sample collection system. The collection system includes a filter holder to collect the sample of record and an in-line detector head and filter for monitoring beta radiation-emitting particles.

Glissmeyer, J.A.; Maughan, A.D.

1997-09-01T23:59:59.000Z

195

Generic effluent monitoring system certification for AP-40 exhauster stack  

SciTech Connect

Tests were conducted to verify that the Generic Effluent Monitoring System (GEMS), as applied to the AP-40 exhauster stack, meets all applicable regulatory performance criteria for air sampling systems at nuclear facilities. These performance criteria address both the suitability of the air sampling probe location and the transport of the sample to the collection devices. The criteria covering air sampling probe location ensure that the contaminants in the stack are well mixed with the airflow at the probe location such that the extracted sample represents the whole. The sample transport criteria ensure that the sampled contaminants are quantitatively delivered to the collection device. The specific performance criteria are described in detail in the report. The tests demonstrated that the GEMS/AP-40 system meets all applicable performance criteria. The contaminant mixing tests were conducted by Pacific Northwest National Laboratory (PNNL) at the wind tunnel facility, 331-H Building, using a mockup of the actual stack. The particle sample transport tests were conducted by PNNL at the Numatec Hanford Company`s 305 Building. The AP-40 stack is typical of several 10-in. diameter stacks that discharge the filtered ventilation air from tank farms at the U.S. Department of Energy`s Hanford Site in Richland, Washington. The GEMS design features a probe with a single shrouded sampling nozzle, a sample delivery line, and sample collection system. The collection system includes a filter holder to collect the sample of record and an in-line detector head and filter for monitoring beta radiation-emitting particles. Unrelated to the performance criteria, it was found that the record sample filter holder exhibited symptoms of sample bypass around the particle collection filter. This filter holder should either be modified or replaced with a different type. 10 refs., 8 figs., 6 tabs.

Glissmeyer, J.A.; Davis, W.E.; Bussell, J.H.; Maughan, A.D.

1997-09-01T23:59:59.000Z

196

CASL milestone validates reactor model using TVA data | ornl.gov  

NLE Websites -- All DOE Office Websites (Extended Search)

Ron Walli Ron Walli Communications 865.576.0226 CASL milestone validates reactor model using TVA data This CASL visualization shows the thermal distribution of neutrons in Watts Bar Unit 1 Cycle 1 reactor core at initial criticality, as calculated by the VERA program. Image courtesy of Oak Ridge National Laboratory. This CASL visualization shows the thermal distribution of neutrons in Watts Bar Unit 1 Cycle 1 reactor core at initial criticality, as calculated by the VERA program. Image courtesy of Oak Ridge National Laboratory. (hi-res image) OAK RIDGE, Tenn., July 10, 2013 -Today, the Consortium for Advanced Simulation of Light Water Reactors announced that its scientists have successfully completed their first full-scale simulation of an operating nuclear reactor. CASL is modeling nuclear reactors on supercomputers to

197

Diagnostics for hybrid reactors  

SciTech Connect

The Hybrid Reactor(HR) can be considered an attractive actinide-burner or a fusion assisted transmutation for destruction of transuranic(TRU) nuclear waste. The hybrid reactor has two important subsystems: the tokamak neutron source and the blanket which includes a fuel zone where the TRU are placed and a tritium breeding zone. The diagnostic system for a HR must be as simple and robust as possible to monitor and control the plasma scenario, guarantee the protection of the machine and monitor the transmutation.

Orsitto, Francesco Paolo [ENEA Unita' Tecnica Fusione , Associazione ENEA-EURATOM sulla Fusione C R Frascati v E Fermi 45 00044 Frascati (Italy)

2012-06-19T23:59:59.000Z

198

Structural materials for fusion reactors  

Science Journals Connector (OSTI)

Fusion Reactors will require specially engineered structural materials, which ... on safety considerations. The fundamental differences between fusion and other nuclear reactors arise due to the 14MeV neutronics ...

P. M. Raole; S. P. Deshpande

2009-04-01T23:59:59.000Z

199

Transition to longitudinal instability of detonation waves is generically associated with Hopf bifurcation to  

E-Print Network (OSTI)

Transition to longitudinal instability of detonation waves is generically associated with Hopf We show that transition to longitudinal instability of strong detonation solu- tions of reactive detonations . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1.5 Structure of the equations

Texier, Benjamin - Institut de Mathématiques de Jussieu, Université Paris 7

200

Generic Library Extension in a Heterogeneous Environment Cosmin Oancea Stephen M. Watt  

E-Print Network (OSTI)

Generic Library Extension in a Heterogeneous Environment Cosmin Oancea Stephen M. Watt Department of Computer Science The University of Western Ontario London Ontario, Canada N6A 5B7 {coancea,watt

Watt, Stephen M.

Note: This page contains sample records for the topic "reactor generic distributed" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
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201

Performance Analysis of Generics in Scientific Computing Laurentiu Dragan Stephen M. Watt  

E-Print Network (OSTI)

Performance Analysis of Generics in Scientific Computing Laurentiu Dragan Stephen M. Watt Ontario Research Centre for Computer Algebra University of Western Ontario London, Ontario, Canada N6A 5B7 {ldragan,watt

Watt, Stephen M.

202

A Technique for Generic Iteration and Its Optimization Stephen M. Watt  

E-Print Network (OSTI)

A Technique for Generic Iteration and Its Optimization Stephen M. Watt Department of Computer Science University of Western Ontario London Ontario, Canada N6A 5B7 watt@csd.uwo.ca Abstract Software

Watt, Stephen M.

203

Generic Approach for Dispersing Single-Walled Carbon Nanotubes:? The Strength of a Weak Interaction  

Science Journals Connector (OSTI)

A generic noncovalent approach for dispersing high concentrations of individual single-walled carbon nanotubes (SWNT) in organic as well as aqueous solutions of synthetic block copolymers is presented. It is suggested that a weak, long-ranged entropic ...

Rina Shvartzman-Cohen; Yael Levi-Kalisman; Einat Nativ-Roth; Rachel Yerushalmi-Rozen

2004-06-24T23:59:59.000Z

204

Automation of the collection and processing of X-ray diffraction data - a generic approach  

Science Journals Connector (OSTI)

Fully automated data collection and processing would greatly enhance the efficiency of third-generation synchrotron beamlines. A scheme for achieving this goal in a cost-effective and generic way is outlined.

Leslie, A.G.W.

2002-10-21T23:59:59.000Z

205

Reactor Materials | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Benefits Crosscutting Technology Development Reactor Materials Advanced Sensors and Instrumentation Proliferation and Terrorism Risk Assessment Advanced Methods for Manufacturing...

206

Reactor operation safety information document  

SciTech Connect

The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

Not Available

1990-01-01T23:59:59.000Z

207

Performance Assessment Modeling and Sensitivity Analyses of Generic Disposal System Concepts.  

SciTech Connect

directly, rather than through simplified abstractions. It also a llows for complex representations of the source term, e.g., the explicit representation of many individual waste packages (i.e., meter - scale detail of an entire waste emplacement drift). This report fulfills the Generic Disposal System Analysis Work Packa ge Level 3 Milestone - Performance Assessment Modeling and Sensitivity Analyses of Generic Disposal System Concepts (M 3 FT - 1 4 SN08080 3 2 ).

Sevougian, S. David; Freeze, Geoffrey A. [Sandia National Laboratories, Albuquerque, NM; Gardner, William Payton [Sandia National Laboratories, Albuquerque, NM; Hammond, Glenn Edward [Sandia National Laboratories, Albuquerque, NM; Mariner, Paul [Sandia National Laboratories, Albuquerque, NM

2014-09-01T23:59:59.000Z

208

Fossil fuel furnace reactor  

DOE Patents (OSTI)

A fossil fuel furnace reactor is provided for simulating a continuous processing plant with a batch reactor. An internal reaction vessel contains a batch of shale oil, with the vessel having a relatively thin wall thickness for a heat transfer rate effective to simulate a process temperature history in the selected continuous processing plant. A heater jacket is disposed about the reactor vessel and defines a number of independent controllable temperature zones axially spaced along the reaction vessel. Each temperature zone can be energized to simulate a time-temperature history of process material through the continuous plant. A pressure vessel contains both the heater jacket and the reaction vessel at an operating pressure functionally selected to simulate the continuous processing plant. The process yield from the oil shale may be used as feedback information to software simulating operation of the continuous plant to provide operating parameters, i.e., temperature profiles, ambient atmosphere, operating pressure, material feed rates, etc., for simulation in the batch reactor.

Parkinson, William J. (Los Alamos, NM)

1987-01-01T23:59:59.000Z

209

Thermal Reactor Safety  

SciTech Connect

Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

Not Available

1980-06-01T23:59:59.000Z

210

NETL - Chemical Looping Reactor  

ScienceCinema (OSTI)

NETL's Chemical Looping Reactor unit is a high-temperature integrated CLC process with extensive instrumentation to improve computational simulations. A non-reacting test unit is also used to study solids flow at ambient temperature. The CLR unit circulates approximately 1,000 pounds per hour at temperatures around 1,800 degrees Fahrenheit.

None

2014-06-26T23:59:59.000Z

211

Baseline Concept Description of a Small Modular High Temperature Reactor  

SciTech Connect

The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the Generation IV program and its specific R&D needs will be included in this report when appropriate for comparison. The distinguishing features of the HTGR are the refractory (TRISO) coated particle fuel, the low-power density, graphite-moderated core, and the high outlet temperature of the inert helium coolant. The low power density and fuel form effectively eliminate the possibility of core melt, even upon a complete loss of coolant pressure and flow. The graphite, which constitutes the bulk of the core volume and mass, provides a large thermal buffer that absorbs fission heat such that thermal transients occur over a timespan of hours or even days. As chemically-inert helium is already a gas, there is no coolant temperature or void feedback on the neutronics and no phase change or corrosion product that could degrade heat transfer. Furthermore, the particle coatings and interstitial graphite retain fission products such that the source terms at the plant boundary remain well below actionable levels under all anticipated nominal and off-normal operating conditions. These attributes enable the reactor to supply process heat to a collocated industrial plant with negligible risk of contamination and minimal dynamic coupling of the facilities (Figure 1). The exceptional retentive properties of coated particle fuel in a graphite matrix were first demonstrated in the DRAGON reactor, a European research facility that began operation in 1964.

Hans Gougar

2014-05-01T23:59:59.000Z

212

Method of producing gaseous products using a downflow reactor  

DOE Patents (OSTI)

Reactor systems and methods are provided for the catalytic conversion of liquid feedstocks to synthesis gases and other noncondensable gaseous products. The reactor systems include a heat exchange reactor configured to allow the liquid feedstock and gas product to flow concurrently in a downflow direction. The reactor systems and methods are particularly useful for producing hydrogen and light hydrocarbons from biomass-derived oxygenated hydrocarbons using aqueous phase reforming. The generated gases may find used as a fuel source for energy generation via PEM fuel cells, solid-oxide fuel cells, internal combustion engines, or gas turbine gensets, or used in other chemical processes to produce additional products. The gaseous products may also be collected for later use or distribution.

Cortright, Randy D; Rozmiarek, Robert T; Hornemann, Charles C

2014-09-16T23:59:59.000Z

213

Neutronic analysis of a fusion hybrid reactor  

SciTech Connect

In a PHYSOR 2010 paper(1) we introduced a fusion hybrid reactor whose fusion component is the gasdynamic mirror (GDM), and whose blanket was made of thorium - 232. The thrust of that study was to demonstrate the performance of such a reactor by establishing the breeding of uranium - 233 in the blanket, and the burning thereof to produce power. In that analysis, we utilized the diffusion equation for one-energy neutron group, namely, those produced by the fusion reactions, to establish the power distribution and density in the system. Those results should be viewed as a first approximation since the high energy neutrons are not effective in inducing fission, but contribute primarily to the production of actinides. In the presence of a coolant, however, such as water, these neutrons tend to thermalize rather quickly, hence a better assessment of the reactor performance would require at least a two group analysis, namely the fast and thermal groups. We follow that approach and write an approximate set of equations for the fluxes of these groups. From these relations we deduce the all-important quantity, k{sub eff}, which we utilize to compute the multiplication factor, and subsequently, the power density in the reactor. We show that k{sub eff} can be made to have a value of 0.99, thus indicating that 100 thermal neutrons are generated per fusion neutron, while allowing the system to function as 'subcritical.' Moreover, we show that such a hybrid reactor can generate hundreds of megawatts of thermal power per cm of length depending on the flux of the fusion neutrons impinging on the blanket. (authors)

Kammash, T. [Univ. of Michigan, NERS, 2355 Bonisteel Blvd., Ann Arbor, MI 48109 (United States)

2012-07-01T23:59:59.000Z

214

Reactor vessel support system. [LMFBR  

DOE Patents (OSTI)

A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

Golden, M.P.; Holley, J.C.

1980-05-09T23:59:59.000Z

215

Fusion reactor systems  

Science Journals Connector (OSTI)

In this review we consider deuterium-tritium (D-T) fusion reactors based on four different plasma-confinement and heating approaches: the tokamak, the theta-pinch, the magnetic-mirror, and the laser-pellet system. We begin with a discussion of the dynamics of reacting plasmas and basic considerations of reactor power balance. The essential plasma physical aspects of each system are summarized, and the main characteristics of the corresponding conceptual power plants are described. In tokamak reactors the plasma densities are about 1020 m-3, and the ? values (ratio of plasma pressure to confining magnetic pressure) are approximately 5%. Plasma burning times are of the order of 100-1000 sec. Large superconducting dc magnets furnish the toroidal magnetic field, and 2-m thick blankets and shields prevent heat deposition in the superconductor. Radially diffusing plasma is diverted away from the first wall by means of null singularities in the poloidal (or transverse) component of the confining magnetic field. The toroidal theta-pinch reactor has a much smaller minor diameter and a much larger major diameter, and operates on a 10-sec cycle with 0.1-sec burning pulses. It utilizes shock heating from high-voltage sources and adabatic-compression heating powered by low-voltage, pulsed cryogenic magnetic or inertial energy stores, outside the reactor core. The plasma has a density of about 1022 m-3 and ? values of nearly unity. In the power balance of the reactor, direct-conversion energy obtained by expansion of the burning high-? plasma against the containing magnetic field is an important factor. No divertor is necessary since neutral-gas flow cools and replaces the "spent" plasma between pulses. The open-ended mirror reactor uses both thermal conversion of neutron energy and direct conversion of end-loss plasma energy to dc electrical power. A fraction of this direct-convertor power is then fed back to the ioninjection system to sustain the reaction and maintain the plasma. The average ion energy is 600 keV, plasma diameter 6 m, and the plasma beta 85%. The power levels of the three magnetic-confinement devices are in the 500-2000 MWe range, with the exception of the mirror reactor, for which the output is approximately 200 MWe. In Laser-Pellet reactors, frozen D-T pellets are ignited in a cavity which absorbs the electromagnetic, charged particle, and neutron energy from the fusion reaction. The confinement is "inertial," since the fusion reaction occurs during the disassembly of the heated pellet. A pellet-cavity unit would produce about 200 MWt in pulses with a repetition rate of the order of 10 sec-1. Such units could be clustered to give power plants with outputs in the range of 1000 MWe.

F. L. Ribe

1975-01-01T23:59:59.000Z

216

Nuclear reactor construction with bottom supported reactor vessel  

DOE Patents (OSTI)

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core within the pool. The reactor vessel has an open top end, a closed flat bottom end wall and a continuous cylindrical closed side wall interconnecting the top end and bottom end wall. The reactor also has a generally cylindrical concrete containment structure surrounding the reactor vessel and being formed by a cylindrical side wall spaced outwardly from the reactor vessel side wall and a flat base mat spaced below the reactor vessel bottom end wall. A central support pedestal is anchored to the containment structure base mat and extends upwardly therefrom to the reactor vessel and upwardly therefrom to the reactor core so as to support the bottom end wall of the reactor vessel and the lower end of the reactor core in spaced apart relationship above the containment structure base mat. Also, an annular reinforced support structure is disposed in the reactor vessel on the bottom end wall thereof and extends about the lower end of the core so as to support the periphery thereof. In addition, an annular support ring having a plurality of inward radially extending linear members is disposed between the containment structure base mat and the bottom end of the reactor vessel wall and is connected to and supports the reactor vessel at its bottom end on the containment structure base mat so as to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event. The reactor construction also includes a bed of insulating material in sand-like granular form, preferably being high density magnesium oxide particles, disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall on the containment structure base mat so as to insulate the reactor vessel bottom end wall from the containment structure base mat and allow the reactor vessel bottom end wall to freely expand as it heats up while providing continuous support thereof. Further, a deck is supported upon the side wall of the containment structure above the top open end of the reactor vessel, and a plurality of serially connected extendible and retractable annular bellows extend between the deck and the top open end of the reactor vessel and flexibly and sealably interconnect the reactor vessel at its top end to the deck. An annular guide ring is disposed on the containment structure and extends between its side wall and the top open end of the reactor vessel for providing lateral support of the reactor vessel top open end by limiting imposition of lateral loads on the annular bellows by the occurrence of a lateral seismic event.

Sharbaugh, John E. (Bullskin Township, Fayette County, PA)

1987-01-01T23:59:59.000Z

217

Nuclear plant-aging research on reactor protection systems  

SciTech Connect

This report presents the rsults of a review of the Reactor Trip System (RTS) and the Engineered Safety Feature Actuating System (ESFAS) operating experiences reported in Licensee Event Reports (LER)s, the Nuclear Power Experience data base, Nuclear Plant Reliability Data System, and plant maintenance records. Our purpose is to evaluate the potential significance of aging, including cycling, trips, and testing as contributors to degradation of the RTS and ESFAS. Tables are presented that show the percentage of events for RTS and ESFAS classified by cause, components, and subcomponents for each of the Nuclear Steam Supply System vendors. A representative Babcock and Wilcox plant was selected for detailed study. The US Nuclear Regulatory Commission's Nuclear Plant Aging Research guidelines were followed in performing the detailed study that identified materials susceptible to aging, stressors, environmental factors, and failure modes for the RTS and ESFAS as generic instrumentation and control systems. Functional indicators of degradation are listed, testing requirements evaluated, and regulatory issues discussed.

Meyer, L.C.

1988-01-01T23:59:59.000Z

218

OXIDATIVE COUPLING OF METHANE USING INORGANIC MEMBRANE REACTORS  

SciTech Connect

The objective of this research is to study the oxidative coupling of methane in catalytic inorganic membrane reactors. A specific target is to achieve conversion of methane to C{sub 2} hydrocarbons at very high selectivity and higher yields than in conventional non-porous, co-feed, fixed bed reactors by controlling the oxygen supply through the membrane. A membrane reactor has the advantage of precisely controlling the rate of delivery of oxygen to the catalyst. This facility permits balancing the rate of oxidation and reduction of the catalyst. In addition, membrane reactors minimize the concentration of gas phase oxygen thus reducing non selective gas phase reactions, which are believed to be a main route for the formation of CO{sub x} products. Such gas phase reactions are a cause of decreased selectivity in the oxidative coupling of methane in conventional flow reactors. Membrane reactors could also produce higher product yields by providing better distribution of the reactant gases over the catalyst than the conventional plug flow reactors. Membrane reactor technology also offers the potential for modifying the membranes both to improve catalytic properties as well as to regulate the rate of the permeation/diffusion of reactants through the membrane to minimize by-product generation. Other benefits also exist with membrane reactors, such as the mitigation of thermal hot-spots for highly exothermic reactions such as the oxidative coupling of methane. The application of catalytically active inorganic membranes has potential for drastically increasing the yield of reactions which are currently limited by either thermodynamic equilibria, product inhibition, or kinetic selectivity.

Dr. Y.H. Ma; Dr. W.R. Moser; Dr. A.G. Dixon; Dr. A.M. Ramachandra; Dr. Y. Lu; C. Binkerd

1998-04-01T23:59:59.000Z

219

Generic Repository Concepts and Thermal Analysis for Advanced Fuel Cycles  

SciTech Connect

The current posture of the used nuclear fuel management program in the U.S. following termination of the Yucca Mountain Project, is to pursue research and development (R&D) of generic (i.e., non-site specific) technologies for storage, transportation and disposal. Disposal R&D is directed toward understanding and demonstrating the performance of reference geologic disposal concepts selected to represent the current state-of-the-art in geologic disposal. One of the principal constraints on waste packaging and emplacement in a geologic repository is management of the waste-generated heat. This paper describes the selection of reference disposal concepts, and thermal management strategies for waste from advanced fuel cycles. A geologic disposal concept for spent nuclear fuel (SNF) or high-level waste (HLW) consists of three components: waste inventory, geologic setting, and concept of operations. A set of reference geologic disposal concepts has been developed by the U.S. Department of Energy (DOE) Used Fuel Disposition Campaign, for crystalline rock, clay/shale, bedded salt, and deep borehole (crystalline basement) geologic settings. We performed thermal analysis of these concepts using waste inventory cases representing a range of advanced fuel cycles. Concepts of operation consisting of emplacement mode, repository layout, and engineered barrier descriptions, were selected based on international progress and previous experience in the U.S. repository program. All of the disposal concepts selected for this study use enclosed emplacement modes, whereby waste packages are in direct contact with encapsulating engineered or natural materials. The encapsulating materials (typically clay-based or rock salt) have low intrinsic permeability and plastic rheology that closes voids so that low permeability is maintained. Uniformly low permeability also contributes to chemically reducing conditions common in soft clay, shale, and salt formations. Enclosed modes are associated with temperature constraints that limit changes to the encapsulating materials, and they generally have less capacity to dissipate heat from the waste package and its immediate surroundings than open modes such as that proposed for a repository at Yucca Mountain, Nevada. Open emplacement modes can be ventilated for many years prior to permanent closure of the repository, limiting peak temperatures both before and after closure, and combining storage and disposal functions in the same facility. Open emplacement modes may be practically limited to unsaturated host formations, unless emplacement tunnels are effectively sealed everywhere prior to repository closure. Thermal analysis of disposal concepts and waste inventory cases has identified important relationships between waste package size and capacity, and the duration of surface decay storage needed to meet temperature constraints. For example, the choice of salt as the host medium expedites the schedule for geologic disposal by approximately 50 yr (other factors held constant) thereby reducing future reliance on surface decay storage. Rock salt has greater thermal conductivity and stability at higher temperatures than other media considered. Alternatively, the choice of salt permits the use of significantly larger waste packages for SNF. The following sections describe the selection of reference waste inventories, geologic settings, and concepts of operation, and summarize the results from the thermal analysis.

Hardin, Ernest [Sandia National Laboratories (SNL)] [Sandia National Laboratories (SNL); Blink, James [Lawrence Livermore National Laboratory (LLNL)] [Lawrence Livermore National Laboratory (LLNL); Carter, Joe [Savannah River National Laboratory (SRNL)] [Savannah River National Laboratory (SRNL); Massimiliano, Fratoni [Lawrence Livermore National Laboratory (LLNL)] [Lawrence Livermore National Laboratory (LLNL); Greenberg, Harris [Lawrence Livermore National Laboratory (LLNL)] [Lawrence Livermore National Laboratory (LLNL); Howard, Rob L [ORNL] [ORNL

2011-01-01T23:59:59.000Z

220

Spherical torus fusion reactor  

DOE Patents (OSTI)

A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

Peng, Yueng-Kay M. (Oak Ridge, TN)

1989-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor generic distributed" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


221

Nuclear divisional reactor  

SciTech Connect

A nuclear divisional reactor including a reactor core having side and top walls, a heat exchanger substantially surrounding the core, the heat exchanger including a plurality of separate fluid holding and circulating chambers each in contact with a portion of the core, control rod means associated with the core and external of the heat exchanger including control rods and means for moving said control rods, each of the chambers having separate means for delivering and removing fluid therefrom, separate means associated with each of the delivering and removing means for producing useable energy external of the chambers, each of the means for producing useable energy having separate variable capacity energy outputs thereby making available a plurality of individual sources of useable energy of varying degrees.

Administratrix, A.P.; Rugh, J.L.

1982-11-02T23:59:59.000Z

222

Fusion reactor pumped laser  

DOE Patents (OSTI)

A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam. 10 figs.

Jassby, D.L.

1987-09-04T23:59:59.000Z

223

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

2010-09-01T23:59:59.000Z

224

Infinite Dimensional Port Hamiltonian Representation of Chemical Reactors  

E-Print Network (OSTI)

Infinite Dimensional Port Hamiltonian Representation of Chemical Reactors W. Zhou B. Hamroun Y.le.gorrec@ens2m.fr) Abstract: Infinite dimensional Port Hamiltonian representation of non isothermal chemical: Port Hamiltonian Systems, Distributed Systems, Irreversible Thermodynamics 1. INTRODUCTION Models

Paris-Sud XI, Université de

225

Summary of space nuclear reactor power systems, 1983--1992  

SciTech Connect

This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

Buden, D.

1993-08-11T23:59:59.000Z

226

Thermionic Reactor Design Studies  

SciTech Connect

During the 1960's and early 70's the author performed extensive design studies, analyses, and tests aimed at thermionic reactor concepts that differed significantly from those pursued by other investigators. Those studies, like most others under Atomic Energy Commission (AEC and DOE) and the National Aeronautics and Space Administration (NASA) sponsorship, were terminated in the early 1970's. Some of this work was previously published, but much of it was never made available in the open literature. U.S. interest in thermionic reactors resumed in the early 80's, and was greatly intensified by reports about Soviet ground and flight tests in the late 80's. This recent interest resulted in renewed U.S. thermionic reactor development programs, primarily under Department of Defense (DOD) and Department of Energy (DOE) sponsorship. Since most current investigators have not had an opportunity to study all of the author's previous work, a review of the highlights of that work may be of value to them. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling. Where the author's concepts differed from the later Topaz-2 design was in the relative location of the emitter and the collector. Placing the fueled emitter on the outside of the cylindrical diodes permits much higher axial conductances to reduce ohmic losses in the electrodes of full-core-height diodes. Moreover, placing the fuel on the outside of the diode makes possible reactors with much higher fuel volume fractions, which enable power-flattened fast reactors scalable to very low power levels without the need for life-limiting hydride moderators or the use of efficiency-limiting driver fuel. In addition, with the fuel on the outside its swelling does not increase the emitter diameter or reduce the interelectrode gap. This should permit long lifetimes even with closer spacings, which can significantly improve the system efficiences. This was confirmed by coupled neutronic, thermal, thermionic, and electrical system analyses - some of which are presented in this paper - and by subsequent experiments. A companion paper presented next describes the fabrication and testing of full-scale converter elements, both fueled and unfueled, and summarizes the test results obtained. There is a duplicate copy in the file.

Schock, Alfred

1994-06-01T23:59:59.000Z

227

Massive Hanford Test Reactor Removed - Plutonium Recycle Test...  

Office of Environmental Management (EM)

Massive Hanford Test Reactor Removed - Plutonium Recycle Test Reactor removed from Hanford's 300 Area Massive Hanford Test Reactor Removed - Plutonium Recycle Test Reactor removed...

228

Nuclear Reactor Severe Accident Experiments  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Reactor Severe Accident Experiments Nuclear Reactor Severe Accident Experiments Capabilities Engineering Experimentation Reactor Safety Testing and Analysis Overview Nuclear Reactor Severe Accident Experiments MAX NSTF SNAKE Aerosol Experiments System Components Laser Applications Robots Applications Other Facilities Other Capabilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Nuclear Reactor Severe Accident Experiments 1 2 3 4 5 6 7 We perform experiments simulating reactor core melt phenomena in which molten core debris ("corium") erodes the concrete floor of a containment building. This occurred during the Fukushima nuclear power plant accident though the extent of concrete damage is yet unknown. This video shows the top view of a churning molten pool of uranium oxide at 2000°C (3600°F) seen during an experiment at Argonne. Corium behaves much like lava.

229

Solid oxide electrochemical reactor science.  

SciTech Connect

Solid-oxide electrochemical cells are an exciting new technology. Development of solid-oxide cells (SOCs) has advanced considerable in recent years and continues to progress rapidly. This thesis studies several aspects of SOCs and contributes useful information to their continued development. This LDRD involved a collaboration between Sandia and the Colorado School of Mines (CSM) ins solid-oxide electrochemical reactors targeted at solid oxide electrolyzer cells (SOEC), which are the reverse of solid-oxide fuel cells (SOFC). SOECs complement Sandia's efforts in thermochemical production of alternative fuels. An SOEC technology would co-electrolyze carbon dioxide (CO{sub 2}) with steam at temperatures around 800 C to form synthesis gas (H{sub 2} and CO), which forms the building blocks for a petrochemical substitutes that can be used to power vehicles or in distributed energy platforms. The effort described here concentrates on research concerning catalytic chemistry, charge-transfer chemistry, and optimal cell-architecture. technical scope included computational modeling, materials development, and experimental evaluation. The project engaged the Colorado Fuel Cell Center at CSM through the support of a graduate student (Connor Moyer) at CSM and his advisors (Profs. Robert Kee and Neal Sullivan) in collaboration with Sandia.

Sullivan, Neal P. (Colorado School of Mines, Golden, CO); Stechel, Ellen Beth; Moyer, Connor J. (Colorado School of Mines, Golden, CO); Ambrosini, Andrea; Key, Robert J. (Colorado School of Mines, Golden, CO)

2010-09-01T23:59:59.000Z

230

Report on THMC Modeling of the Near Field Evolution of a Generic Clay  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

on THMC Modeling of the Near Field Evolution of a Generic on THMC Modeling of the Near Field Evolution of a Generic Clay Repository: Model Validation and Demonstration Rev 2 Report on THMC Modeling of the Near Field Evolution of a Generic Clay Repository: Model Validation and Demonstration Rev 2 Shale and clay-rich rock formations have been considered as potential host rocks for geological disposal of high-level radioactive waste throughout the world. Coupled thermal, hydrological, mechanical, and chemical (THMC) processes have a significant impact on the long-term safety of a repository in this type of rocks. The validity of the two-part Hooke's model (TPHM), a new constitutive relationship, and associated formulations regarding rock hydraulic/mechanical properties is demonstrated by the consistency between observations from a mine-by test at the Mont Terri site

231

Electricity Distribution  

Science Journals Connector (OSTI)

High voltage (HV) distribution grids have nominal voltages of up ... the grid that connects distribution to the transmission substations and also supplies large industrial customers requiri...

Tomás Gómez

2013-01-01T23:59:59.000Z

232

Nuclear Reactor Materials and Fuels  

Science Journals Connector (OSTI)

Nuclear reactor materials and fuels can be classified into six categories: Nuclear fuel materials Nuclear clad materials Nuclear coolant materials Nuclear poison materials Nuclear moderator materials

Dr. James S. Tulenko

2012-01-01T23:59:59.000Z

233

Thermonuclear Reflect AB-Reactor  

E-Print Network (OSTI)

The author offers a new kind of thermonuclear reflect reactor. The remarkable feature of this new reactor is a three net AB reflector, which confines the high temperature plasma. The plasma loses part of its energy when it contacts with the net but this loss can be compensated by an additional permanent plasma heating. When the plasma is rarefied (has a small density), the heat flow to the AB reflector is not large and the temperature in the triple reflector net is lower than 2000 - 3000 K. This offered AB-reactor has significantly less power then the currently contemplated power reactors with magnetic or inertial confinement (hundreds-thousands of kW, not millions of kW). But it is enough for many vehicles and ships and particularly valuable for tunnelers, subs and space apparatus, where air to burn chemical fuel is at a premium or simply not available. The author has made a number of innovations in this reactor, researched its theory, developed methods of computation, made a sample computation of typical project. The main point of preference for the offered reactor is its likely cheapness as a power source. Key words: Micro-thermonuclear reactor, Multi-reflex AB-thermonuclear reactor, Self-magnetic AB-thermonuclear reactor, aerospace thermonuclear engine.

Alexander Bolonkin

2008-03-26T23:59:59.000Z

234

Light Water Reactor Sustainability Newsletter  

NLE Websites -- All DOE Office Websites (Extended Search)

hydraulics software RELAP-7 (which is under development in the Light Water Reactor Sustainability LWRS Program). A novel interaction between the probabilistic part (i.e., RAVEN)...

235

Reactor coolant pump flywheel  

DOE Patents (OSTI)

A flywheel for a pump, and in particular a flywheel having a number of high density segments for use in a nuclear reactor coolant pump. The flywheel includes an inner member and an outer member. A number of high density segments are provided between the inner and outer members. The high density segments may be formed from a tungsten based alloy. A preselected gap is provided between each of the number of high density segments. The gap accommodates thermal expansion of each of the number of segments and resists the hoop stress effect/keystoning of the segments.

Finegan, John Raymond; Kreke, Francis Joseph; Casamassa, John Joseph

2013-11-26T23:59:59.000Z

236

Biparticle fluidized bed reactor  

DOE Patents (OSTI)

A fluidized bed reactor system utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves. 3 figs.

Scott, C.D.; Marasco, J.A.

1995-04-25T23:59:59.000Z

237

Biparticle fluidized bed reactor  

DOE Patents (OSTI)

A fluidized bed reactor system is described which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary and tertiary particulate phases, continuously introduced and removed simultaneously in the cocurrent and countercurrent mode, act in a role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Means for introducing and removing the sorbent phases include feed screw mechanisms and multivane slurry valves. 3 figs.

Scott, C.D.; Marasco, J.A.

1996-02-27T23:59:59.000Z

238

Nuclear reactor control apparatus  

DOE Patents (OSTI)

Nuclear reactor core safety rod release apparatus comprises a control rod having a detent notch in the form of an annular peripheral recess at its upper end, a control rod support tube for raising and lowering the control rod under normal conditions, latches pivotally mounted on the control support tube with free ends thereof normally disposed in the recess in the control rod, and cam means for pivoting the latches out of the recess in the control rod when a scram condition occurs. One embodiment of the invention comprises an additional magnetically-operated latch for releasing the control rod under two different conditions, one involving seismic shock.

Sridhar, Bettadapur N. (Cupertino, CA)

1983-11-01T23:59:59.000Z

239

Generic Wave-Function Description of Fractional Quantum Anomalous Hall States and Fractional Topological Insulators  

Science Journals Connector (OSTI)

We propose a systematical approach to construct generic fractional quantum anomalous Hall states, which are generalizations of the fractional quantum Hall states to lattice models with zero net magnetic field and full lattice translation symmetry. Local and translationally invariant Hamiltonians can also be constructed, for which the proposed states are unique ground states. Our result demonstrates that generic chiral topologically ordered states can be realized in lattice models, without requiring magnetic translation symmetry and Landau level structure. We further generalize our approach to fractional topological insulators, and provide the first explicit wave-function description of fractional topological insulators in the absence of spin conservation.

Xiao-Liang Qi

2011-09-15T23:59:59.000Z

240

A generic approach to development of integrated environments for engineering design  

E-Print Network (OSTI)

enables older models to remain in the system for future use, if desired. Summary of Intent The intent of this Section was to detail the generic architecture proposed by this thesis. Each of the objects of the object oriented design was described... enables older models to remain in the system for future use, if desired. Summary of Intent The intent of this Section was to detail the generic architecture proposed by this thesis. Each of the objects of the object oriented design was described...

Hart, Paul Bradley

2012-06-07T23:59:59.000Z

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While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
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to obtain the most current and comprehensive results.


241

Modular Pebble Bed Reactor High Temperature Gas Reactor  

E-Print Network (OSTI)

Modular Pebble Bed Reactor High Temperature Gas Reactor Andrew C Kadak Massachusetts Institute For 1150 MW Combined Heat and Power Station Oil Refinery Hydrogen Production Desalinization Plant VHTR/Graphite Discrimination system Damaged Sphere ContainerGraphiteReturn FuelReturn Fresh Fuel Container Spent Fuel Tank #12

242

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Assessment of High Value Surveillance Materials Assessment of High Value Surveillance Materials Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Assessment of High Value Surveillance Materials The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the operation of commercial nuclear power plants require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely

243

Axial xenon stability considerations in VVER-1000 reactors  

SciTech Connect

Frequent problems experienced in VVER-1000 reactors with xenon oscillation control have been reported. Modern Pressurized Water Reactors (PWRs) are designed with operational strategies to help avoid such axial xenon oscillations. Uncontrolled xenon oscillations can cause operational problems, requiring frequent operator intervention and leading to power reductions (or plant trips) due to high core peaking factors, thereby reducing overall plant capacity factors. In the worst case, an uncontrolled xenon oscillation can lead to violations of safety requirements on pellet clad interaction (PCI), departure from nucleate boiling or fuel centerline melting if the plant does not have adequate safety protection systems that account for limiting axial power distributions.

Doshi, P.K.; Miller, R.W.

1993-12-31T23:59:59.000Z

244

DOE Drops Plan to Restart Reactor  

Science Journals Connector (OSTI)

...longer in flux. Hanford research reactor...decision to scrap the Hanford reactor, which...research. At public meetings, however...decision to scrap the Hanford reactor, which...research. At public meetings, however, FFTF...

Robert F. Service

2000-12-01T23:59:59.000Z

245

Operational Analysis of Multiregional Nuclear Reactor Kinetics  

Science Journals Connector (OSTI)

......Operational Analysis of Multiregional Nuclear Reactor Kinetics NASSAR H. S. HAIDAR...analytically for a multiregional nuclear reactor whose subregions are of arbitrary...Operational Analysis of Multiregional Nuclear Reactor Kinetics NASSAU H. S. HAIDAR......

NASSAR H. S. HAIDAR

1983-05-01T23:59:59.000Z

246

Solvent refined coal reactor quench system  

DOE Patents (OSTI)

There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream. 1 fig.

Thorogood, R.M.

1983-11-08T23:59:59.000Z

247

Iterative methods for solving nonlinear problems of nuclear reactor criticality  

SciTech Connect

The paper presents iterative methods for calculating the neutron flux distribution in nonlinear problems of nuclear reactor criticality. Algorithms for solving equations for variations in the neutron flux are considered. Convergence of the iterative processes is studied for two nonlinear problems in which macroscopic interaction cross sections are functionals of the spatial neutron distribution. In the first problem, the neutron flux distribution depends on the water coolant density, and in the second one, it depends on the fuel temperature. Simple relationships connecting the vapor content and the temperature with the neutron flux are used.

Kuz'min, A. M., E-mail: mephi.kam@mail.ru [National Research Nuclear University MEPhI (Russian Federation)

2012-12-15T23:59:59.000Z

248

Temperature effects on chemical reactor  

Science Journals Connector (OSTI)

In this paper we had to study some characteristics of the chemical reactors from which we can understand the reactor operation in different circumstances; from these and the most important factor that has a great effect on the reactor operation is the temperature it is a mathematical processing of a chemical problem that was already studied but it may be developed by introducing new strategies of control; in our case we deal with the analysis of a liquid?gas reactor which can make the flotation of the benzene to produce the ethylene; this type of reactors can be used in vast domains of the chemical industry especially in refinery plants where we find the oil separation and its extractions whether they are gases or liquids which become necessary for industrial technology especially in our century.

M. Azzouzi

2008-01-01T23:59:59.000Z

249

Frontiers in Reactor Engineering  

Science Journals Connector (OSTI)

...multicomponent, multiphase systems (7, 8). On the...Simulations in multiphase systems are vastly aided by imaging tools for the assessment of phase distributions...g., methane steam reforming), leading...for a number of systems that use UOP...

Milorad P. Dudukovic

2009-08-07T23:59:59.000Z

250

THE MATERIALS OF FAST BREEDER REACTORS  

E-Print Network (OSTI)

metal fast breeder reactor (LMFBR) concern the behavior ofmetal fast breeder reactor (LMFBR). Despite the simplicityinduced by irradiation. LMFBR funding is the largest single

Olander, Donald R.

2013-01-01T23:59:59.000Z

251

Nuclear reactors in the United States  

Science Journals Connector (OSTI)

Nuclear reactors in the United States ... A chart listing the operating and planned nuclear reactors in the United States. ... Nuclear / Radiochemistry ...

Hubert N. Alyea

1956-01-01T23:59:59.000Z

252

Advanced Reactor Research and Development Funding Opportunity...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Reactor Research and Development Funding Opportunity Announcement Advanced Reactor Research and Development Funding Opportunity Announcement The U.S. Department of Energy (DOE)...

253

LIGHT WATER REACTOR SUSTAINABILITY PROGRAM: INTRODUCTION  

NLE Websites -- All DOE Office Websites (Extended Search)

LIGHT WATER REACTOR SUSTAINABILITY PROGRAM: INTRODUCTION The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1...

254

MOOSE simulating nuclear reactor CRUD buildup  

SciTech Connect

This simulation uses multiple physical models to show how the buildup of boron deposits on reactor fuel can affect performance and the reactor's power profile.

None

2014-02-06T23:59:59.000Z

255

Advanced Reactor Technologies | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Advanced Reactor Advanced Reactor Technologies Advanced Reactor Technologies Advanced Reactor Technologies Advanced Reactor Technologies The Office of Advanced Reactor Technologies (ART) sponsors research, development and deployment (RD&D) activities through its Next Generation Nuclear Plant (NGNP), Advanced Reactor Concepts (ARC), and Advanced Small Modular Reactor (aSMR) programs to promote safety, technical, economical, and environmental advancements of innovative Generation IV nuclear energy technologies. The Office of Nuclear Energy (NE) will pursue these advancements through RD&D activities at the Department of Energy (DOE) national laboratories and U.S. universities, as well as through collaboration with industry and international partners. These activities will focus on advancing scientific

256

Granular Dynamics in Pebble Bed Reactor Cores  

E-Print Network (OSTI)

pebble bed reactor,” Nuclear Engineering and Design, vol.the AVR reactor,” Nuclear Engineering and Design, vol. 121,Operating Experience,” Nuclear Engineering and Design, vol.

Laufer, Michael Robert

2013-01-01T23:59:59.000Z

257

F Reactor Inspection | Department of Energy  

NLE Websites -- All DOE Office Websites (Extended Search)

Inspection F Reactor Inspection Addthis Description Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor last week before...

258

Physics of nuclear reactor safety  

Science Journals Connector (OSTI)

Provides a concise review of the physical aspects of safety of nuclear fission reactors. It covers the developments of roughly the last decade. The introductory chapter contains an analysis of the changes in safety philosophy that are characteristic of the last decade and that have given rise to an increased importance of physical aspects because of the emphasis on passive or natural safety. The second chapter focuses on the basics of reactor safety, identifying the main risk sources and the main principles for a safe design. The third chapter concerns a systematic treatment of the physical processes that are fundamental for the properties of fission chain reacting processes and the control of those processes. Because of the rather specialized nature of the field of reactor physics, each paragraph contains a very concise description of the theory of the phenomenon under consideration, before presenting a review of the developments. Chapter 4 contains a short review of the thermal aspects of reactor safety, restricted to those aspects that are characteristic of the nuclear reactor field, because thermal hydraulics of fission reactors is not principally different from that of other physical systems. In chapter 5 the consequences of the physics treated in the preceding chapters for the dynamics and safety of actual reactors are reviewed. The systematics of the treatment is mainly based on a division of reactors into three categories according to the type of coolant, which to a large extent determines the safety properties of the reactors. The last chapter contains a physical analysis of the Chernobyl accident that occurred in 1986. The reason for an attempt to give a review of this accident, as complete as possible within the space limits set by the editors, is twofold: the Chernobyl accident is the most severe accident in history and physical properties of the reactor played a decisive role, thereby serving as an illustration of the material of the preceding chapters.

H van Dam

1992-01-01T23:59:59.000Z

259

KNOWLEDGE MANAGEMENT FOR INDUSTRIAL SAFETY, GENERIC RESOURCE PLATFORM COMBINED WITH AN ONTOLOGY BASED APPROACH  

E-Print Network (OSTI)

safety culture to enable effective use of available knowledge for the prevention of major accidentsKNOWLEDGE MANAGEMENT FOR INDUSTRIAL SAFETY, GENERIC RESOURCE PLATFORM COMBINED WITH AN ONTOLOGY to manage risks and maintain industrial safety is largely based on the capacity of various actors to acquire

Boyer, Edmond

260

Embedding Ergonomic Rules as Generic Requirements in a Formal Development Process of Interactive Software  

E-Print Network (OSTI)

Embedding Ergonomic Rules as Generic Requirements in a Formal Development Process of Interactive a formal framework for the development of interactive software that bridges the gap between ergonomic development process from requirements to model­based execution. It also embeds ergonomic knowledge

Farenc, Christelle

Note: This page contains sample records for the topic "reactor generic distributed" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


261

A Generic Framework based on Ergonomics Rules for Computer Aided Design of User Interface  

E-Print Network (OSTI)

A Generic Framework based on Ergonomics Rules for Computer Aided Design of User Interface Cedex, France E­mail:{farenc, palanque}@univ­tlse1.fr Key words: Ergonomic rules, usability evaluation. Abstract: Ergonomic rules are supposed to help developers to build UI respecting human factor principles

Farenc, Christelle

262

2-3. Generic Approaches Towards Water Quality Monitoring Based on Paleolimnology  

E-Print Network (OSTI)

phosphorus analysis of Lake St-Charles, the principal drinking water supply for Québec City, #12;62 R environmental records for lake and river ecosystems provide a valuable generic tool for water quality management by way of water quality research on three ecosystems in Québec, Canada. Lake St-Augustin is a small lake

Vincent, Warwick F.

263

Transition to longitudinal instability of detonation waves is generically associated with Hopf bifurcation to  

E-Print Network (OSTI)

Transition to longitudinal instability of detonation waves is generically associated with Hopf show that transition to longitudinal instability of strong detonation solu- tions of reactive of an associated Evans function, and obtain the first complete nonlinear stability result for strong detonations

Hélein, Frédéric - Institut de Mathématiques de Jussieu, Université Paris 7

264

Automating the Transfer of a Generic Set of Behaviors Onto a Virtual Character  

E-Print Network (OSTI)

of specialists, despite the broad availability of the models, assets and simulation environments. To address this problem, we present a system that allows the rapid incorpo- ration of high-fidelity humanoid 3D modelsAutomating the Transfer of a Generic Set of Behaviors Onto a Virtual Character Andrew Feng1

Kallmann, Marcelo

265

Quality engineering process for the Program Design Phase of a generic software life cycle  

E-Print Network (OSTI)

Quality engineering process for the Program Design Phase of a generic software life cycle Witold.georgiadou@mdx.ac.uk Abstract This paper presents the design of a quality engineering process applicable in the program design place between the program designer and the software quality engineer. The paper also discusses

Suryn, Witold

266

Towards a Generic Context-Aware Framework for Self-Adaptation of Service-Oriented Architectures  

E-Print Network (OSTI)

Towards a Generic Context-Aware Framework for Self-Adaptation of Service-Oriented Architectures context-aware frame- work to build self-organizing service-oriented architectures for cloud computing Application (SBA) is a software architecture where the basic element is the service, performing a single

Paris-Sud XI, Université de

267

A Harmonic Potential Approach for Simultaneous Planning and Control of a Generic UAV Platform  

E-Print Network (OSTI)

A Harmonic Potential Approach for Simultaneous Planning and Control of a Generic UAV Platform Ahmad and control of a large variety of unmanned aerial vehicles (UAVs) is tackled using the harmonic potential to regulate the velocity of the UAV concerned in a manner that would propel the UAV to a target point while

Masoud, Ahmad A.

268

A Practical Activity Recognition Approach Based on the Generic Activity Framework  

Science Journals Connector (OSTI)

In spite of the obvious importance of activity recognition technology for human centric applications, state-of-the-art activity recognition technology is not practical enough for real world deployments because of the insufficient accuracy and lack of ... Keywords: Activity Model, Activity Recognition, Activity Theory, Generic Activity Framework, Multi Layer Neural Network

Sumi Helal, Eunju Kim

2012-07-01T23:59:59.000Z

269

Solution Ionic Strength Engineering as a Generic Strategy to Coat Graphene Oxide (GO)  

E-Print Network (OSTI)

Solution Ionic Strength Engineering as a Generic Strategy to Coat Graphene Oxide (GO) on Various Functional Particles and Its Application in High-Performance Lithium- Sulfur (Li-S) Batteries Jiepeng Rong Angeles, California 90089, United States Graphene oxide (GO) synthesis GO used in this study was prepared

Zhou, Chongwu

270

Generic Models in the Design of Solar Commercial Buildings , N. ISAACS1  

E-Print Network (OSTI)

of generic models were developed. These include: Standard data on building materials used in commercial buildings in New Zealand; Materials data collated into descriptions of standard buildings which are representative of commercial building `types'; Standard building model descriptions which are intended to provide

Amor, Robert

271

Development of generic floor response spectra for equipment qualification for seismic loads  

SciTech Connect

A generic floor response spectra has been developed for use in the qualification of electrical and mechanical equipment in operating nuclear power plants. Actual PWR and BWR - Mark I structural models were used as representative of a class of structures. For each model, the stiffness properties were varied, with the same mass, so as to extend the fundamental base structure natural frequency from 2 cps to 36 cps. This resulted in fundamental mode coupled natural frequencies as low as 0.86 cps and as high as 30 cps. The characteristics of 1000 floor response spectra were studied to determine the generic spectra. A procedure for its application to any operating plant has been established. The procedure uses as much or as little information that currently exists at the plant relating to the question of equipment qualification. A generic floor response spectra is proposed for the top level of a generic structure. Reduction factors are applied to the peak acceleration for equipment at lower levels.

Curreri, J.R.; Costantino, C.J.

1984-10-01T23:59:59.000Z

272

13 - Generation IV reactor designs, operation and fuel cycle  

Science Journals Connector (OSTI)

Abstract: This chapter looks at Generation IV nuclear reactors, such as the very high-temperature reactor (VHTR), the supercritical water reactor (SCWR), the molten salt reactor (MSR), the sodium-cooled fast reactor (SFR), the lead-cooled fast reactor (LFR) and the gas-cooled fast reactor (GFR). Reactor designs and fuel cycles are also described.

N. Cerullo; G. Lomonaco

2012-01-01T23:59:59.000Z

273

New Remote Method for Estimation of Contamination Levels of Reactor Equipment - 13175  

SciTech Connect

Projects for decommissioning of shutdown reactors and reactor facilities carried out in several countries, including Russia. In the National Research Centre 'Kurchatov Institute' decontamination and decommissioning of the research reactor MR (Material Testing Reactor) has been initiated. The research reactor MR has a long history and consists of nine loop facilities for experiments with different kinds of fuel. During the operation of main and auxiliary equipment of reactors it was subjected to strong radioactive contamination. The character of this contamination requires individual strategies for the decontamination work. This requires information about the character of the distribution of radioactive contamination of equipment in the premises. A detailed radiation survey of these premises using standard dosimetric equipment is almost impossible because of high levels of radiation and high-density of the equipment that does not allow identifying the most active fragments using standard tools of measurement. The problem can be solved using the method of remote measurements of distribution of radioactivity with help of the collimated gamma-ray detectors. For radiation surveys of the premises of loop installations remotely operated spectrometric collimated system was used [1, 2, 3]. As a result of the work, maps of the distribution of activity and dose rate for surveyed premises were plotted and superimposed on its photo. The new results of measurements in different areas of the reactor and at its loop installations, with emphasis on the radioactive survey of highly-contaminated samples, are presented. (authors)

Danilovich, Alexey; Ivanov, Oleg; Potapov, Victor; Semenov, Sergey; Semin, Ilya; Smirnov, Sergey; Stepanov, Vyacheslav; Volkovich, Anatoly [National Research Centre 'Kurchatov Institute', Moscow (Russian Federation)] [National Research Centre 'Kurchatov Institute', Moscow (Russian Federation)

2013-07-01T23:59:59.000Z

274

University Reactor Conversion Lessons Learned Workshop for Purdue University Reactor  

SciTech Connect

The Department of Energy’s Idaho National Laboratory, under its programmatic responsibility for managing the University Research Reactor Conversions, has completed the conversion of the reactor at Purdue University Reactor. With this work completed and in anticipation of other impending conversion projects, the INL convened and engaged the project participants in a structured discussion to capture the lessons learned. The lessons learned process has allowed us to capture gaps, opportunities, and good practices, drawing from the project team’s experiences. These lessons will be used to raise the standard of excellence, effectiveness, and efficiency in all future conversion projects.

Eric C. Woolstenhulme; Dana M. Hewit

2008-09-01T23:59:59.000Z

275

Assessment of torsatrons as reactors  

SciTech Connect

Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors because stellarators have no dangerous disruptions and no need for continuous current drive or power recirculated to the plasma, both easing the first wall, blanket, and shield design; less severe constraints on the plasma parameters and profiles; and better access for maintenance. This study shows that a reactor based on the torsatron configuration (a stellarator variant) could also have up to double the mass utilization efficiency (MUE) and a significantly lower cost of electricity (COE) than a conventional tokamak reactor (ARIES-I) for a range of assumptions. Torsatron reactors can have much smaller coil systems than tokamak reactors because the coils are closer to the plasma and they have a smaller cross section (higher average current density because of the lower magnetic field). The reactor optimization approach and the costing and component models are those used in the current stage of the ARIES-I tokamak reactor study. Typical reactor parameters for a 1-GW(e) Compact Torsatron reactor example are major radius R[sub 0] = 6.6-8.8 m, on-axis magnetic field B[sup 0] = 4.8-7.5 T, B[sub max] (on coils) = 16 T, MUE 140-210 kW(e)/tonne, and COE (in constant 1990 dollars) = 67-79 mill/kW(e)h. The results are relatively sensitive to assumptions on the level of confinement improvement and the blanket thickness under the inboard half of the helical windings but relatively insensitive to other assumptions.

Lyon, J.F. (Oak Ridge National Lab., TN (United States)) [Oak Ridge National Lab., TN (United States); Painter, S.L. (Australian National Univ., Canberra, ACT (Australia)) [Australian National Univ., Canberra, ACT (Australia)

1992-12-01T23:59:59.000Z

276

Assessment of torsatrons as reactors  

SciTech Connect

Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors because stellarators have no dangerous disruptions and no need for continuous current drive or power recirculated to the plasma, both easing the first wall, blanket, and shield design; less severe constraints on the plasma parameters and profiles; and better access for maintenance. This study shows that a reactor based on the torsatron configuration (a stellarator variant) could also have up to double the mass utilization efficiency (MUE) and a significantly lower cost of electricity (COE) than a conventional tokamak reactor (ARIES-I) for a range of assumptions. Torsatron reactors can have much smaller coil systems than tokamak reactors because the coils are closer to the plasma and they have a smaller cross section (higher average current density because of the lower magnetic field). The reactor optimization approach and the costing and component models are those used in the current stage of the ARIES-I tokamak reactor study. Typical reactor parameters for a 1-GW(e) Compact Torsatron reactor example are major radius R{sub 0} = 6.6-8.8 m, on-axis magnetic field B{sup 0} = 4.8-7.5 T, B{sub max} (on coils) = 16 T, MUE 140-210 kW(e)/tonne, and COE (in constant 1990 dollars) = 67-79 mill/kW(e)h. The results are relatively sensitive to assumptions on the level of confinement improvement and the blanket thickness under the inboard half of the helical windings but relatively insensitive to other assumptions.

Lyon, J.F. [Oak Ridge National Lab., TN (United States); Painter, S.L. [Australian National Univ., Canberra, ACT (Australia)

1992-12-01T23:59:59.000Z

277

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Initial Assessment of Thermal Annealing Needs and Challenges Initial Assessment of Thermal Annealing Needs and Challenges Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges The most life-limiting structural component in light-water reactors (LWR) is the reactor pressure vessel (RPV) because replacement of the RPV is not considered a viable option at this time. LWR licenses are now being extended from 40y to 60y by the U.S. Nuclear Regulatory Commission (NRC) with intentions to extend licenses to 80y and beyond. The RPV materials exhibit varying degrees of sensitivity to irradiation-induced embrittlement (decreased toughness) , as shown in Fig. 1.1, and extending operation from 40y to 80y implies a doubling of the neutron exposure for the RPV. Thus,

278

Nuclear reactor downcomer flow deflector  

DOE Patents (OSTI)

A nuclear reactor having a coolant flow deflector secured to a reactor core barrel in line with a coolant inlet nozzle. The flow deflector redirects incoming coolant down an annulus between the core barrel and the reactor vessel. The deflector has a main body with a front side facing the fluid inlet nozzle and a rear side facing the core barrel. The rear side of the main body has at least one protrusion secured to the core barrel so that a gap exists between the rear side of the main body adjacent the protrusion and the core barrel. Preferably, the protrusion is a relief that circumscribes the rear side of the main body.

Gilmore, Charles B. (Greensburg, PA); Altman, David A. (Pittsburgh, PA); Singleton, Norman R. (Murrysville, PA)

2011-02-15T23:59:59.000Z

279

Migration and retention of elements at the Oklo natural reactor  

SciTech Connect

The Oklo natural reactor, Gabon, permits study of fission-produced elemental behavior in a natural geologic environment. The uranium ore that sustained fission reactions formed about 2 billion years before present (BYBP), and the reactor was operative for about 5 x 10/sup 5/ yrs between about 1.95 to 2 BYBP. The many tons of fission products can, for the most part, be studied for their abundance and distribution today. Since reactor shutdown, many fissiogenic elements have not migrated from host pitchblende, and several others have migrated only a few tens of meters from the reactor ore. Only Xe and Kr have apparently been largely removed from the reactor zones. An element by element assessment of the Oklo rocks' ability to retain the fission products, and actinides and radiogenic Pb and Bi as well, leads to the conclusion that no widespread migration of the elements occurred. This suggests that rocks with more favorable geologic characteristics are indeed well suited for consideration for the storage of radioactive waste.

Brookins, D.G.

1982-01-01T23:59:59.000Z

280

Modification of the Core Cooling System of TRIGA 2000 Reactor  

SciTech Connect

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

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

2010-06-22T23:59:59.000Z

Note: This page contains sample records for the topic "reactor generic distributed" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


281

2012 Annual Report Research Reactor Infrastructure Program  

SciTech Connect

The content of this report is the 2012 Annual Report for the Research Reactor Infrastructure Program.

Douglas Morrell

2012-11-01T23:59:59.000Z

282

Tritium diagnostics in a fusion reactor  

Science Journals Connector (OSTI)

Methods for controlling tritium in a fusion reactor are reviewed. The characteristic features of the...

A. I. Markin; N. I. Syromyatnikov; A. M. Belov

2010-05-01T23:59:59.000Z

283

Combustion synthesis continuous flow reactor  

DOE Patents (OSTI)

The present invention is a reactor for combustion synthesis of inorganic powders. The reactor includes a reaction vessel having a length and a first end and a second end. The reaction vessel further has a solution inlet and a carrier gas inlet. The reactor further has a heater for heating both the solution and the carrier gas. In a preferred embodiment, the reaction vessel is heated and the solution is in contact with the heated reaction vessel. It is further preferred that the reaction vessel be cylindrical and that the carrier gas is introduced tangentially into the reaction vessel so that the solution flows helically along the interior wall of the reaction vessel. As the solution evaporates and combustion produces inorganic material powder, the carrier gas entrains the powder and carries it out of the reactor.

Maupin, Gary D. (Richland, WA); Chick, Lawrence A. (West Richland, WA); Kurosky, Randal P. (Maple Valley, WA)

1998-01-01T23:59:59.000Z

284

Interfacial effects in fast reactors  

E-Print Network (OSTI)

The problem of increased resonance capture rates near zone interfaces in fast reactor media has been examined both theoretically and experimentally. An interface traversing assembly was designed, constructed and employed ...

Saidi, Mohammad Said

1979-01-01T23:59:59.000Z

285

Unique features of space reactors  

SciTech Connect

Space reactors are designed to meet a unique set of requirements; they must be sufficiently compact to be launched in a rocket to their operational location, operate for many years without maintenance and servicing, operate in extreme environments, and reject heat by radiation to space. To meet these restrictions, operating temperatures are much greater than in terrestrial power plants, and the reactors tend to have a fast neutron spectrum. Currently, a new generation of space reactor power plants is being developed. The major effort is in the SP-100 program, where the power plant is being designed for seven years of full power, and no maintenance operation at a reactor outlet operating temperature of 1350 K. 8 refs., 3 figs., 1 tab.

Buden, D.

1990-01-01T23:59:59.000Z

286

Nuclear Reactors and Technology; (USA)  

SciTech Connect

Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database (EDB) during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency's Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on EDB and Nuclear Science Abstracts (NSA) database. Current information, added daily to EDB, is available to DOE and its contractors through the DOE integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user's needs.

Cason, D.L.; Hicks, S.C. (eds.)

1991-01-01T23:59:59.000Z

287

Reactor physics project final report  

E-Print Network (OSTI)

This is the final report in an experimental and theoretical program to develop and apply single- and few-element methods for the determination of reactor lattice parameters. The period covered by the report is January 1, ...

Driscoll, Michael J.

1970-01-01T23:59:59.000Z

288

Novel Catalytic Membrane Reactors  

SciTech Connect

There are many industrial catalytic organic reversible reactions with amines or alcohols that have water as one of the products. Many of these reactions are homogeneously catalyzed. In all cases removal of water facilitates the reaction and produces more of the desired chemical product. By shifting the reaction to right we produce more chemical product with little or no additional capital investment. Many of these reactions can also relate to bioprocesses. Given the large number of water-organic compound separations achievable and the ability of the Compact Membrane Systems, Inc. (CMS) perfluoro membranes to withstand these harsh operating conditions, this is an ideal demonstration system for the water-of-reaction removal using a membrane reactor. Enhanced reaction synthesis is consistent with the DOE objective to lower the energy intensity of U.S. industry 25% by 2017 in accord with the Energy Policy Act of 2005 and to improve the United States manufacturing competitiveness. The objective of this program is to develop the platform technology for enhancing homogeneous catalytic chemical syntheses.

Stuart Nemser, PhD

2010-10-01T23:59:59.000Z

289

Evaluation of Torsatrons as reactors  

SciTech Connect

Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors. This scoping study, which uses an integrated cost-minimization code that incorporates costing and reactor component models self-consistently with a 1-D energy transport calculation, shows that a torsatron reactor could also be economically competitive with a tokamak reactor. The projected cost of electricity (COE) estimated using the Advanced Reactor Innovation and Evaluation Studies (ARIES) costing algorithms is 65.6 mill/kW(e)h in constant 1992 dollars for a reference 1-GW(e) Compact Torsatron reactor case. The COE is relatively insensitive (<10% variation) over a wide range of assumptions, including variations in the maximum field allowed on the coils, the coil elongation, the shape of the density profile, the beta limit, the confinement multiplier, and the presence of a large loss region for alpha particles. The largest variations in the COE occur for variations in the electrical power output demanded and the plasma-coil separation ratio.

Lyon, J.F. [Oak Ridge National Lab., TN (United States); Gulec, K. [Univ. of Tennessee, Knoxville, TN (United States); Miller, R.L. [Los Alamos National Lab., NM (United States); El-Guebaly, L. [Univ. of Wisconsin, Madison, WI (United States)

1994-03-01T23:59:59.000Z

290

When Do Commercial Reactors Permanently Shut Down?  

Reports and Publications (EIA)

For those wishing to obtain current data, the following resources are available: U.S. reactors, go to the Energy Information Administration's nuclear reactor shutdown list. (Note: As of April 30, 2010, the last U.S. reactor to permanently shut down was Big Rock Point in 1997.) Foreign Reactors, go to the Power Reactor Information System (PRIS) on the International Atomic Energy Agency's website.

2011-01-01T23:59:59.000Z

291

Summary Notes from the 10 July 2007 Generic Technical Issue Discussion on Point of Compliance  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

the 10 July 2007 Generic Technical Issue Discussion on Point of the 10 July 2007 Generic Technical Issue Discussion on Point of Compliance Attendees: Representatives from Department of Energy-Savannah River (DOE-SR), DOE-Headquarters (DOE-HQ), and the U.S. Nuclear Regulatory Commission (NRC), met at the NRC offices in Rockville, Maryland on 10 July 2007. Representatives from the South Carolina Department of Health and Environmental Control (SCDHEC) and State of Idaho participated in the meeting via a teleconference link. Discussion: DOE believes that based on the position papers provided prior to the meeting, DOE and NRC staff have many areas of agreement and no significant areas of disagreement with respect to the specific point of compliance requirements articulated in the respective DOE and NRC requirements. The NRC position paper was based on

292

Buildings Energy Data Book: 1.5 Generic Fuel Quad and Comparison  

Buildings Energy Data Book (EERE)

1 1 Key Definitions Quad: Quadrillion Btu (10^15 or 1,000,000,000,000,000 Btu) Generic Quad for the Buildings Sector: One quad of primary energy consumed in the buildings sector (includes the residential and commercial sectors), apportioned between the various primary fuels used in the sector according to their relative consumption in a given year. To obtain this value, electricity is converted into its primary energy forms according to relative fuel contributions (or shares) used to produce electricity in the given year. Electric Quad (Generic Quad for the Electric Utility Sector): One quad of primary energy consumed at electric utility power plants to supply electricity to end-users, shared among various fuels according to their relative contribution in

293

Summary Notes from 3 October 2007 Generic Technical Issue Discussion on Concentration Averaging  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

3 October 2007 Generic Technical Issue Discussion on 3 October 2007 Generic Technical Issue Discussion on Concentration Averaging Attendees: Representatives from Department of Energy-Headquarters (DOE-HQ) and the U.S. Nuclear Regulatory Commission (NRC) met at the DOE offices in Germantown, Maryland on 3 October 2007. Representatives from Department of Energy-Savannah River (DOE-SR) and the South Carolina Department of Health and Environmental Control (SCDHEC) participated in the meeting via a teleconference link. Discussion: DOE believes that based on the position papers provided prior to the meeting, DOE and NRC staff have many areas of agreement and no significant areas of disagreement with respect to the concentration averaging requirements articulated in the respective DOE and NRC requirements. The NRC position paper was based on NUREG-

294

Catalytic Reactor For Oxidizing Mercury Vapor  

DOE Patents (OSTI)

A catalytic reactor (10) for oxidizing elemental mercury contained in flue gas is provided. The catalyst reactor (10) comprises within a flue gas conduit a perforated corona discharge plate (30a, b) having a plurality of through openings (33) and a plurality of projecting corona discharge electrodes (31); a perforated electrode plate (40a, b, c) having a plurality of through openings (43) axially aligned with the through openings (33) of the perforated corona discharge plate (30a, b) displaced from and opposing the tips of the corona discharge electrodes (31); and a catalyst member (60a, b, c, d) overlaying that face of the perforated electrode plate (40a, b, c) opposing the tips of the corona discharge electrodes (31). A uniformly distributed corona discharge plasma (1000) is intermittently generated between the plurality of corona discharge electrode tips (31) and the catalyst member (60a, b, c, d) when a stream of flue gas is passed through the conduit. During those periods when corona discharge (1000) is not being generated, the catalyst molecules of the catalyst member (60a, b, c, d) adsorb mercury vapor contained in the passing flue gas. During those periods when corona discharge (1000) is being generated, ions and active radicals contained in the generated corona discharge plasma (1000) desorb the mercury from the catalyst molecules of the catalyst member (60a, b, c, d), oxidizing the mercury in virtually simultaneous manner. The desorption process regenerates and activates the catalyst member molecules.

Helfritch, Dennis J. (Baltimore, MD)

1998-07-28T23:59:59.000Z

295

A reactor core on-line monitoring program - COMP  

SciTech Connect

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

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

2012-07-01T23:59:59.000Z

296

The backflow cell model for fluidized bed catalytic reactors  

E-Print Network (OSTI)

that the backmixing of gas in a small fluidized bed with high length to diameter rati. o is relatively small. Hence, it was recommended. that reaction rate studies in fluidized bed reactors be correlated on the basis oi' piston flow~ neglecting mixing. Nay (19...) points out that the straight line obtained on plotting the results of Gilliland's ex- periment on a paper with semilogarithmic coordinates, can be used to characterize the residence time distribution introduced by Danckwerts (6). A steep slope, he...

Ganapathy, E. V

2012-06-07T23:59:59.000Z

297

Comments on Americium Volatilization during Fuel Fabrication for Fast Reactors  

SciTech Connect

The physical processes relevant to the fabrication of metallic and ceramic nuclear fuels are analyzed, with attention to recycling of fuels containing U, Pu, and minor volatile actinides for the use in fast reactors. This analysis is relevant to the development of a process model that can be used for the numerical simulation and prediction of the spatial distribution of composition in the fuel, an important factor in fuel performance.

Sabau, Adrian S [ORNL; Ohriner, Evan Keith [ORNL

2008-01-01T23:59:59.000Z

298

Exposure conditions of reactor internals of Rovno VVER-440 NPP units 1 and 2  

SciTech Connect

Results of determination of irradiation conditions for vessel internals of VVER-440 reactor No. 1 and 2 at Rovno Nuclear Power Plant, obtained by specialists at Inst. for Nuclear Research Kyiv (Ukraine)), and Nuclear Research Inst. Rez (Czech Republic)), are presented. To calculate neutron transport, detailed calculation models of these reactors were prepared. Distribution of neutron flux functionals on the surface of reactor VVER-440 baffle and core barrel for different core loads was studied. Agreement between results obtained by specialists at Inst. for Nuclear Research and at Nuclear Research Inst. is shown. (authors)

Grytsenko, O.V.; Pugach, S.M.; Diemokhin, V.L.; Bukanov, V.N. [Inst. for Nuclear Research, Kyiv, 03680 (Ukraine); Marek, M.; Vandlik, S. [Nuclear Research Inst. Rez Plc., Rez, 25068 (Czech Republic)

2011-07-01T23:59:59.000Z

299

Distribution Workshop  

Energy.gov (U.S. Department of Energy (DOE))

On September 24-26, 2012, the GTT presented a workshop on grid integration on the distribution system at the Sheraton Crystal City near Washington, DC.

300

(Liquid metal reactor/fast breeder reactor research and development)  

SciTech Connect

The second meeting of the UJCC was held in Japan on June 6--8, 1990. The first day was devoted to presentations of the status of the US and Japanese Fast Breeder Reactor (FBR) programs and the status of specific areas of cooperative work. Briefly, the Japanese are following the FBR development program which has been in place since the 1970s. This program includes an FBR test reactor (JOYO), a pilot-scale reactor (MONJU), a demonstration-scale plant, and commercial-scale plants by about 2020. The US program has been redirected toward an actinide recycle mission using metal fuel and pyroprocessing of spent fuel to recovery both Pu and the higher actinides for return to the Liquid Metal Reactor (LMR). The second day was spent traveling from Tokyo to Tsuruga for a tour of the MONJU reactor. The tour was especially interesting. The third day was spent writing the minutes of the meeting and the return trip to Tokyo.

Homan, F.J.

1990-06-20T23:59:59.000Z

Note: This page contains sample records for the topic "reactor generic distributed" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


301

Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency  

SciTech Connect

Suggested for Track 7: Advances in Reactor Core Design and In-Core Management _____________________________________________________________________________________ Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency R. Wigeland and K. Hamman Idaho National Laboratory Given the ability of fast reactors to effectively transmute the transuranic elements as are present in spent nuclear fuel, fast reactors are being considered as one element of future nuclear power systems to enable continued use and growth of nuclear power by limiting high-level waste generation. However, a key issue for fast reactors is higher electricity cost relative to other forms of nuclear energy generation. The economics of the fast reactor are affected by the amount of electric power that can be produced from a reactor, i.e., the thermal efficiency for electricity generation. The present study is examining the potential for fast reactor subassembly design changes to improve the thermal efficiency by increasing the average coolant outlet temperature without increasing peak temperatures within the subassembly, i.e., to make better use of current technology. Sodium-cooled fast reactors operate at temperatures far below the coolant boiling point, so that the maximum coolant outlet temperature is limited by the acceptable peak temperatures for the reactor fuel and cladding. Fast reactor fuel subassemblies have historically been constructed using a large number of small diameter fuel pins contained within a tube of hexagonal cross-section, or hexcan. Due to this design, there is a larger coolant flow area next to the hexcan wall as compared to flow area in the interior of the subassembly. This results in a higher flow rate near the hexcan wall, overcooling the fuel pins next to the wall, and a non-uniform coolant temperature distribution. It has been recognized for many years that this difference in sodium coolant temperature was detrimental to achieving greater thermal efficiency, since it causes the fuel pins in the center of the subassembly to operate at higher temperatures than those near the hexcan walls, and it is the temperature limit(s) for those fuel pins that limits the average coolant outlet temperature. Fuel subassembly design changes are being investigated using computational fluid dynamics (CFD) to quantify the effect that the design changes have on reducing the intra-subassembly coolant flow and temperature distribution. Simulations have been performed for a 19-pin test subassembly geometry using typical fuel pin diameters and wire wrap spacers. The results have shown that it may be possible to increase the average coolant outlet temperature by 20 C or more without changing the peak temperatures within the subassembly. These design changes should also be effective for reactor designs using subassemblies with larger numbers of fuel pins. R. Wigeland, Idaho National Laboratory, P.O. Box 1625, Mail Stop 3860, Idaho Falls, ID, U.S.A., 83415-3860 email – roald.wigeland@inl.gov fax (U.S.) – 208-526-2930

R. Wigeland; K. Hamman

2009-09-01T23:59:59.000Z

302

Advanced Nuclear Reactors | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Advanced Nuclear Advanced Nuclear Reactors Advanced Nuclear Reactors Turbulent Flow of Coolant in an Advanced Nuclear Reactor Visualizing Coolant Flow in Sodium Reactor Subassemblies Sodium-cooled Fast Reactor (SFR) Coolant Flow At the heart of a nuclear power plant is the reactor. The fuel assembly is placed inside a reactor vessel where all the nuclear reactions occur to produce the heat and steam used for power generation. Nonetheless, an entire power plant consists of many other support components and key structures like coolant pipes; pumps and tanks including their surrounding steel framing; and concrete containment and support structures. The Reactors Product Line within NEAMS is concerned with modeling the reactor vessel as well as those components of a complete power plant that

303

Advanced Reactor Technology Documents | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Nuclear Reactor Technologies » Advanced Reactor Nuclear Reactor Technologies » Advanced Reactor Technologies » Advanced Reactor Technology Documents Advanced Reactor Technology Documents January 30, 2013 Advanced Reactor Concepts Technical Review Panel Report This report documents the establishment of a technical review process and the findings of the Advanced Reactor Concepts (ARC) Technical Review Panel (TRP).1 The intent of the process is to identify R&D needs for viable advanced reactor concepts in order to inform DOE-NE R&D investment decisions. A goal of the process is to facilitate greater engagement between DOE and industry. The process involved establishing evaluation criteria, conducting a pilot review, soliciting concept inputs from industry entities, reviewing the concepts by TRP members and compiling the

304

Microsoft Word - power_reactors_briggs.doc  

NLE Websites -- All DOE Office Websites (Extended Search)

Most common - Boiling Water and Pressurized Most common - Boiling Water and Pressurized Water Reactors About 80% of the world's nuclear reactors used for generating electricity are either boiling water reactors or pressurized water reactors. Of these, about 30% are boiling water reactors and 70% are pressurized water reactors. All power reactors currently in use in the United States are of these two types. Both types of reactors have been very successfully used for reliable, on-demand, emissions-free electricity generation for decades. How does a boiling water reactor work? Water flows from the bottom of the fuel to the top of the fuel, and as it moves past the fuel, it carries away the heat produced by the

305

PIA - Advanced Test Reactor National Scientific User Facility...  

Energy Savers (EERE)

Advanced Test Reactor National Scientific User Facility Users Week 2009 PIA - Advanced Test Reactor National Scientific User Facility Users Week 2009 PIA - Advanced Test Reactor...

306

Global Optimization of Chemical Reactors and Kinetic Optimization  

E-Print Network (OSTI)

Model; 3-D; Monolith; Reactor; Optimization Introduction TheAngeles Global Optimization of Chemical Reactors and KineticGlobal Optimization of Chemical Reactors and Kinetic

ALHUSSEINI, ZAYNA ISHAQ

2013-01-01T23:59:59.000Z

307

Green's Functions for Surface Waves in a Generic Velocity Structure 1 Victor C. Tsai and Sarun Atiganyanun* 3  

E-Print Network (OSTI)

1 Green's Functions for Surface Waves in a Generic Velocity Structure 1 and Green's functions have been well established 14 for many decades. However, or Green's function surface displacement. We address this gap in the 19 literature

308

UCLA program in reactor studies: The ARIES tokamak reactor study  

SciTech Connect

The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D-{sup 3}He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs.

Not Available

1991-01-01T23:59:59.000Z

309

Rapid starting methanol reactor system  

DOE Patents (OSTI)

The invention relates to a methanol-to-hydrogen cracking reactor for use with a fuel cell vehicular power plant. The system is particularly designed for rapid start-up of the catalytic methanol cracking reactor after an extended shut-down period, i.e., after the vehicular fuel cell power plant has been inoperative overnight. Rapid system start-up is accomplished by a combination of direct and indirect heating of the cracking catalyst. Initially, liquid methanol is burned with a stoichiometric or slightly lean air mixture in the combustion chamber of the reactor assembly. The hot combustion gas travels down a flue gas chamber in heat exchange relationship with the catalytic cracking chamber transferring heat across the catalyst chamber wall to heat the catalyst indirectly. The combustion gas is then diverted back through the catalyst bed to heat the catalyst pellets directly. When the cracking reactor temperature reaches operating temperature, methanol combustion is stopped and a hot gas valve is switched to route the flue gas overboard, with methanol being fed directly to the catalytic cracking reactor. Thereafter, the burner operates on excess hydrogen from the fuel cells.

Chludzinski, Paul J. (38 Berkshire St., Swampscott, MA 01907); Dantowitz, Philip (39 Nancy Ave., Peabody, MA 01960); McElroy, James F. (12 Old Cart Rd., Hamilton, MA 01936)

1984-01-01T23:59:59.000Z

310

Development of a PIRT (phenomena identification and ranking table) for a postulated double-ended guillotine break in a production reactor  

SciTech Connect

The US Nuclear Regulatory Commission has developed a generic methodology to quantify the uncertainty in best estimate computer codes used to license commercial light water reactors. This same methodology is equally applicable to other reactor designs with regards to providing a technical basis which supports the establishment and demonstration of compliance with safe operating margins. One of the cornerstones of the method is the identification and ranking of phenomena that are important to the postulated scenario. This paper references descriptions of the total methodology, describes the first three steps (i.e, through the identification and ranking of phenomena), and summarizes the results of the application of the methodology to a double-ended guillotine break loss of coolant accident in a production reactor. 6 refs., 8 figs., 5 tabs.

Hanson, R.G.; Wilson, G.E.; Ortiz, M.G. (EG and G Idaho, Inc., Idaho Falls, ID (USA)); Griggs, D.P. (Savannah River Lab., Aiken, SC (USA))

1989-01-01T23:59:59.000Z

311

N Reactor Placed In Interim Safe Storage: Largest Hanford Reactor Cocooning  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

N Reactor Placed In Interim Safe Storage: Largest Hanford Reactor N Reactor Placed In Interim Safe Storage: Largest Hanford Reactor Cocooning Project Now Complete N Reactor Placed In Interim Safe Storage: Largest Hanford Reactor Cocooning Project Now Complete June 14, 2012 - 12:00pm Addthis Media Contacts Cameron Hardy Cameron.Hardy@rl.doe.gov 509-376-5365 Mark McKenna mmckenna@wch-rcc.com 509-372-9032 RICHLAND, WASH. - The U.S. Department of Energy's (DOE's) River Corridor contractor, Washington Closure Hanford, has completed placing N Reactor in interim safe storage, a process also known as "cocooning." N Reactor was the last of nine plutonium production reactors to be shut down at DOE's Hanford Site in southeastern Washington state. It was Hanford's longest-running reactor, operating from 1963 to 1987. "In the 1960's, N Reactor represented the future of energy in America.

312

Graphite Reactor | ornl.gov  

NLE Websites -- All DOE Office Websites (Extended Search)

Graphite Reactor Graphite Reactor 'In the early, desperate days of World War II, the United States launched the top-secret, top-priority Manhattan Project...' In the early, desperate days of U.S. involvement in World War II, American scientists began to fear that the German discovery of uranium fission in 1939 might enable the Nazis to develop a super bomb. Afraid of losing this crucial race, the United States launched the top-secret, top-priority Manhattan Project. The plan was to create two atomic weapons-one fueled by plutonium, the other by enriched uranium. Hanford, Washington, was selected as the site for plutonium production, but before large reactors could be built there, a pilot plant was necessary to prove the feasibility of scaling up from laboratory experiments. A secluded, rural area near Clinton, Tennessee, was

313

Business Opportunities for Small Reactors  

SciTech Connect

This report assesses the market potential and identifies a number of potential paths for developing the small nuclear reactor business. There are several potential opportunities identified and evaluated. Selecting a specific approach for the business development requires additional information related to a specific market and sources of capital to support the investment. If and how a market for small nuclear plants may develop is difficult to predict because of the complexity of the economic and institutional factors that will influence such development. Key factors are; economics, safety, proliferation resistance and investment risk. The economic and political interest of any of the identified markets is also dependent on successful demonstration of the safety and reliability of small nuclear reactor. Obtaining a US-NRC Standard Design approval would be an important development step toward establishing a market for small reactors. (authors)

Minato, Akio; Nishimura, Satoshi [Central Research Institute of Electric Power Industry - CRIEPI, 2-11-1 Iwado-Kita, Komae, Tokyo 201-8511 (Japan); Brown, Neil W. [Lawrence Livermore National Laboratory - LLNL, PO Box 808, Livermore, CA 94551 (United States)

2007-07-01T23:59:59.000Z

314

Level III probabilistic risk assessment for N Reactor  

SciTech Connect

A Level III probabilistic risk assessment (PRA) has been performed for N Reactor, a Department of Energy (DOE) production reactor located on the Hanford reservation in Washington. The objectives of the PRA are to assess the risks to the public and the Hanford site workers posed by the operation of N Reactor, to compare those risks to proposed DOE safety goals, and to identify changes to the plant that could reduce the risk. The scope of the PRA is comprehensive, excluding only sabotage and operation errors of commission. State-of-the-art methodology is employed based largely on the methods developed by Sandia for the US Nuclear Regulatory Commission in support of the NUREG-1150 study of five commercial nuclear power plants. The structure of the probabilistic models allowed complex interactions and dependencies between systems to be explicitly considered. Latin Hypercube sampling techniques were used to develop uncertainty distributions for the risks associated with postulated core damage events initiated by fire, seismic, and internal events as well as the overall combined risk. The combined risk results show that N Reactor meets the primary DOE safety goals and compared favorably to the plants considered in the NUREG-1150 analysis. 36 figs., 81 tabs.

Camp, A.L.; Kunsman, D.M.; Miller, L.A.; Sprung, J.L.; Wheeler, T.A.; Wyss, G.D. (Sandia National Labs., Albuquerque, NM (USA))

1990-04-01T23:59:59.000Z

315

Development of Improved Burnable Poisons for Commercial Nuclear Power Reactors  

SciTech Connect

Burnable poisons are used in nuclear reactors to produce a more level distribution of power in the reactor core and to reduce to necessity for a large control system. An ideal burnable poison would burn at the same rate as the fuel. In this study, separation of neutron-absorbing isotopes was investigated in order to eliminate isotopes that remain as absorbers at the end of fuel life, thus reducing useful fuel life. The isotopes Gd-157, Dy-164, and Er-167 were found to have desirable properties. These isotopes were separated from naturally occurring elements by means of plasma separation to evaluate feasibility and cost. It was found that pure Gd-157 could save approximately $6 million at the end of four years. However, the cost of separation, using the existing facility, made separation cost- ineffective. Using a magnet with three times the field strength is expected to reduce the cost by a factor of ten, making isotopically separated burnable poisons a favorable method of increasing fuel life in commercial reactors, in particular Generation-IV reactors. The project also investigated various burnable poison configurations, and studied incorporation of metallic burnable poisons into fuel cladding.

M. L. Grossbeck J-P.A. Renier Tim Bigelow

2003-09-30T23:59:59.000Z

316

Actinide Burning in CANDU Reactors  

SciTech Connect

Actinide burning in CANDU reactors has been studied as a method of reducing the actinide content of spent nuclear fuel from light water reactors, and thereby decreasing the associated long term decay heat load. In this work simulations were performed of actinides mixed with natural uranium to form a mixed oxide (MOX) fuel, and also mixed with silicon carbide to form an inert matrix (IMF) fuel. Both of these fuels were taken to a higher burnup than has previously been studied. The total transuranic element destruction calculated was 40% for the MOX fuel and 71% for the IMF. (authors)

Hyland, B.; Dyck, G.R. [Atomic Energy of Canada Limited, Chalk River, Ontario, K0J 1J0 (Canada)

2007-07-01T23:59:59.000Z

317

Wealth Distribution  

Science Journals Connector (OSTI)

Walter: What is a just wealth distribution? In my view, it is one that results from respect for proper initial homesteading, for resulting private property rights, and, finally, from any legitimate subsequent ...

Four Arrows; Walter Block

2011-01-01T23:59:59.000Z

318

Special Distribution  

Office of Legacy Management (LM)

Special Distribution Special Distribution Issued: December 1977 ',, Radiological Survey and Decontamination of the Former Main Technical Area (TA-1) at Los Alamos, New Mexico Compiled by A. John Ahlquist Alan K. Stoker Linda K. Trocki c laboratory of, the University of California LOS ALAMOS, NEW MEXICO 87545 An Alfirmdve Action/Equal Opportunity Employer ..-_- .-- .--.-. c T -,--... _ _._-r..l __,.. - .-,_.. ..- _._ -- .--. " . . _ . - . c- - . . . _ -. . _ . - . - . _ - - n - _ _~ ~_. __ _ ~~_ --..&e+ L.';; CONTENTS ABSTRACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .._____ 1 EXECUTIVE SUMMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .._... _._ 2 I. BACKGROUND .............................................. 15

319

BNL | Our History: Reactors as Research Tools  

NLE Websites -- All DOE Office Websites (Extended Search)

> See also: Accelerators > See also: Accelerators Brookhaven History: Using Reactors as Research Tools BGRR Brookhaven Graphite Research Reactor The Brookhaven Graphite Research Reactor (BGRR) was the Laboratory's first big machine and the first peace-time reactor built in the United States following World War II. The reactor's primary mission was to produce neutrons for scientific experimentation and to refine reactor technology. At the time, the BGRR could accommodate more simultaneous experiments than any other reactor. Scientists and engineers from every corner of the U.S. came to use the reactor, which was not only a source of neutrons for experiments, but also an excellent training facility. Researchers used the BGRR's neutrons as tools for studying atomic nuclei and the structure of solids, and to investigate many physical, chemical and

320

New fast-reactor approach. [LMFBR  

SciTech Connect

The design parameters for a 1000 MW LMFBR type reactor are presented. The design requires the multiple primary coolant pumps and heat exchangers to be located around the core within the reactor vessel.

Folkrod, J.R.; Kann, W.J.; Klocksieben, R.H.

1983-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor generic distributed" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


321

Reactor accelerator coupling experiments: a feasability study  

E-Print Network (OSTI)

The Reactor Accelerator Coupling Experiments (RACE) are a set of neutron source driven subcritical experiments under temperature feedback conditions. These experiments will involve coupling an accelerator driven neutron source to a TRIGA reactor...

Woddi Venkat Krishna, Taraknath

2006-08-16T23:59:59.000Z

322

Reactivity control assembly for nuclear reactor  

DOE Patents (OSTI)

Reactivity control assembly for nuclear reactor comprises supports stacked above reactor core for holding control rods. Couplers associated with the supports and a vertically movable drive shaft have lugs at their lower ends for engagement with the supports.

Bollinger, Lawrence R. (Schenectady, NY)

1984-01-01T23:59:59.000Z

323

Inherent safety concepts in nuclear power reactors  

Science Journals Connector (OSTI)

Different inherent safety concepts being considered in fast and thermal reactors are presented after outlining the basic goals of nuclear reactor safety, the ‘defence in depth’ philosophy to achieve these goal...

O M Pal Singh; R Shankar Singh

1989-06-01T23:59:59.000Z

324

Choice of coils for a fusion reactor  

Science Journals Connector (OSTI)

...configurations. The most ambitious is the International Thermonuclear Experimental Reactor, a large tokamak planned for construction...configuration has features in common with the International Thermonuclear Experimental Reactor experiment. Mathematical Model We...

Romeo Alexander; Paul R. Garabedian

2007-01-01T23:59:59.000Z

325

Computational Analysis of Fluid Flow in Pebble Bed Modular Reactor  

E-Print Network (OSTI)

High Temperature Gas-cooled Reactor (HTGR) is a Generation IV reactor under consideration by Department of Energy and in the nuclear industry. There are two categories of HTGRs, namely, Pebble Bed Modular Reactor (PBMR) and Prismatic reactor. Pebble...

Gandhir, Akshay

2012-10-19T23:59:59.000Z

326

The development of structural materials for fusion reactors  

Science Journals Connector (OSTI)

...severely exposed parts of future fusion reactors and pose key problems...successful implementation of fusion reactors as an efficient source...conditions in the International Thermonuclear Experimental Reactor (ITER...environmental attractiveness of fusion reactors. In this paper...

1999-01-01T23:59:59.000Z

327

Utilization of Refractory Metals and Alloys in Fusion Reactor Structures  

Science Journals Connector (OSTI)

In design of fusion reactors, structural material selection is very crucial to improve reactor’s performance. Different types of materials have been proposed for use in fusion reactor structures. Among these mate...

Mustafa Übeyli; ?enay Yalç?n

2006-12-01T23:59:59.000Z

328

Reliability assessment of Passive Containment Cooling System of an Advanced Reactor using APSRA methodology  

Science Journals Connector (OSTI)

Abstract Passive Systems are increasingly playing a prominent role in the advanced nuclear reactor systems and are being utilised in normal operations as well as safety systems of the reactors following an accident. The Passive Containment Cooling System (PCCS) is one of the several passive safety features in an Advanced Reactor (AHWR). In this paper, the APSRA methodology has been employed for reliability evaluation of the PCCS of AHWR. Performance assessment of the PCCS is initially performed during a postulated design basis LOCA using the best-estimate code RELAP5/Mod 3.2. The parameters affecting the system performance are then identified and considered for further analysis. Based on some pre-determined failure criterion, the failure surface for the system is predicted using the best-estimate code taking into account the deviations of the identified parameters from their nominal states as well as the model uncertainties inherent to the best estimate code. Root diagnosis is then carried out to determine the various failure causes, which occurs mainly due to malfunctioning of mechanical components. The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. The reliability of the PCCS is then evaluated from the probability of availability of these components.

Mukesh Kumar; Aranyak Chakravarty; A.K. Nayak; Hari Prasad; V. Gopika

2014-01-01T23:59:59.000Z

329

On the estimation of the unknown reactivity coefficients in a CANDU reactor  

Science Journals Connector (OSTI)

A space-time kinetics based inverse architecture method is suggested to analyze the reactivity variations associated with power excursions in a generic CANDU reactor. It is intended to provide diagnosis tools to gain enhanced control thereby ensuring safe operation of the plant. A methodology for analyzing the data available from the in core flux detectors and extracting the unknown reactivity coefficients is presented. The proposed system uses a reference model in conjunction with an optimal estimator. The reference model is composed of a state space representation of the space-time dynamics of neutron flux in the core, based on modal expansion approximation, and a time domain optimal estimator filter. We investigated three different estimation techniques based on recursive prediction error method (RPEM), dual extended Kalman filter (DEKF), and joint extended Kalman filter (JEKF). We compared their applicability to the estimation of coolant-void dynamic reactivity in loss-of-coolant accident in a CANDU reactor. The state equations also include the characteristics of the detector responses. The thermal hydraulic models were not included in the calculations. Two different types of detectors are considered in this analysis, the over prompt responsive Platinum detector of the reactor shutdown systems, and the under delayed responsive Vanadium detector of the flux mapping system.

Lobat Tayebi; Daryoosh Vashaee

2013-01-01T23:59:59.000Z

330

Digital computer operation of a nuclear reactor  

DOE Patents (OSTI)

A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state.

Colley, Robert W. (Richland, WA)

1984-01-01T23:59:59.000Z

331

Liquid metal cooled nuclear reactor plant system  

DOE Patents (OSTI)

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA)

1993-01-01T23:59:59.000Z

332

Digital computer operation of a nuclear reactor  

DOE Patents (OSTI)

A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state.

Colley, R.W.

1982-06-29T23:59:59.000Z

333

Light Water Reactor Sustainability (LWRS) Program  

NLE Websites -- All DOE Office Websites (Extended Search)

Light Water Reactor Sustainability (LWRS) Program Login Instructions go here. User ID: Password: Log In Forgot your password?...

334

High-Fidelity Light Water Reactor Analysis with the Numerical Nuclear Reactor  

Science Journals Connector (OSTI)

Technical Paper / Mathematics and Computation, Supercomputing, Reactor Physics and Nuclear and Biological Applications

David P. Weber; Tanju Sofu; Won Sik Yang; Thomas J. Downar; Justin W. Thomas; Zhaopeng Zhong; Jin Young Cho; Kang Seog Kim; Tae Hyun Chun; Han Gyu Joo; Chang Hyo Kim

335

How far is a Fusion Power Reactor from an Experimental Reactor?  

E-Print Network (OSTI)

be able to move directly and safely to a "first of a kind" reactor. The main conditions to be satisfied / experimental evidence. To assess the reactor relevance of ITER, rather than a comparison between ITER and one1 How far is a Fusion Power Reactor from an Experimental Reactor? R. Toschi(1) , P. Barabaschi(2

336

Monte Carlo Domain Decomposition for Robust Nuclear Reactor Analyses  

Science Journals Connector (OSTI)

Abstract Monte Carlo (MC) neutral particle transport codes are considered the gold-standard for nuclear simulations, but they cannot be robustly applied to high-fidelity nuclear reactor analysis without accommodating several terabytes of materials and tally data. While this is not a large amount of aggregate data for a typical high performance computer, MC methods are only embarrassingly parallel when the key data structures are replicated for each processing element, an approach which is likely infeasible on future machines. The present work explores the use of spatial domain decomposition to make full-scale nuclear reactor simulations tractable with Monte Carlo methods, presenting a simple implementation in a production-scale code. Good performance is achieved for mesh-tallies of up to 2.39TB distributed across 512 compute nodes while running a full-core reactor benchmark on the Mira Blue Gene/Q supercomputer at the Argonne National Laboratory. In addition, the effects of load imbalances are explored with an updated performance model that is empirically validated against observed timing results. Several load balancing techniques are also implemented to demonstrate that imbalances can be largely mitigated, including a new and efficient way to distribute extra compute resources across coarse domain meshes.

Nicholas Horelik; Andrew Siegel; Benoit Forget; Kord Smith

2014-01-01T23:59:59.000Z

337

Beyond QoS signaling: A new generic IP signaling framework  

Science Journals Connector (OSTI)

This paper describes the design principles and an introduction of a framework and protocols for generic IP signaling, namely the Cross-Application Signaling Protocol (CASP) and its signaling applications. While reusing certain features of the existing RSVP protocol, CASP overcomes its shortcomings and may be deployed as a replacement technology to provide simpler, mobility-supported, more extensible and more secure signaling services in IP based networks. This paper discusses challenges of today’s IP signaling protocols and addresses fundamentals and key aspects of CASP and its current signaling applications. In addition, a comparison with previous signaling protocol proposals and an outlook of future work in this area are also given.

Xiaoming Fu; Hannes Tschofenig; Dieter Hogrefe

2006-01-01T23:59:59.000Z

338

Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation  

E-Print Network (OSTI)

Physics Optimization of Breed and Burn Fast Reactor Systems.reactors: Fabrication and properties and their optimization.

Heidet, Florent

2010-01-01T23:59:59.000Z

339

DOSE RATES FROM NEUTRON ACTIVATION OF FUSION REACTOR COMPONENTS  

E-Print Network (OSTI)

NEUTRON ACTIVATION OF FUSION REACTOR C01WONENTS LawrenceNeutron Activation of Fusion Reactor Components Lawrence

Ruby, Lawrence

2014-01-01T23:59:59.000Z

340

Nuclear Reactor Safety Design Criteria  

Directives, Delegations, and Requirements

The order establishes nuclear safety criteria applicable to the design, fabrication, construction, testing, and performance requirements of nuclear reactor facilities and safety class structures, systems, and components (SSCs) within these facilities. Cancels paragraphs 8a and 8b of DOE 5480.6. Cancels DOE O 5480.6 in part. Certified 11-18-10.

1993-01-19T23:59:59.000Z

Note: This page contains sample records for the topic "reactor generic distributed" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


341

Computer aided nuclear reactor modeling  

E-Print Network (OSTI)

Nuclear reactor modeling is an important activity that lets us analyze existing as well as proposed systems for safety, correct operation, etc. The quality of a analysis is directly proportional to the quality of the model used. In this work we look...

Warraich, Khalid Sarwar

2012-06-07T23:59:59.000Z

342

Nozzle for electric dispersion reactor  

DOE Patents (OSTI)

A nozzle for an electric dispersion reactor includes two concentric electrodes, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode.

Sisson, Warren G. (Oak Ridge, TN); Basaran, Osman A. (Oak Ridge, TN); Harris, Michael T. (Knoxville, TN)

1998-01-01T23:59:59.000Z

343

Nozzle for electric dispersion reactor  

DOE Patents (OSTI)

A nozzle for an electric dispersion reactor includes two concentric electrodes, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode. 4 figs.

Sisson, W.G.; Basaran, O.A.; Harris, M.T.

1998-04-14T23:59:59.000Z

344

Laminar Entrained Flow Reactor (Fact Sheet)  

SciTech Connect

The Laminar Entrained Flow Reactor (LEFR) is a modular, lab scale, single-user reactor for the study of catalytic fast pyrolysis (CFP). This system can be employed to study a variety of reactor conditions for both in situ and ex situ CFP.

Not Available

2014-02-01T23:59:59.000Z

345

International Journal of Chemical Reactor Engineering  

E-Print Network (OSTI)

International Journal of Chemical Reactor Engineering Volume 3 2005 Article A17 Optimal Operation, a single re- action takes place in the reactor and the operational objective is to compute the optimal feed is illustrated via simulation of two semi-batch reactor applications. KEYWORDS: Dynamic Optimization, Batch

Palanki, Srinivas

346

The Integral Fast Reactor (IFR) - Reactors designed/built by Argonne  

NLE Websites -- All DOE Office Websites (Extended Search)

Integral Fast Reactor Integral Fast Reactor About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

347

Design of generic coal conversion facilities: Process release---Refining and upgrading  

SciTech Connect

The refinery and upgrade process development unit (PDU) is designed to upgrade liquid hydrocarbon products from the direct and indirect liquefaction PDU's to transportation fuels. The refinery will comprise of the following reactor systems: (a) Hydrotreating (b) Hydrocracking (c) Reforming. The three reactor systems will share common feed preparation, product separation and fractionation sections. The refinery is being designed to operate independently of the other PDU's. The use of common feed and product handling systems will permit operation of one process reactor system at a time in the refinery. In addition, the hydrotreater and hydrocracker will be operable in series. The process is designed to utilize intermediate storage and maximize the use of equipment.

Not Available

1991-09-01T23:59:59.000Z

348

Design of generic coal conversion facilities: Process release---Refining and upgrading  

SciTech Connect

The refinery and upgrade process development unit (PDU) is designed to upgrade liquid hydrocarbon products from the direct and indirect liquefaction PDU`s to transportation fuels. The refinery will comprise of the following reactor systems: (a) Hydrotreating (b) Hydrocracking (c) Reforming. The three reactor systems will share common feed preparation, product separation and fractionation sections. The refinery is being designed to operate independently of the other PDU`s. The use of common feed and product handling systems will permit operation of one process reactor system at a time in the refinery. In addition, the hydrotreater and hydrocracker will be operable in series. The process is designed to utilize intermediate storage and maximize the use of equipment.

Not Available

1991-09-01T23:59:59.000Z

349

Heterogeneous Recycling in Fast Reactors  

SciTech Connect

Current sodium fast reactor (SFR) designs have avoided the use of depleted uranium blankets over concerns of creating weapons grade plutonium. While reducing proliferation risks, this restrains the reactor design space considerably. This project will analyze various blanket and transmutation target configurations that could broaden the design space while still addressing the non-proliferation issues. The blanket designs will be assessed based on the transmutation efficiency of key minor actinide (MA) isotopes and also on mitigation of associated proliferation risks. This study will also evaluate SFR core performance under different scenarios in which depleted uranium blankets are modified to include minor actinides with or without moderators (e.g. BeO, MgO, B4C, and hydrides). This will be done in an effort to increase the sustainability of the reactor and increase its power density while still offering a proliferation resistant design with the capability of burning MA waste produced from light water reactors (LWRs). Researchers will also analyze the use of recycled (as opposed to depleted) uranium in the blankets. The various designs will compare MA transmutation efficiency, plutonium breeding characteristics, proliferation risk, shutdown margins and reactivity coefficients with a current reference sodium fast reactor design employing homogeneous recycling. The team will also evaluate the out-of-core accumulation and/or burn-down rates of MAs and plutonium isotopes on a cycle-by-cycle basis. This cycle-by-cycle information will be produced in a format readily usable by the fuel cycle systems analysis code, VISION, for assessment of the sustainability of the deployment scenarios.

Dr. Benoit Forget; Michael Pope; Piet, Steven J.; Michael Driscoll

2012-07-30T23:59:59.000Z

350

Control of reactor coolant flow path during reactor decay heat removal  

DOE Patents (OSTI)

An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

Hunsbedt, Anstein N. (Los Gatos, CA)

1988-01-01T23:59:59.000Z

351

Grid integrated distributed PV (GridPV).  

SciTech Connect

This manual provides the documentation of the MATLAB toolbox of functions for using OpenDSS to simulate the impact of solar energy on the distribution system. The majority of the functions are useful for interfacing OpenDSS and MATLAB, and they are of generic use for commanding OpenDSS from MATLAB and retrieving information from simulations. A set of functions is also included for modeling PV plant output and setting up the PV plant in the OpenDSS simulation. The toolbox contains functions for modeling the OpenDSS distribution feeder on satellite images with GPS coordinates. Finally, example simulations functions are included to show potential uses of the toolbox functions. Each function in the toolbox is documented with the function use syntax, full description, function input list, function output list, example use, and example output.

Reno, Matthew J.; Coogan, Kyle [Georgia Institute of Technology, Atlanta, GA

2013-08-01T23:59:59.000Z

352

New Methods for Analysis of Spatial Distribution and Coaggregation of Microbial Populations in Complex Biofilms  

Science Journals Connector (OSTI)

...biofilms from different reactors. For this purpose...enables detailed analyses of biofilms grown...the digital image analysis software daime...the statistical reliability of the results...Vertical-distribution analysis. The biofilms in...reasons such as reactor design, shear forces...

Robert Almstrand; Holger Daims; Frank Persson; Fred Sörensson; Malte Hermansson

2013-07-26T23:59:59.000Z

353

AIAA 94-1214: Using generic tool kits to build intelligent systems  

SciTech Connect

The Intelligent Systems and Robotics Center at Sandia National Laboratories is developing technologies for the automation of processes associated with environmental remediation and information-driven manufacturing. These technologies, which focus on automated planning and programming and sensor-based and model-based control, are used to build intelligent systems which are able to generate plans of action, program the necessary devices, and use sensors to react to changes in the environment. By automating tasks through the use of programmable devices tied to computer models which are augmented by sensing, requirements for faster, safer, and cheaper systems are being satisfied. However, because of the need for rapid cost-effective prototyping and multi-laboratory teaming, it is also necessary to define a consistent approach to the construction of controllers for such systems. As a result, the Generic Intelligent System Controller (GISC) concept has been developed. This concept promotes the philosophy of producing generic tool kits which can be used and reused to build intelligent control systems.

Miller, D.J.

1993-12-31T23:59:59.000Z

354

Reactor and Nuclear Systems Division (RNSD)  

NLE Websites -- All DOE Office Websites (Extended Search)

RNSD Home RNSD Home Research Groups Advanced Reactor Systems & Safety Nuclear Data & Criticality Safety Nuclear Security Modeling Radiation Safety Information Computational Center Radiation Transport Reactor Physics Thermal Hydraulics & Irradiation Engineering Used Fuel Systems Staff Details (CV/Bios) Publications Org Chart Contact Us ORNL Staff Only Research Groups Advanced Reactor Systems & Safety Nuclear Data & Criticality Safety Nuclear Security Modeling Radiation Safety Information Computational Center Radiation Transport Reactor Physics Thermal Hydraulics & Irradiation Engineering Used Fuel Systems Reactor and Nuclear Systems Division News Highlights U.S. Rep. Fleischmann touts ORNL as national energy treasure Martin Peng wins Fusion Power Associates Leadership Award

355

Shutdown system for a nuclear reactor  

DOE Patents (OSTI)

An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion.

Groh, Edward F. (Naperville, IL); Olson, Arne P. (Western Springs, IL); Wade, David C. (Naperville, IL); Robinson, Bryan W. (Oak Lawn, IL)

1984-01-01T23:59:59.000Z

356

Reactor monitoring and safeguards using antineutrino detectors  

Science Journals Connector (OSTI)

Nuclear reactors have served as the antineutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these very weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Measurements made with antineutrino detectors could therefore orer an alternative means for verifying the power history and fissile inventory of a reactors, as part of International Atomic Energy Agency (IAEA) and other reactor safeguards regimes. Several erorts to develop this monitoring technique are underway across the globe.

N S Bowden

2008-01-01T23:59:59.000Z

357

Self isolating high frequency saturable reactor  

DOE Patents (OSTI)

The present invention discloses a saturable reactor and a method for decoupling the interwinding capacitance from the frequency limitations of the reactor so that the equivalent electrical circuit of the saturable reactor comprises a variable inductor. The saturable reactor comprises a plurality of physically symmetrical magnetic cores with closed loop magnetic paths and a novel method of wiring a control winding and a RF winding. The present invention additionally discloses a matching network and method for matching the impedances of a RF generator to a load. The matching network comprises a matching transformer and a saturable reactor.

Moore, James A. (Powell, TN)

1998-06-23T23:59:59.000Z

358

Fast-acting nuclear reactor control device  

DOE Patents (OSTI)

A fast-acting nuclear reactor control device for moving and positioning a fety control rod to desired positions within the core of the reactor between a run position in which the safety control rod is outside the reactor core, and a shutdown position in which the rod is fully inserted in the reactor core. The device employs a hydraulic pump/motor, an electric gear motor, and solenoid valve to drive the safety control rod into the reactor core through the entire stroke of the safety control rod. An overrunning clutch allows the safety control rod to freely travel toward a safe position in the event of a partial drive system failure.

Kotlyar, Oleg M. (Idaho Falls, ID); West, Phillip B. (Idaho Falls, ID)

1993-01-01T23:59:59.000Z

359

Research Program of a Super Fast Reactor  

SciTech Connect

Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is not breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)

Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie; Terai, Takayuki; Nagasaki, Shinya; Muroya, Yusa; Abe, Hiroaki [Nuclear Professional School / Department of Nuclear Engineering and Management, The University of Tokyo, Tokaimura, Naka-gun, Ibaraki, 319-1188 (Japan); Mori, Hideo [Department of Mechanical Engineering, Kyushu University (Japan); Akiba, Masato; Akimoto, Hajime; Okumura, Keisuke; Akasaka, Naoaki [Japan Atomic Energy Agency (Japan); GOTO, Shoji [Tokyo Electric Power Company (Japan)

2006-07-01T23:59:59.000Z

360

Nuclear reactor vessel fuel thermal insulating barrier  

DOE Patents (OSTI)

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

2013-03-19T23:59:59.000Z

Note: This page contains sample records for the topic "reactor generic distributed" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

History of Research Reactors at Brookhaven  

NLE Websites -- All DOE Office Websites (Extended Search)

History of Research Reactors at Brookhaven History of Research Reactors at Brookhaven Brookhaven National Laboratory has three nuclear reactors on its site that were used for scientific research. The reactors are all shut down, and the Laboratory is addressing environmental issues associated with their operations. photo of BGRR Brookhaven Graphite Research Reactor - Beginning operations in 1950, the graphite reactor was used for research in medicine, biology, chemistry, physics and nuclear engineering. One of the most significant achievements at this facility was the development of technetium-99m, a radiopharmaceutical widely used to image almost any organ in the body. The graphite reactor was shut down in 1969. Parts of it have been decommissioned, with the remainder to be addressed by 2011. More history

362

Nuclear reactor alignment plate configuration  

DOE Patents (OSTI)

An alignment plate that is attached to a core barrel of a pressurized water reactor and fits within slots within a top plate of a lower core shroud and upper core plate to maintain lateral alignment of the reactor internals. The alignment plate is connected to the core barrel through two vertically-spaced dowel pins that extend from the outside surface of the core barrel through a reinforcement pad and into corresponding holes in the alignment plate. Additionally, threaded fasteners are inserted around the perimeter of the reinforcement pad and into the alignment plate to further secure the alignment plate to the core barrel. A fillet weld also is deposited around the perimeter of the reinforcement pad. To accomodate thermal growth between the alignment plate and the core barrel, a gap is left above, below and at both sides of one of the dowel pins in the alignment plate holes through with the dowel pins pass.

Altman, David A; Forsyth, David R; Smith, Richard E; Singleton, Norman R

2014-01-28T23:59:59.000Z

363

The ARIES tokamak reactor study  

SciTech Connect

The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D{sup 3}He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions.

Not Available

1989-10-01T23:59:59.000Z

364

Nuclear power reactor education and training at the Ford nuclear reactor  

SciTech Connect

Since 1977, staff members of the University of Michigan's Ford nuclear reactor have provided courses and reactor laboratory training programs for reactor operators, engineers, and technicians from seven electric utilities, including Cleveland Electric Illuminating, Consumers Power, Detroit Edison, Indiana and Michigan Electric, Nebraska Public Power, Texas Utilities Generating Company, and Toledo Edison. Reactor laboratories, instrument technician training, and reactor physics courses have been conducted at the university. Courses conducted at plant sites include reactor physics, thermal sciences, materials sciences, and health physics and radiation protection.

Burn, R.R.

1989-01-01T23:59:59.000Z

365

Using reactor operating experience to improve the design of a new Broad Application Test Reactor  

SciTech Connect

Increasing regulatory demands and effects of plant aging are limiting the operation of existing test reactors. Additionally, these reactors have limited capacities and capabilities for supporting future testing missions. A multidisciplinary team of experts developed sets of preliminary safety requirements, facility user needs, and reactor design concepts for a new Broad Application Test Reactor (BATR). Anticipated missions for the new reactor include fuels and materials irradiation testing, isotope production, space testing, medical research, fusion testing, intense positron research, and transmutation doping. The early BATR design decisions have benefited from operating experiences with existing reactors. This paper discusses these experiences and highlights their significance for the design of a new BATR.

Fletcher, C.D.; Ryskamp, J.M.; Drexler, R.L.; Leyse, C.F.

1993-07-01T23:59:59.000Z

366

DISTRIBUTION CATEGORY  

Office of Scientific and Technical Information (OSTI)

DISTRIBUTION CATEGORY DISTRIBUTION CATEGORY uc-11 I A W E N C E LIVERMORE IABORATORY University of Cahfmia/Livermore, California/94550 UCRL-52658 CALCULATION OF CHEMICAL EQUILIBRIUM BETWEEN AQUEOUS SOLUTION AND MINERALS: THE EQ3/6 - - SOFTWARE PACKAGE T. J. Wolery MS. date: February 1, 1979 . . - . . - . Tho rcpon rn prepared as an account of work sponsored by the United Stater Government. Seither Lhc Urutcd Stater nor the Umted Stater Department of Energy, nor any of their employees. nor any of their E O ~ ~ ~ B C I O I S . rubcontracton. o r their employees. makes any warranr)., exprcs or !mplwd. or assumes any legal liability or respanability io: the ~ c c u o c y . complctencn or uvfulneu of any miormarlon. apparatcr. product or p r o m s dtwlorcd. or r c p r e v n u that its UP would not infringe privately owned r

367

Advanced Safeguards Approaches for New Fast Reactors  

SciTech Connect

This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.

Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

2007-12-15T23:59:59.000Z

368

Summary Notes from 5 March 2008 Generic Technical Issue Discussion on Long-Term Grout Performance  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

8 Page 1 of 6 8 Page 1 of 6 Summary Notes from 5 March 2008 Generic Technical Issue Discussion on Long-Term Grout Performance Attendees: Representatives from Department of Energy-Headquarters (DOE-HQ) and the U.S. Nuclear Regulatory Commission staff (NRC) met at the DOE offices in Germantown, Maryland on 5 March 2008. Representatives from Department of Energy- Savannah River (DOE-SR), Department of Energy-Idaho (DOE-ID), Department of Energy-Richland (DOE-RL), Department of Energy-River Protection (DOE-ORP), and the Center for Nuclear Waste Regulatory Analysis participated in the meeting via a teleconference link. Discussion: NRC Staff prepared and disseminated a paper summarizing issues and considerations relative to grout degradation and associated modeling issues. The purpose

369

Generic tool for modelling and simulation of semiconductor intrabay material handling system  

Science Journals Connector (OSTI)

Semiconductor manufacturing facilities are migrating to 300 mm technology, necessitating the implementation of automated material handling systems (AMHS) for a variety of ergonomic and safety considerations. A predictive tool, such as software simulation, is needed at the planning stage to estimate the performance of these relatively new systems. Two forms of AMHS are in general use in industry one which handles material within a group of machines (a bay) and one which transfers material between bays. This paper presents a generic tool for modelling and simulation of an intrabay AMHS. The model utilises a library of different blocks representing the different components of any intrabay material handling system, providing a tool that allows rapid building and analysis of an AMHS under different operating conditions. The ease of use of the system means that inexpert users have the ability to generate good models.

K.S. El-Kilany; P. Young; M.A. El Baradie

2004-01-01T23:59:59.000Z

370

Feasibility of establishing and operating a generic oil shale test facility  

SciTech Connect

The December 19, 1985, Conference Report on House Joint Resolution 465, Further continuing appropriations for Fiscal Year 1986, included instruction to DOE to conduct a feasibility study for a generic oil shale test facility. The study was completed, as directed, and its findings are documented in this report. To determine the feasibility of establishing and operating such a facility, the following approach was used: examine the nature of the resource, and establish and basic functions associated with recovery of the resource; review the history of oil shale development to help put the present discussion in perspective; describe a typical oil shale process; define the relationship between each oil shale system component (mining, retorting, upgrading, environmental) and its cost. Analyze how research could reduce costs; and determine the scope of potential research for each oil shale system component.

Not Available

1986-12-01T23:59:59.000Z

371

Efficiencies of a molecular motor: A generic hybrid model applied to the F1-ATPase  

E-Print Network (OSTI)

In a single molecule assay, the motion of a molecular motor is often inferred from measuring the stochastic trajectory of a large probe particle attached to it. We discuss a simple model for this generic set-up taking into account explicitly the elastic coupling between probe and motor. The combined dynamics consists of discrete steps of the motor and continuous Brownian motion of the probe. Motivated by recent experiments on the F1-ATPase, we investigate three types of efficiencies both in simulations and a Gaussian approximation. Overall, we obtain good quantitative agreement with the experimental data. In particular, we clarify the conditions under which one of these efficiencies becomes larger than one.

Zimmermann, Eva

2012-01-01T23:59:59.000Z

372

Efficiencies of a molecular motor: A generic hybrid model applied to the F1-ATPase  

E-Print Network (OSTI)

In a single molecule assay, the motion of a molecular motor is often inferred from measuring the stochastic trajectory of a large probe particle attached to it. We discuss a simple model for this generic set-up taking into account explicitly the elastic coupling between probe and motor. The combined dynamics consists of discrete steps of the motor and continuous Brownian motion of the probe. Motivated by recent experiments on the F1-ATPase, we investigate three types of efficiencies both in simulations and a Gaussian approximation. Overall, we obtain good quantitative agreement with the experimental data. In particular, we clarify the conditions under which one of these efficiencies becomes larger than one.

Eva Zimmermann; Udo Seifert

2012-09-17T23:59:59.000Z

373

Modified Bosonic Gas Trapped in a Generic 3-dim Power Law Potential  

E-Print Network (OSTI)

We analyze the consequences caused by an anomalous single-particle dispersion relation suggested in several quantum-gravity models, upon the thermodynamics of a Bose-Einstein condensate trapped in a generic 3-dimensional power-law potential. We prove that the condensation temperature is shifted as a consequence of such deformation and show that this fact could be used to provide bounds on the deformation parameters. Additionally, we show that the shift in the condensation temperature, described as a non-trivial function of the number of particles and the trap parameters, could be used as a criterion to analyze the effects caused by a deformed dispersion relation in weakly interacting systems and also in finite size systems.

E. Castellanos; C. Laemmerzahl

2014-02-25T23:59:59.000Z

374

Critical Temperature of an Interacting Bose Gas in a Generic Power-Law Potential  

E-Print Network (OSTI)

We investigate the critical temperature of an interacting Bose gas confined in a trap described by a generic isotropic power-law potential. We compare the results with respect to the non-interacting case. In particular, we derive an analytical formula for the shift of the critical temperature holding to first order in the scattering length. We show that this shift scales as $N^{n\\over 3(n+2)}$, where $N$ is the number of Bosons and $n$ is the exponent of the power-law potential. Moreover, the sign of the shift critically depends on the power-law exponent $n$. Finally, we find that the shift of the critical temperature due to finite-size effects vanishes as $N^{-{2n\\over 3(n+2)}}$.

Luca Salasnich

2002-04-18T23:59:59.000Z

375

Unified Link Layer API: A generic and open API to manage wireless media access  

Science Journals Connector (OSTI)

We present the Unified Link Layer API (ULLA) framework: an open and extensible API framework that incorporates a number of requirements related to a wide range of applications, including multi-mode and cross-layer optimisation scenarios. This work has been mainly motivated by the complexity and interoperability problems related to the large number of wireless \\{APIs\\} available today. ULLA provides database and object oriented service abstractions to applications through a generic query mechanism, a method to setup asynchronous notifications and a command interface. It encapsulates link level heterogeneity by defining a unified model for link technologies. We describe design details, various implementation options and discuss how the proposed ULLA design provides an extensible, scalable and platform independent framework, enabling seamless link access and control in various types of device platforms. Application programming using ULLA is illustrated using code examples. Numerous usage scenarios for ULLA are presented, highlighting unified access to heterogeneous link standards while encouraging application innovation.

Mahesh Sooriyabandara; Tim Farnham; Costas Efthymiou; Matthias Wellens; Janne Riihijärvi; Petri Mähönen; Alain Gefflaut; José Antonio Galache; Diego Melpignano; Arthur van Rooijen

2008-01-01T23:59:59.000Z

376

Experimental facility for containment sump reliability studies (Generic Task A-43). [PWR; BWR  

SciTech Connect

On July 3, 1979, Sandia National Laboratories (Sandia) contracted the Alden Research Laboratory (ARL) to conduct tests on unresolved safety issues associated with containment sump performance during the recirculation mode (Generic Task A-43). This report describes the test facility constructed and completed under Phase I, Task III of the contract. Sump performance is determined through the observation of vortex formation in the main tank and the measurement of swirl, pressure gradient, and entrained air in the suction pipes. The use of electrically operated valves and a sophisticated data acquisition system, with computer interface, allows the test flow parameters to be set and test data to be taken (with the exception of vortex observations) from a single central office.

Durgin, W. W.; Padmanabhan, M.; Janik, C. R.

1980-12-01T23:59:59.000Z

377

Fabrication Flaws in Reactor Pressure Vessel Repair Welds  

SciTech Connect

This paper describes the fabrication flaw distribution and characterization in the repair weld metal of reactor pressure vessels. This work indicates that the large flaws occur in these repairs. These results show that repair flaws are complex in composition and sometimes include cracks on the repair ends. Parametric analysis using an exponential fit is performed on the data. A description of repair flaw morphology is provided. Fabrication flaws in repairs are characterized using high sensitivity nondestructive ultrasonic testing, validation by other nondestructive evaluation (NDE) techniques, and complemented by destructive testing.

Schuster, George J.; Doctor, Steven R.

2007-12-01T23:59:59.000Z

378

Articulated limiter blade for a tokamak fusion reactor  

DOE Patents (OSTI)

A limiter blade for a large tokomak fusion reactor includes three articulated blade sections for enabling the limiter blade to be adjusted for plasmas of different sizes. Each blade section is formed of a rigid backing plate carrying graphite tiles coated with titanium carbide, and the limiter blade forms a generally elliptic contour in both the poloidal and toroidal directions to uniformly distribute the heat flow to the blade. The limiter blade includes a central blade section movable along the major radius of the vacuum vessel, and upper and lower pivotal blade sections which may be pivoted by linear actuators having rollers held to the back surface of the pivotal blade sections.

Doll, D.W.

1982-10-21T23:59:59.000Z

379

Progress in evaluation and improvement in nondestructive examination reliability for inservice inspection of Light Water Reactors (LWRs) and characterize fabrication flaws in reactor pressure vessels  

SciTech Connect

This paper is a review of the work conducted under two programs. One (NDE Reliability Program) is a multi-year program addressing the reliability of nondestructive evaluation (NDE) for the inservice inspection (ISI) of light water reactor components. This program examines the reliability of current NDE, the effectiveness of evolving technologies, and provides assessments and recommendations to ensure that the NDE is applied at the right time, in the right place with sufficient effectiveness that defects of importance to structural integrity will be reliably detected and accurately characterized. The second program (Characterizing Fabrication Flaws in Reactor Pressure Vessels) is assembling a data base to quantify the distribution of fabrication flaws that exist in US nuclear reactor pressure vessels with respect to density, size, type, and location. These programs will be discussed as two separate sections in this report. 4 refs., 7 figs.

Doctor, S.R.; Bowey, R.E.; Good, M.S.; Friley, J.R.; Kurtz, R.J.; Simonen, F.A.; Taylor, T.T.; Heasler, P.G.; Andersen, E.S.; Diaz, A.A.; Greenwood, M.S.; Hockey, R.L.; Schuster, G.J.; Spanner, J.C.; Vo, T.V.

1991-10-01T23:59:59.000Z

380

E-Print Network 3.0 - argonne fast source reactor Sample Search...  

NLE Websites -- All DOE Office Websites (Extended Search)

of the Omega Reactor Facility, Summary: fission. The benefits of a fast reactor over the water boiler reactor were a high intensity source offast... Reactors at Other Locations...

Note: This page contains sample records for the topic "reactor generic distributed" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


381

X-10 Graphite Reactor | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

X-10 Graphite Reactor X-10 Graphite Reactor X-10 Graphite Reactor X-10 Graphite Reactor When President Roosevelt in December 1942 authorized the Manhattan Project, the Oak Ridge site in eastern Tennessee had already been obtained and plans laid for an air-cooled experimental pile, a pilot chemical separation plant, and support facilities. The X-10 Graphite Reactor, designed and built in ten months, went into operation on November 4, 1943. The X-10 used neutrons emitted in the fission of uranium-235 to convert uranium-238 into a new element, plutonium-239. The reactor consists of a huge block of graphite, measuring 24 feet on each side, surrounded by several feet of high-density concrete as a radiation shield. The block is pierced by 1,248 horizontal diamond-shaped channels in

382

Ordered bed modular reactor design proposal  

SciTech Connect

The Ordered Bed Modular Reactor (OBMR) is a design as an advanced modular HTGR in which the annular reactor core is filled with an ordered bed of fuel spheres. This arrangement allows fuel elements to be poured into the core cavity which is shaped so that an ordered bed is formed and to be discharged from the core through the opening holes in the reactor top. These operations can be performed in a shutdown shorter time. The OBMR has the most of advantages from both the pebble bed reactor and block type reactor. Its core has great structural flexibility and stability, which allow increasing reactor output power and outlet gas temperature as well as decreasing core pressure drop. This paper introduces ordered packing bed characteristics, unloading and loading technique of the fuel spheres and predicted design features of the OBMR. (authors)

Tian, J. [Inst. of Nuclear Energy Technology, Tsinghua Univ., Beijing 100084 (China)

2006-07-01T23:59:59.000Z

383

Nuclear Reactor Technologies | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Reactor Technologies Reactor Technologies Nuclear Reactor Technologies TVA Watts Bar Nuclear Power Plant | Photo courtesy of Tennessee Valley Authority TVA Watts Bar Nuclear Power Plant | Photo courtesy of Tennessee Valley Authority Nuclear power has reliably and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Small Modular Reactor Technologies Small modular reactors can also be made in factories and transported to sites where they would be ready to "plug and play" upon arrival, reducing both capital costs and construction times. The smaller size also makes these reactors ideal for small electric grids and for locations that

384

Reactors: Modern-Day Alchemy - Argonne's Nuclear Science and Technology  

NLE Websites -- All DOE Office Websites (Extended Search)

Achievements > Achievements > Legacy > Reactors: Modern-Day Alchemy About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy

385

Achievements: Nuclear Reactors designed/built by Argonne National  

NLE Websites -- All DOE Office Websites (Extended Search)

Achievements > Achievements > Argonne National Laboratory Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy

386

Ignition reactor and pump pulse parameters in a reactor–laser system  

Science Journals Connector (OSTI)

The experience gained in operating a demonstration nuclear-pumped laser in stand B (Physics and Power- Engineering Institute (FEI)) with a pulsed ignition reactor based on the 235U BARS-6 reactor is analyzed. It ...

P. P. D’yachenko; G. N. Fokin

2012-09-01T23:59:59.000Z

387

Solid tags for identifying failed reactor components  

DOE Patents (OSTI)

A solid tag material which generates stable detectable, identifiable, and measurable isotopic gases on exposure to a neutron flux to be placed in a nuclear reactor component, particularly a fuel element, in order to identify the reactor component in event of its failure. Several tag materials consisting of salts which generate a multiplicity of gaseous isotopes in predetermined ratios are used to identify different reactor components.

Bunch, Wilbur L. (Richland, WA); Schenter, Robert E. (Richland, WA)

1987-01-01T23:59:59.000Z

388

Argonne step closer to safer nuclear reactor  

Science Journals Connector (OSTI)

Argonne step closer to safer nuclear reactor ... "A key technological link" toward development of meltdown-immune nuclear reactors is now in the demonstration phase at Argonne National Laboratory near Chicago. ... The technique is part of Argonne's continuing interest in the sodium-cooled integral fast reactor (IFR), whose immunity to meltdown derives from molten sodium's function as a heat sink and the use of metallic fuel that conducts heat better than conventional oxide fuels. ...

WARD WORTHY

1988-05-30T23:59:59.000Z

389

Small Reactor for Deep Space Exploration  

ScienceCinema (OSTI)

This is the first demonstration of a space nuclear reactor system to produce electricity in the United States since 1965, and an experiment demonstrated the first use of a heat pipe to cool a small nuclear reactor and then harvest the heat to power a Stirling engine at the Nevada National Security Site's Device Assembly Facility confirms basic nuclear reactor physics and heat transfer for a simple, reliable space power system.

None

2014-05-30T23:59:59.000Z

390

Fuel handling apparatus for a nuclear reactor  

DOE Patents (OSTI)

Fuel handling apparatus for transporting fuel elements into and out of a nuclear reactor and transporting them within the reactor vessel extends through a penetration in the side of the reactor vessel. A lateral transport device carries the fuel elements laterally within the vessel and through the opening in the side of the vessel, and a reversible lifting device raises and lowers the fuel elements. In the preferred embodiment, the lifting device is supported by a pair of pivot arms.

Hawke, Basil C. (Solana Beach, CA)

1987-01-01T23:59:59.000Z

391

Small Reactor for Deep Space Exploration  

SciTech Connect

This is the first demonstration of a space nuclear reactor system to produce electricity in the United States since 1965, and an experiment demonstrated the first use of a heat pipe to cool a small nuclear reactor and then harvest the heat to power a Stirling engine at the Nevada National Security Site's Device Assembly Facility confirms basic nuclear reactor physics and heat transfer for a simple, reliable space power system.

None

2012-11-29T23:59:59.000Z

392

Neutron shielding panels for reactor pressure vessels  

DOE Patents (OSTI)

In a nuclear reactor neutron panels varying in thickness in the circumferential direction are disposed at spaced circumferential locations around the reactor core so that the greatest radial thickness is at the point of highest fluence with lesser thicknesses at adjacent locations where the fluence level is lower. The neutron panels are disposed between the core barrel and the interior of the reactor vessel to maintain radiation exposure to the vessel within acceptable limits.

Singleton, Norman R. (Murrysville, PA)

2011-11-22T23:59:59.000Z

393

Percent Distribution  

Gasoline and Diesel Fuel Update (EIA)

. . Percent Distribution of Natural Gas Supply and Disposition by State, 1996 Table State Estimated Proved Reserves (dry) Marketed Production Total Consumption Alabama................................................................... 3.02 2.69 1.48 Alaska ...................................................................... 5.58 2.43 2.04 Arizona..................................................................... NA 0 0.55 Arkansas.................................................................. 0.88 1.12 1.23 California.................................................................. 1.25 1.45 8.23 Colorado .................................................................. 4.63 2.90 1.40 Connecticut.............................................................. 0 0 0.58 D.C...........................................................................

394

Distributed Generation  

NLE Websites -- All DOE Office Websites (Extended Search)

Untapped Value of Backup Generation Untapped Value of Backup Generation While new guidelines and regulations such as IEEE (Institute of Electrical and Electronics Engineers) 1547 have come a long way in addressing interconnection standards for distributed generation, utilities have largely overlooked the untapped potential of these resources. Under certain conditions, these units (primarily backup generators) represent a significant source of power that can deliver utility services at lower costs than traditional centralized solutions. These backup generators exist today in large numbers and provide utilities with another option to reduce peak load, relieve transmission congestion, and improve power reliability. Backup generation is widely deployed across the United States. Carnegie Mellon's Electricity

395

VVER-440 and VVER-1000 reactor dosimetry benchmark - BUGLE-96 versus ALPAN VII.0  

SciTech Connect

Document available in abstract form only, full text of document follows: Analytical results of the vodo-vodyanoi energetichesky reactor-(VVER-) 440 and VVER-1000 reactor dosimetry benchmarks developed from engineering mockups at the Nuclear Research Inst. Rez LR-0 reactor are discussed. These benchmarks provide accurate determination of radiation field parameters in the vicinity and over the thickness of the reactor pressure vessel. Measurements are compared to calculated results with two sets of tools: TORT discrete ordinates code and BUGLE-96 cross-section library versus the newly Westinghouse-developed RAPTOR-M3G and ALPAN VII.0. The parallel code RAPTOR-M3G enables detailed neutron distributions in energy and space in reduced computational time. ALPAN VII.0 cross-section library is based on ENDF/B-VII.0 and is designed for reactor dosimetry applications. It uses a unique broad group structure to enhance resolution in thermal-neutron-energy range compared to other analogous libraries. The comparison of fast neutron (E > 0.5 MeV) results shows good agreement (within 10%) between BUGLE-96 and ALPAN VII.O libraries. Furthermore, the results compare well with analogous results of participants of the REDOS program (2005). Finally, the analytical results for fast neutrons agree within 15% with the measurements, for most locations in all three mockups. In general, however, the analytical results underestimate the attenuation through the reactor pressure vessel thickness compared to the measurements. (authors)

Duo, J. I. [Radiation Engineering and Analysis, Westinghouse Electric Company LLC, 1000 Westinghouse Dr., Cranberry Township, PA 16066 (United States)

2011-07-01T23:59:59.000Z

396

Uranium redistribution under oxidizing conditions in Oklo natural reactor zone 2, Gabon  

SciTech Connect

This mineralogical study was completed to elucidate the relationships between uranium distribution and alteration products of the host rock of natural reactor zone clays just below the reactor core. Uraninite is preserved without any alteration in the reactor core. Uranium minerals are found to be present in the fractures in the reactor zone clays associated with iron-mineral veins, galena and Ti-bearing minerals. Uranium, for which the phases could not be identified, occurs in iron-mineral veins and the iron-mineral rim of pyrite grains in the reactor zone clays. Uranium is not associated with granular iron minerals occurring in the illite matrix of the reactor zone clays. The degree of crystallinity and uranium content of the three iron-bearing alteration products suggest that they formed under different conditions; the granular iron minerals, under alteration conditions where uranium was not mobilized while the iron-mineral veins and the iron-mineral rim of pyrite, under conditions in which uranium is mobilized after the formation of the granular iron minerals.

Isobe, H.; Ohnuki, T. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan); Murakami, T. [Ehime Univ., Matsuyama, Ehime (Japan); Gauthier-Lafaye, F. [CNRS, Strasbourg (France). Centre de Geochemie de la Surface

1995-12-31T23:59:59.000Z

397

An attempt to minimize the temperature gradient along a plug-flow methane/steam reforming reactor by adopting locally controlled heating zones  

Science Journals Connector (OSTI)

Plug flow reactors are very common in the chemical process industry, including methane/steam reforming applications. Their operation presents many challenges, such as a strong dependence of temperature and composition distribution on the inlet conditions. The strongly endothermic methane/steam reforming reaction might result in a temperature drop at the inlet of the reactor and consequently the occurrence of large temperature gradients. The strongly non-uniform temperature distribution due to endothermic chemical reaction can have tremendous consequences on the operation of the reactor, such as catalyst degradation, undesired side reactions and thermal stresses. To avoid such unfavorable conditions, thermal management of the reactor becomes an important issue. To carry out thermal management properly, detailed modeling and corresponding numerical analyses of the phenomena occurring inside the reforming system is required. This paper presents experimental and numerical studies on the methane/steam reforming process inside a plug-flow reactor. To optimize the reforming reactors, detailed data about the entire reforming process is required. In this study the kinetics of methane/steam reforming on the Ni/YSZ catalyst was experimentally investigated. Measurements including different thermal boundary conditions, the fuel flow rate and the steam- to-methane ratios were performed. The reforming rate equation derived from experimental data was used in the numerical model to predict gas composition and temperature distribution along the steam-reforming reactor. Finally, an attempt was made to control the temperature distribution by adopting locally controlled heating zones.

M Mozdzierz; G Brus; A Sciazko; Y Komatsu; S Kimijima; J S Szmyd

2014-01-01T23:59:59.000Z

398

Nuclear Energy Enabling Technologies (NEET) Reactor Materials  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Enabling Technologies (NEET) Reactor Materials Enabling Technologies (NEET) Reactor Materials Award Recipient Estimated Award Amount* Award Location Supporting Organizations Project Description University of Nebraska $979,978 Lincoln, NE Massachusetts Institute of Technology (Cambridge, MA), Texas A&M (College Station, TX) Project will explore the development of advanced metal/ceramic composites. These improvements could lead to more efficient production of electricity in advanced reactors. Oak Ridge National Laboratory $849,000 Oak Ridge, TN University of Wisconsin-Madison (Madison, WI) Project will develop novel high-temperature high-strength steels with the help of computational modeling, which could lead to increased efficiency in advanced reactors. Pacific Northwest National Laboratory

399

Accelerators for Subcritical Molten-Salt Reactors  

SciTech Connect

Accelerator parameters for subcritical reactors have usually been based on using solid nuclear fuel much like that used in all operating critical reactors as well as the thorium burning accelerator-driven energy amplifier proposed by Rubbia et al. An attractive alternative reactor design that used molten salt fuel was experimentally studied at ORNL in the 1960s, where a critical molten salt reactor was successfully operated using enriched U235 or U233 tetrafluoride fuels. These experiments give confidence that an accelerator-driven subcritical molten salt reactor will work better than conventional reactors, having better efficiency due to their higher operating temperature, having the inherent safety of subcritical operation, and having constant purging of volatile radioactive elements to eliminate their accumulation and potential accidental release in dangerous amounts. Moreover, the requirements to drive a molten salt reactor can be considerably relaxed compared to a solid fuel reactor, especially regarding accelerator reliability and spallation neutron targetry, to the point that much of the required technology exists today. It is proposed that Project-X be developed into a prototype commercial machine to produce energy for the world by, for example, burning thorium in India and nuclear waste from conventional reactors in the USA.

Johnson, Roland (Muons, Inc.) [Muons, Inc.

2011-08-03T23:59:59.000Z

400

Subcritical Fission Reactor Based on Linear Collider  

E-Print Network (OSTI)

The beams of Linear Collider after main collision can be utilized to build an accelerator--driven sub--critical reactor.

I. F. Ginzburg

2005-07-29T23:59:59.000Z

Note: This page contains sample records for the topic "reactor generic distributed" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


401

Italian hybrid and fission reactors scenario analysis  

SciTech Connect

Italy is a country where a long tradition of studies both in the fission and fusion field is consolidated; nevertheless a strong public opinion concerned with the destination of the Spent Nuclear Fuel hinders the development of nuclear power. The possibility to a severe reduction of the NSF mass generated from a fleet of nuclear reactors employing an hypothetical fusionfission hybrid reactor has been investigated in the Italian framework. The possibility to produce nuclear fuel for the fission nuclear reactors with the hybrid reactor was analyzed too.

Ciotti, M.; Manzano, J.; Sepielli, M. [ENEA CR Frascati, Via Enrico Fermi, 45, 00044, Frascati, Roma (Italy); ENEA CR casaccia, Via Anguillarese, 301, 00123, Santa Maria di Galeria, Roma (Italy)

2012-06-19T23:59:59.000Z

402

Nuclear reactor multiphysics via bond graph formalism  

E-Print Network (OSTI)

This work proposes a simple and effective approach to modeling nuclear reactor multiphysics problems using bond graphs. Conventional multiphysics simulation paradigms normally use operator splitting, which treats the ...

Sosnovsky, Eugeny

2014-01-01T23:59:59.000Z

403

Recovery Act Progress Update: Reactor Closure Feature  

ScienceCinema (OSTI)

A Recovery Act Progress Update. Decommissioning of two nuclear reactor sites at the Department of Energy's facilities has been approved and is underway.

Cody, Tom

2012-06-14T23:59:59.000Z

404

Computational evaluation of two reactor benchmark problems.  

E-Print Network (OSTI)

??A neutronic evaluation of two reactor benchmark problems was performed. The benchmark problems describe typical PWR uranium and plutonium (mixed oxide) fueled lattices. WIMSd4m, a… (more)

Cowan, James Anthony

2012-01-01T23:59:59.000Z

405

Heat dissipating nuclear reactor with metal liner  

DOE Patents (OSTI)

Disclosed is a nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

Gluekler, Emil L. (San Jose, CA); Hunsbedt, Anstein (Los Gatos, CA); Lazarus, Jonathan D. (Sunnyvale, CA)

1987-01-01T23:59:59.000Z

406

Hallam, Nebraska, Decommissioned Reactor Site Fact Sheet  

Office of Legacy Management (LM)

Program. Objectives for the reactor were fulfilled by 1966, and the Nebraska Public Power District decommissioned and dismantled the facility between 1967 and 1969. Facility...

407

Tanden Mirror Reactor Systems Code (TMRSC)  

SciTech Connect

This paper describes a computer code developed to model a tandem mirror reactor. This is the first tandem mirror reactor model to couple the highly linked physics, magnetics, and neutronic analysis into a single code. Results from this code for two sensitivity studies are included in this paper. These studies are designed (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power and (2) to determine the impact of reactor power level on cost.

Reid, R.L.; Rothe, K.E.; Barrett, R.J.

1985-01-01T23:59:59.000Z

408

Light Water Reactor Sustainability Program Contact Information  

NLE Websites -- All DOE Office Websites (Extended Search)

Program Organization LWRS Program Management Richard Reister Federal Project Director Light Water Reactor Deployment Office of Nuclear Energy U.S. Department of Energy...

409

Green's Functions for Surface Waves in a Generic Velocity Structure by Victor C. Tsai and Sarun Atiganyanun*  

E-Print Network (OSTI)

Short Note Green's Functions for Surface Waves in a Generic Velocity Structure by Victor C. Tsai displacement/stress eigenfunctions and Green's functions have been well established for many decades. However on frequency, or Green's function surface displacement. We address this gap in the liter- ature and here

410

Video Primal Sketch: A Generic Middle-Level Representation of Video Institute for Information and System Sciences  

E-Print Network (OSTI)

Video Primal Sketch: A Generic Middle-Level Representation of Video Zhi Han Zongben Xu Institute This paper presents a middle-level video representation named Video Primal Sketch (VPS), which integrates two. This paper makes three contributions: i) learning a dictionary of video primitives as parametric generative

Zhu, Song Chun

411

Solution Ionic Strength Engineering As a Generic Strategy to Coat Graphene Oxide (GO) on Various Functional Particles and Its  

E-Print Network (OSTI)

in improving the properties of particle materials. KEYWORDS: Graphene oxide, sulfur, lithium-sulfur batteriesSolution Ionic Strength Engineering As a Generic Strategy to Coat Graphene Oxide (GO) on Various Functional Particles and Its Application in High-Performance Lithium-Sulfur (Li-S) Batteries Jiepeng Rong

Zhou, Chongwu

412

A Generic Space-Time-Frequency Correlation Model and Its Corresponding Simulation Model for Narrowband MIMO Channels  

E-Print Network (OSTI)

A Generic Space-Time-Frequency Correlation Model and Its Corresponding Simulation Model.Laurenson d9ed. ac. uk Keyword: Wireless channels, NIIMO, STF correlation, deterministic simulation model theoretical reference model, a deterministic simulation model is then proposed and its 3-D STF correlation

Wang, Cheng-Xiang

413

Thermal stabilization of chemical reactors. I The mathematical description of the Endex reactor  

Science Journals Connector (OSTI)

...efficiently by steam generation. Conversely...of fossil or nuclear fuels, which...limits of the reactor. The physico...wasted. The Endex reactor can be thought...conventional steam generation that is currently...Rates of heat generation by reaction...functions of reactor temperature...

1999-01-01T23:59:59.000Z

414

NEW OPTIMIZATION-BASED APPROACH TO CHEMICAL REACTOR SYNTHESIS TOWARDS THE FULL INTEGRATION OF REACTOR  

E-Print Network (OSTI)

NEW OPTIMIZATION-BASED APPROACH TO CHEMICAL REACTOR SYNTHESIS ­ TOWARDS THE FULL INTEGRATION solutions. However, it does not provide optimal reactor design from both economical and environmental and methods for reactor design. It also explores the possibilities for actuation improvement for the optimal

Van den Hof, Paul

415

Analysis of ITU TRIGA Mark II research reactor using Monte Carlo method  

Science Journals Connector (OSTI)

Abstract Research reactors include many complicated components with various shapes and sizes. Such complex parts also observed in TRIGA core are modelled by the researchers as simplified physical geometries when a particle transport computer code is used to analyse the reactors. These models are used to gain information on possible modifications in the reactors with no cost except a certain computational time demand. Besides, they can be used to understand the fabrication uncertainties of the core components and the methodologies used in the design process. The main objective of this study is to make a detailed three-dimensional full-core model of ITU (Istanbul Technical University) TRIGA Mark II research reactor for the use of Monte Carlo method and making a comparison of the simulation with the experimental observations. In case of lacking of experimental values reported, Final Safety Analysis Report values are used as reference. Furthermore, it is aimed to observe possible influences of using various neutron cross-section libraries (ENDF/Bs and JEFFs) onto the simulation results. The Monte Carlo simulations are carried out by using MCNP5 radiation transport code. All unsteady conditions are ignored, assuming the reactor operates at cold-zero power under the steady-state condition. For comparison, effective core multiplication factor (keff) and effective delayed neutron fraction (?eff) are computed. Reactivity worth ($) of control rods with rod position is presented. Pin power distribution within the fuel elements, axial power peaking distribution along the fuel length and normalized distribution of fast/thermal neutron flux throughout the core are analysed. The simulation results show that MCNP5 model of the reactor is properly established with sufficient detail in such a way that all simulation results are in an excellent agreement with the experimental data (or FSAR values). Results also show that the model yields more or less the same value even different neutron libraries are used.

Mehmet Türkmen; Üner Çolak

2014-01-01T23:59:59.000Z

416

Killing the Golden Goose or Just Chasing it Around the Farmyard?: Generic Entry and the Incentives for Early-Stage Pharmaceutical Innovation  

E-Print Network (OSTI)

Innovation ABSTRACT Over the last decade, generic penetration in the U.S. pharmaceutical market has increased substantially, providing significant consumer surplus gains. What impact has this rise in generic penetration had on the rate and direction of early stage pharmaceutical innovation? We explore this question using

Sekhon, Jasjeet S.

417

The performance of ENDF/B-V. 2 nuclear data for fast reactor calculations  

SciTech Connect

Calculations with ENDF/B-V.2 data have been made for twenty-five fast-spectrum integral assemblies covering a wide range of sizes and compositions. Analysis was done by transport codes with refined cross section processing methods and detailed reactor modelling. The predictions of fission rate distributions and control rod worths were emphasized for the more prototypic benchmark cores. The results show considerable improvements in agreement with experiment compared with analysis using ENDF/B-IV data, but it is apparent that significant errors remain for fast reactor design calculations.

Atkinson, C.A.; Collins, P.J.

1987-01-01T23:59:59.000Z

418

Modal approximation of xenon oscillations in nuclear reactors — an FEM-based approach  

Science Journals Connector (OSTI)

A modal expansion method for modelling dynamic space-dependent effects (xenon oscillations) in large nuclear reactors is presented. Based on the finite element method (FEM), modal expansion functions for various cases with practical significance can be calculated. In this way it is possible to derive a low order state space model describing the considered effect. The use of this model in on-line controller design and simulation is intended. Modal methods in connection with modern numerical algorithms turned out to be effective to treat distributed parameter systems governed by elliptic differential equations. This will be demonstrated by application to a VVER-1000 pressurized water reactor.

J.Chr. Kalkkuhl; M.G. Döring

1993-01-01T23:59:59.000Z

419

Secondary heat exchanger design and comparison for advanced high temperature reactor  

SciTech Connect

Next generation nuclear reactors such as the advanced high temperature reactor (AHTR) are designed to increase energy efficiency in the production of electricity and provide high temperature heat for industrial processes. The efficient transfer of energy for industrial applications depends on the ability to incorporate effective heat exchangers between the nuclear heat transport system and the industrial process heat transport system. This study considers two different types of heat exchangers - helical coiled heat exchanger and printed circuit heat exchanger - as possible options for the AHTR secondary heat exchangers with distributed load analysis and comparison. Comparison is provided for all different cases along with challenges and recommendations. (authors)

Sabharwall, P. [Idaho National Laboratory, Idaho Falls, ID 83415-3860 (United States); Kim, E. S. [Seoul National Univ., P.O. Box 1625, Idaho Falls, ID 83415-3860 (United States); Siahpush, A.; McKellar, M.; Patterson, M. [Idaho National Laboratory, Idaho Falls, ID 83415-3860 (United States)

2012-07-01T23:59:59.000Z

420

Percent Distribution  

Gasoline and Diesel Fuel Update (EIA)

. . Percent Distribution of Natural Gas Delivered to Consumers by State, 1996 Table State Residential Commercial Industrial Vehicle Fuel Electric Utilities Alabama..................................... 1.08 0.92 2.27 0.08 0.23 Alaska ........................................ 0.31 0.87 0.85 - 1.16 Arizona....................................... 0.53 0.92 0.30 3.91 0.70 Arkansas.................................... 0.88 0.98 1.59 0.11 1.24 California.................................... 9.03 7.44 7.82 43.11 11.64 Colorado .................................... 2.12 2.18 0.94 0.58 0.20 Connecticut................................ 0.84 1.26 0.37 1.08 0.38 D.C............................................. 0.33 0.52 - 0.21 - Delaware.................................... 0.19 0.21 0.16 0.04 0.86 Florida........................................

Note: This page contains sample records for the topic "reactor generic distributed" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


421

Distribution Category:  

Office of Legacy Management (LM)

- - Distribution Category: Remedial Action and Decommissioning Program (UC-70A) DOE/EV-0005/48 ANL-OHS/HP-84-104 ARGONNE NATIONAL LABORATORY 9700 South Cass Avenue Argonne, Illinois 60439 FORMERLY UTILIZED MXD/AEC SITES REMEDIAL ACTION PROGRAM RADIOLOGICAL SURVEY OF THE HARSHAW CHEMICAL COMPANY CLEVELAND. OHIO Prepared by R. A. Wynveen Associate Division Director, OHS W. H. Smith Senior Health Physicist C. M. Sholeen Health Physicist A. L. Justus Health Physicist K. F. Flynn Health Physicist Radiological Survey Group Health Physics Section Occupational Health and Safety Division April 1984 Work Performed under Budget Activity DOE KN-03-60-40 and ANL 73706 iii PREFACE AND EXECUTIVE SUMMARY This is one in a series of reports resulting from a program initiated

422

Simulated nuclear reactor fuel assembly  

DOE Patents (OSTI)

An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

Berta, V.T.

1993-04-06T23:59:59.000Z

423

Compound cryopump for fusion reactors  

E-Print Network (OSTI)

We reconsider an old idea: a three-stage compound cryopump for use in fusion reactors such as DEMO. The helium "ash" is adsorbed on a 4.5 K charcoal-coated surface, while deuterium and tritium are adsorbed at 15-22 K on a second charcoal-coated surface. The helium is released by raising the first surface to ~30 K. In a separate regeneration step, deuterium and tritium are released at ~110 K. In this way, the helium can be pre-separated from other species. In the simplest design, all three stages are in the same vessel, with a single valve to close the pump off from the tokamak during regeneration. In an alternative design, the three stages are in separate vessels, connected by valves, allowing the stages to regenerate without interfering with each other. The inclusion of the intermediate stage would not affect the overall pumping speed significantly. The downstream exhaust processing system could be scaled down, as much of the deuterium and tritium could be returned directly to the reactor. This could reduce ...

Kovari, M; Shephard, T

2013-01-01T23:59:59.000Z

424

The application of Plant Reliability Data Information System (PRINS) to CANDU reactor  

SciTech Connect

As risk-informed applications (RIAs) are actively implanted in the nuclear industry, an issue associated with technical adequacy of Probabilistic Safety Assessment (PSA) arises in its modeling and data sourcing. In Korea, PSA for all Korean NPPs has been completed and KHNP(Korea Hydro and Nuclear Power Plant Company) developed the database called the Plant Reliability Data Information System (PRinS). It has several characteristics that distinguish it from other database system such as NPRDs (INPO,1994), PRIS (IAEA), and SRDF (EdF). This database has the function of systematic data management such as automatic data-gathering, periodic data deposition and updating, statistical analysis including Bayesian method, and trend analysis of failure rate or unavailability. In recent PSA for CANDU reactor, the component failure data of EPRI ALWR URD and Component Reliability Database were preferentially used as generic data set. The error factor for most component failure data was estimated by using the information NUREG/CR-4550 and NUREG/CR-4639. Also, annual trend analysis was performed for the functional losses of components using the statistical analysis and chart module of PRinS. Furthermore, the database has been updated regularly and maintained as a living program to reflect the current status. This paper presents the failure data analysis using PRinS which provides Bayesian analysis on main components in the CANDU reactor. (authors)

Hwang, S. W.; Lim, Y. H.; Park, H. C. [Korea Hydro and Nuclear Power Co., Ltd., Naah-ri 260, Yangnam-myun, Gyeongju-si, Gyeong Buk (Korea, Republic of)

2012-07-01T23:59:59.000Z

425

Status of the RERTR (Reduced Enrichment Research and Test Reactor) Program  

SciTech Connect

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1987, the major events, findings and activities of 1988 are reviewed. The US Nuclear Regulatory Commission issued a formal and generic approval of the use of U3Si2-Al dispersion fuel in research and test reactors, with densities up to 4.8 g U/cmT. New significant findings from postirradiation examinations, from ion-beam irradiations, and from analytical modeling, have raised serious doubts about the potential of LEU U3Si-Al dispersion fuel for applications requiring very high uranium densities and high burnups (>6 g U/cmT, >50% burnup). As a result of these findings, the fuel development efforts have been redirected towards three new initiatives: (1) a systematic application of ion-beam irradiations to screen new materials; (2) application of Hot Isostatic Pressing (HIP) procedures to produce U3Si2-Al plates with high uranium densities and thin uniform cladding; and (3) application of HIP procedures to produce plates with U3Si wires imbedded in an aluminum matrix, achieving stability, high uranium density, and thin uniform cladding. The new fuel concepts hold the promise of extraordinary performance potential and require approximately five years to develop.

Travelli, A.

1988-01-01T23:59:59.000Z

426

Seismic probabilistic risk assessment for K Reactor at the DOE Savannah River Site  

SciTech Connect

The evaluation of the risk of core melt for the K Reactor indicated an improvement in the strength of the plant as the result of analyses and structural upgrades done for the reactor to satisfy requirements of the NRC USI A-46 unresolved safety issue. The evaluation for operator errors used the THERP methodology and applied this method to conditions required as the result of earthquakes. The evaluation indicated the strength of structures and equipment was stronger after the walk down and analyses were completed than before, although physical changes to the system were minimal. The net result of the re-analysis indicated the overall yearly risk of core melt frequency from earthquakes decreased from 1.2 {times} 10{sup {minus}4} per year to 7.5 {times}10{sup {minus}5}. The uncertainty analysis indicated the 95 percentile value was 1.2 {times} 10{sup {minus}4}, and the five percentile value was 5.5 {times} 10{sup {minus}5}. The seismic hazard for the site used an evaluation of the Lawrence Livermore (LLNL) and the Electric Power Research Institute (EPRI) Hazard analysis for the plant site, with generic soils factors from the EPRI study. This hazard was determined from the two studies by comparing the predicted hazards with the historical record, and deleting those predictions that did not come close to the historical record.

Wingo, H.E. [Westinghouse Savannah River Co., Aiken, SC (United States); McCann, M.W. [Benjamin (Jack R.) and Associates, Inc., Mountain View, CA (United States)

1993-09-01T23:59:59.000Z

427

A probabilistic safety analysis of incidents in nuclear research reactors  

Science Journals Connector (OSTI)

......System for Research Reactor (IRSRR). Available...System for Research Reactor (IRSRR). Available...76. 7 Manual on reliability data collection for research reactor PSAs. (1992) IAEA...probabilistic safety analysis of incidents in nuclear......

Valdir Maciel Lopes; Gian Maria Agostinho Angelo Sordi; Mauricio Moralles; Tufic Madi Filho

2012-06-01T23:59:59.000Z

428

Natural Gas Transmission and Distribution Module  

Gasoline and Diesel Fuel Update (EIA)

5, DOE/EIA-M062(2005) (Washington, DC, 2005). 5, DOE/EIA-M062(2005) (Washington, DC, 2005). Energy Information Administration/Assumptions to the Annual Energy Outlook 2006 101 Primary Flows Secondary Flows Pipeline Border Crossing Specific LNG Terminals Primary Flows Secondary Flows Pipeline Border Crossing Specific LNG Terminals Generic LNG Terminals Alaska Alaska MacKenzie W. Canada E. Canada Canada Offshore & LNG Pacific (9) Mountain (8) CA (12) AZ/NM (11) W. South Central (7) E. South Central (6) W. North Central (4) E. North Central (3) Mid Atlantic (2) New Engl. (1) S. Atlantic (5) FL (10) Bahamas Mexico Figure 8. Natural Gas Transmission and Distribution Model Regions Source: Energy Information Administration, Office of Integrated Analysis and Forecasting Report #:DOE/EIA-0554(2006) Release date: March 2006 Next release date: March 2007

429

A next-generation reactor concept: The Integral Fast Reactor (IFR)  

SciTech Connect

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

Chang, Y.I.

1992-01-01T23:59:59.000Z

430

A next-generation reactor concept: The Integral Fast Reactor (IFR)  

SciTech Connect

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

Chang, Y.I.

1992-07-01T23:59:59.000Z

431

Fabrication Flaw Density and Distribution in Piping Weldments  

SciTech Connect

The U.S. Nuclear Regulatory Commission supported the Pacific Northwest National Laboratory (PNNL) to develop empirical data on the density and distribution of fabrication flaws in nuclear reactor components. These data are needed to support probabilistic fracture mechanics calculations and studies on component structural integrity. PNNL performed nondestructive examination inspections and destructive testing on archived piping welds to determine the fabrication flaw size and distribution characteristics of the flaws in nuclear power plant piping weldments. Eight different processes and product forms in piping weldments were studied including wrought stainless steel and dissimilar metal weldments. Parametric analysis using an exponential fit was performed on the data. Results were created as a function of the through-wall size of the fabrication flaws as well as the length distribution. The results are compared and contrasted with those developed for reactor pressure vessel processes and product forms. The most significant findings were that the density of fabrication flaws versus through-wall size was higher in piping weldments than that for the reactor pressure vessel weldments, and the density of fabrication flaws versus through-wall size in both reactor pressure vessel weld repairs and piping weldments were greater than the density in the original weldments. Curves showing these distributions are presented.

Doctor, Steven R.

2009-09-01T23:59:59.000Z

432

University of Virginia Reactor Facility Decommissioning Results  

SciTech Connect

The University of Virginia Reactor Facility started accelerated decommissioning in 2002. The facility consists of two licensed reactors, the CAVALIER and the UVAR. This paper will describe the progress in 2002, remaining efforts and the unique organizational structure of the project team.

Ervin, P. F.; Lundberg, L. A.; Benneche, P. E.; Mulder, R. U.; Steva, D. P.

2003-02-24T23:59:59.000Z

433

Hydrogasification reactor and method of operating same  

SciTech Connect

The present invention provides a system and method for evaluating effects of process parameters on hydrogasification processes. The system includes a hydrogasification reactor, a pressurized feed system, a hopper system, a hydrogen gas source, and a carrier gas source. Pressurized carbonaceous material, such as coal, is fed to the reactor using the carrier gas and reacted with hydrogen to produce natural gas.

Hobbs, Raymond; Karner, Donald; Sun, Xiaolei; Boyle, John; Noguchi, Fuyuki

2013-09-10T23:59:59.000Z

434

The Gas Reactor Makes a Comeback  

Science Journals Connector (OSTI)

...while the operators of a gas reactor can leave the...ofhis high 699 temperature gas-cooled reactor (HTGR...from the highly pressured turbine drive system might get...would produce combustible gas-es, creating the potential...too much to complete the remaining contracts. So General...

ELIOT MARSHALL

1984-05-18T23:59:59.000Z

435

Engineering Development of Ceramic Membrane Reactor  

E-Print Network (OSTI)

ceramic Ion Transport Membrane (ITM) reactor system for low-cost conversion of natural gas to hydrogen;7 A Revolutionary Technology Using Ceramic Membranes Ion Transport Membranes (ITM) ­ Non-porous multiEngineering Development of Ceramic Membrane Reactor Systems for Converting Natural Gas to Hydrogen

436

Explosive demolition of K East Reactor Stack  

ScienceCinema (OSTI)

Using $420,000 in Recovery Act funds, the Department of Energy and contractor CH2M HILL Plateau Remediation Company topped off four months of preparations when they safely demolished the exhaust stack at the K East Reactor and equipment inside the reactor building on July 23, 2010.

None

2010-09-02T23:59:59.000Z

437

Dynamic detection of nuclear reactor core incident  

Science Journals Connector (OSTI)

Surveillance, safety and security of evolving systems are a challenge to prevent accident. The dynamic detection of a hypothetical and theoretical blockage incident in the Phenix nuclear reactor is investigated. Such an incident is characterized by abnormal ... Keywords: Contrast, Dynamic detection of perturbations, Evolving system, Fast-neutron reactor, Neighbourhood, Noise

Laurent Hartert; Danielle Nuzillard; Jean-Philippe Jeannot

2013-02-01T23:59:59.000Z

438

SAFETY AND RELIABILITY ANALYSIS OF NUCLEAR REACTORS  

Science Journals Connector (OSTI)

Abstract A survey of the various aspects of safety and reliability analysis of nuclear reactors is presented with particular emphasis on the interrelation between structural reliability and systems reliability. In reactor design this interrelation is of overriding importance since it is the task of the control, protective and containment systems to protect the mechanical system and the structure from accidental overloading.

T.A. JAEGER

1972-01-01T23:59:59.000Z

439

Reactor Antineutrinos Signal all over the world  

E-Print Network (OSTI)

We present an updated estimate of reactor antineutrino signal all over the world, with particular attention to the sites proposed for existing and future geo-neutrino experiment. In our calculation we take into account the most updated data on Thermal Power for each nuclear plant, on reactor antineutrino spectra and on three neutrino oscillation mechanism.

Ricci, B; Baldoncini, M; Esposito, J; Ludhova, L; Zavatarelli, S

2014-01-01T23:59:59.000Z

440

Reactor Antineutrinos Signal all over the world  

E-Print Network (OSTI)

We present an updated estimate of reactor antineutrino signal all over the world, with particular attention to the sites proposed for existing and future geo-neutrino experiment. In our calculation we take into account the most updated data on Thermal Power for each nuclear plant, on reactor antineutrino spectra and on three neutrino oscillation mechanism.

B. Ricci; F. Mantovani; M. Baldoncini; J. Esposito; L. Ludhova; S. Zavatarelli

2014-03-17T23:59:59.000Z

Note: This page contains sample records for the topic "reactor generic distributed" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


441

CASCADE OPTIMIZATION AND CONTROL OF BATCH REACTORS  

E-Print Network (OSTI)

CASCADE OPTIMIZATION AND CONTROL OF BATCH REACTORS Xiangming Hua, Sohrab Rohani and Arthur Jutan ajutan@uwo.ca Abstract: In this study, a cascade closed-loop optimization and control strategy for batch reactor. Using model reduction a cascade system is developed, which can effectively combine optimization

Jutan, Arthur

442

Chemical Reactor Analysis and Optimal Digestion  

E-Print Network (OSTI)

J 310 Chemical Reactor Analysis and Optimal Digestion An optimal digestion theory can be readily derived from basic principles o f chemical reactor analysis and design Deborah L. Penry and Peter for formulating and solving optimization problems (Bellman 1957), the entire process is optimized only

Jumars, Pete

443

The breeder reactor: a fossil fuel viewpoint  

Science Journals Connector (OSTI)

... elegant and simple: to generate electricity and, at the same time, to produce additional fuel from the uranium discarded by the existing thermal reactor system. Without the breeder reactor, ... seems likely that the role of nuclear energy will begin to be constrained by the price and availability of uranium at about the turn of the century. There is, however ...

David Merrick

1976-12-16T23:59:59.000Z

444

Reactor Project Presses Ahead Despite Protests  

Science Journals Connector (OSTI)

...existing research reactors-in Berlin, Braunschweig, Jiilich, Geesthacht, and Munich were built in the 1950s and '60s and, even...the United States and 15 reactors abroad (including one in Geesthacht, Germany) have so far been converted to low-enriched uranium...

Robert Koenig

1995-08-04T23:59:59.000Z

445

Aerosol reactor production of uniform submicron powders  

DOE Patents (OSTI)

A method of producing submicron nonagglomerated particles in a single stage reactor includes introducing a reactant or mixture of reactants at one end while varying the temperature along the reactor to initiate reactions at a low rate. As homogeneously small numbers of seed particles generated in the initial section of the reactor progress through the reactor, the reaction is gradually accelerated through programmed increases in temperature along the length of the reactor to promote particle growth by chemical vapor deposition while minimizing agglomerate formation by maintaining a sufficiently low number concentration of particles in the reactor such that coagulation is inhibited within the residence time of particles in the reactor. The maximum temperature and minimum residence time is defined by a combination of temperature and residence time that is necessary to bring the reaction to completion. In one embodiment, electronic grade silane and high purity nitrogen are introduced into the reactor and temperatures of approximately 770.degree. K. to 1550.degree. K. are employed. In another embodiment silane and ammonia are employed at temperatures from 750.degree. K. to 1800.degree. K.

Flagan, Richard C. (Pasadena, CA); Wu, Jin J. (Pasadena, CA)

1991-02-19T23:59:59.000Z

446

The CANDU Reactor System: An Appropriate Technology  

Science Journals Connector (OSTI)

The CANDU Reactor System: An Appropriate Technology...Chalk River, Ontario, Canada K0J 1J0 CANDU power reactors are characterized by the combination...breeder. These and other features make the CANDU system an appropriate technology for countries...

J. A. L. Robertson

1978-02-10T23:59:59.000Z

447

Ontario to Mothball Two CANDU Reactors  

Science Journals Connector (OSTI)

Ontario to Mothball Two CANDU Reactors 10.1126/science.309.5739...540-megawatt Canada Deuterium-Uranium (CANDU) nuclear reactors more than a decade before...could signal the end of the road for CANDU,” says Tom Adams, executive director...

Paul Webster

2005-08-26T23:59:59.000Z

448

Nuclear reactor characteristics and operational history  

Gasoline and Diesel Fuel Update (EIA)

Nuclear > U.S. reactor operation status tables Nuclear > U.S. reactor operation status tables Nuclear Reactor Operational Status Tables Release date: November 22, 2011 Next release date: November 2012 See also: Table 1. Capacity and Generation, Table 2. Ownership Data Table 3. Nuclear Reactor Characteristics and Operational History PDF XLS Plant Name Generator ID Type Reactor Supplier and Model Construction Start Grid Connection Original Expiration Date License Renewal Application License Renewal Issued Extended Expiration Arkansas Nuclear One 1 PWR Babcock&Wilcox, Lower Loop 10/1/1968 8/17/1974 5/20/2014 2/1/2000 6/20/2001 5/20/2034 Arkansas Nuclear One 2 PWR Combustion Eng. 7/1/1971 12/26/1978 7/17/2018 10/15/2003 6/30/2005 7/17/2038

449

Daya Bay Reactor Neutrino Project at NERSC  

NLE Websites -- All DOE Office Websites (Extended Search)

Daya Bay Reactor Neutrino Daya Bay Reactor Neutrino Experiment Daya Bay Reactor Neutrino Experiment Daya Bay is an international neutrino-oscillation experiment designed to determine the last unknown neutrino mixing angle θ13 using anti-neutrinos produced by the Daya Bay and Ling Ao Nuclear Power Plant reactors. The experiment is being built by blasting three kilometers of tunnel through the granite rock under the mountains where the power plants are located. Data collection is now scheduled to start in in 2011. On the PDSF cluster at NERSC, Daya Bay performs simulations of the detectors, reactors, and surrounding mountains to help design and anticipate detector properties and behavior. Once real data are available, Daya Bay will be using NERSC to analyze data and NERSC HPSS will be the central U.S. repository for all raw

450

Reactivity control assembly for nuclear reactor. [LMFBR  

DOE Patents (OSTI)

This invention, which resulted from a contact with the United States Department of Energy, relates to a control mechanism for a nuclear reactor and, more particularly, to an assembly for selectively shifting different numbers of reactivity modifying rods into and out of the core of a nuclear reactor. It has been proposed heretofore to control the reactivity of a breeder reactor by varying the depth of insertion of control rods (e.g., rods containing a fertile material such as ThO/sub 2/) in the core of the reactor, thereby varying the amount of neutron-thermalizing coolant and the amount of neutron-capturing material in the core. This invention relates to a mechanism which can advantageously be used in this type of reactor control system.

Bollinger, L.R.

1982-03-17T23:59:59.000Z

451

Cooling system for a nuclear reactor  

DOE Patents (OSTI)

A cooling system for a gas-cooled nuclear reactor is disclosed which includes at least one primary cooling loop adapted to pass coolant gas from the reactor core and an associated steam generator through a duct system having a main circulator therein, and at least one auxiliary cooling loop having communication with the reactor core and adapted to selectively pass coolant gas through an auxiliary heat exchanger and circulator. The main and auxiliary circulators are installed in a common vertical cavity in the reactor vessel, and a common return duct communicates with the reactor core and intersects the common cavity at a junction at which is located a flow diverter valve operative to effect coolant flow through either the primary or auxiliary cooling loops.

Amtmann, Hans H. (Rancho Santa Fe, CA)

1982-01-01T23:59:59.000Z

452

Features of a subcritical nuclear reactor  

Science Journals Connector (OSTI)

Abstract A subcritical nuclear reactor is a device where the nuclear-fission chain reaction is initiated and maintained using an external neutron source. It is a valuable educational and research tool where in a safe way many reactor parameters can be measured. Here, we have used the six-factor formula to calculate the effective multiplication factor of a subcritical nuclear reactor Nuclear Chicago model 9000. Using the MCNP5 code, a three-dimensional model of the subcritical reactor was developed to estimate the effective multiplication factor, the neutron spectra, and the total and thermal neutron fluences along the radial and axial axis. The MCNP5 results of the effective multiplication factor were compared with those obtained from the six-factor formula. The effective dose and the Ambient dose equivalent, at three sites outside the reactor, were estimated; the Ambient dose equivalent was also measured and compared with the calculated values.

Hector Rene Vega-Carrillo; Isvi Ruben Esparza-Garcia; Alvaro Sanchez

2015-01-01T23:59:59.000Z

453

Operational control of boiling water reactor stability  

SciTech Connect

Boiling water reactor cores are susceptible to instabilities, which generate power oscillations. Specific reactor operating practices can provide a mechanism for control of the instability phenomenon. An axial separation of the core into a single-phase region and a two-phase region resolves the influence of axial flux shapes on core stability. This separation provides the means to derive a core stability control that ensures significant reactor stability margin. The control is achieved by maintaining the core average bulk coolant saturation elevation above a predetermined axial plane. The control can be reliably and efficiently implemented during reactor operations. Analysis demonstrates that variations in parameters important to stability have only secondary influences on stability margin when the control is in effect. Actual plant experience with a large commercial boiling water reactor confirms the capabilities of this stability control in an operational setting.

Mowry, C.M. [PECO Energy, Wayne, PA (United States); Nir, I. [Entergy Operations, Jackson, MS (United States); Newkirk, D.W. [GE Nuclear Energy, San Jose, CA (United States)

1995-03-01T23:59:59.000Z

454

Self-actuating reactor shutdown system  

DOE Patents (OSTI)

A control system for the automatic or self-actuated shutdown or "scram" of a nuclear reactor. The system is capable of initiating scram insertion by a signal from the plant protection system or by independent action directly sensing reactor conditions of low-flow or over-power. Self-actuation due to a loss of reactor coolant flow results from a decrease of pressure differential between the upper and lower ends of an absorber element. When the force due to this differential falls below the weight of the element, the element will fall by gravitational force to scram the reactor. Self-actuation due to high neutron flux is accomplished via a valve controlled by an electromagnet and a thermionic diode. In a reactor over-power, the diode will be heated to a change of state causing the electromagnet to be shorted thereby actuating the valve which provides the changed flow and pressure conditions required for scramming the absorber element.

Barrus, Donald M. (San Jose, CA); Brummond, Willian A (Livermore, CA); Peterson, Leslie F. (Danville, CA)

1988-01-01T23:59:59.000Z

455

Nuclear reactor characteristics and operational history  

U.S. Energy Information Administration (EIA) Indexed Site

Nuclear > U.S. reactor operation status tables Nuclear > U.S. reactor operation status tables Nuclear Reactor Operational Status Tables Release date: November 22, 2011 Next release date: November 2012 See also: Table 2. Ownership Data, Table 3. Characteristics and Operational History Table 1. Nuclear Reactor, State, Type, Net Capacity, Generation, and Capacity Factor PDF XLS Plant/Reactor Name Generator ID State Type 2009 Summer Capacity Net MW(e)1 2010 Annual Generation Net MWh2 Capacity Factor Percent3 Arkansas Nuclear One 1 AR PWR 842 6,607,090 90 Arkansas Nuclear One 2 AR PWR 993 8,415,588 97 Beaver Valley 1 PA PWR 892 7,119,413 91 Beaver Valley 2 PA PWR 885 7,874,151 102 Braidwood Generation Station 1 IL PWR 1,178 9,196,689 89

456

Small Modular Nuclear Reactors | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Reactor Technologies » Small Modular Reactor Technologies » Small Modular Nuclear Reactors Small Modular Nuclear Reactors Cutaway of 2-Unit Generation mPower SMR Installation. | © 2012 Generation mPower LLC. All Rights Reserved. Reprinted with permission. Cutaway of 2-Unit Generation mPower SMR Installation. | © 2012 Generation mPower LLC. All Rights Reserved. Reprinted with permission. The development of clean, affordable nuclear power options is a key element of the Department of Energy's Office of Nuclear Energy (DOE-NE) Nuclear Energy Research and Development Roadmap. As a part of this strategy, a high priority of the Department has been to help accelerate the timelines for the commercialization and deployment of small modular reactor (SMR) technologies through the SMR Licensing Technical Support program. Begun

457

Early Exploration - Reactors designed/built by Argonne National Laboratory  

NLE Websites -- All DOE Office Websites (Extended Search)

Early Exploration Early Exploration About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

458

Consortium for Advanced Simulation of Light Water Reactors (CASL...  

NLE Websites -- All DOE Office Websites (Extended Search)

behavior in AP1000 reactor core Test run signals emergence of the next generation in nuclear power reactor analysis tools OAK RIDGE, Tenn., Feb. 18, 2014 - Scientists and...

459

SciTech Connect: Nuclear power reactor instrumentation systems...  

Office of Scientific and Technical Information (OSTI)

Nuclear power reactor instrumentation systems handbook. Volume 1 Citation Details In-Document Search Title: Nuclear power reactor instrumentation systems handbook. Volume 1 You...

460

TABLE 1. Nuclear Reactor, State, Type, Net Capacity, Generation...  

U.S. Energy Information Administration (EIA) Indexed Site

TABLE 1. Nuclear Reactor, State, Type, Net Capacity, Generation, and Capacity Factor " "PlantReactor Name","Generator ID","State","Type","2009 Summer Capacity"," 2010 Annual...

Note: This page contains sample records for the topic "reactor generic distributed" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


461

Table 3. Nuclear Reactor Characteristics and Operational History  

U.S. Energy Information Administration (EIA) Indexed Site

3. Nuclear Reactor Characteristics and Operational History" "Plant Name","Generator ID","Type","Reactor Supplier and Model","Construction Start","Grid Connection","Commercial...

462

Consortium for Advanced Simulation of Light Water Reactors (CASL...  

NLE Websites -- All DOE Office Websites (Extended Search)

Presentations 2015 back to top Smith, K., Advances in Reactor Physics and Computational Science, Physor 2014 International Conference, "The Role of Reactor Physics toward a...

463

A Stochastic Reactor Based Virtual Engine Model Employing Detailed...  

Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

A Stochastic Reactor Based Virtual Engine Model Employing Detailed Chemistry for Kinetic Studies of In-Cylinder Combustion and Exhaust Aftertreatment A Stochastic Reactor Based...

464

Light Water Reactors A DOE Energy Innovation Hub for Modeling...  

NLE Websites -- All DOE Office Websites (Extended Search)

Consortium for Advanced Simulation of Light Water Reactors A DOE Energy Innovation Hub for Modeling and Simulation of Nuclear Reactors CASL is focused on three issues for nuclear...

465

Analysis of the antineutrino rate during CANDU reactor startup.  

E-Print Network (OSTI)

??Detection systems used to monitor reactor operations are of significant interest as tools for verification of operator declarations. Current reactor site safeguards are limited to… (more)

Matthews, Christopher

2012-01-01T23:59:59.000Z

466

Fast Spectrum Molten Salt Reactor Options  

SciTech Connect

During 2010, fast-spectrum molten-salt reactors (FS-MSRs) were selected as a transformational reactor concept for light-water reactor (LWR)-derived heavy actinide disposition by the Department of Energy-Nuclear Energy Advanced Reactor Concepts (ARC) program and were the subject of a preliminary scoping investigation. Much of the reactor description information presented in this report derives from the preliminary studies performed for the ARC project. This report, however, has a somewhat broader scope-providing a conceptual overview of the characteristics and design options for FS-MSRs. It does not present in-depth evaluation of any FS-MSR particular characteristic, but instead provides an overview of all of the major reactor system technologies and characteristics, including the technology developments since the end of major molten salt reactor (MSR) development efforts in the 1970s. This report first presents a historical overview of the FS-MSR technology and describes the innovative characteristics of an FS-MSR. Next, it provides an overview of possible reactor configurations. The following design features/options and performance considerations are described including: (1) reactor salt options-both chloride and fluoride salts; (2) the impact of changing the carrier salt and actinide concentration on conversion ratio; (3) the conversion ratio; (4) an overview of the fuel salt chemical processing; (5) potential power cycles and hydrogen production options; and (6) overview of the performance characteristics of FS-MSRs, including general comparative metrics with LWRs. The conceptual-level evaluation includes resource sustainability, proliferation resistance, economics, and safety. The report concludes with a description of the work necessary to begin more detailed evaluation of FS-MSRs as a realistic reactor and fuel cycle option.

Gehin, Jess C [ORNL; Holcomb, David Eugene [ORNL; Flanagan, George F [ORNL; Patton, Bruce W [ORNL; Howard, Rob L [ORNL; Harrison, Thomas J [ORNL

2011-07-01T23:59:59.000Z

467

Buildings Energy Data Book: 1.5 Generic Fuel Quad and Comparison  

Buildings Energy Data Book (EERE)

6 6 Shares of U.S. Buildings Generic Quad (Percent) (1) Renewables Natural Gas Petroleum Coal Hydro. Other Total Nuclear Total 1980 39% 12% 31% 7% 4% 11% 7% 100% 1981 38% 11% 32% 7% 4% 11% 8% 100% 1982 37% 10% 33% 8% 4% 12% 8% 100% 1983 35% 10% 34% 9% 4% 13% 8% 100% 1984 35% 10% 34% 8% 4% 12% 8% 100% 1985 34% 10% 35% 7% 4% 11% 10% 100% 1986 32% 10% 36% 7% 4% 11% 11% 100% 1987 32% 10% 37% 6% 4% 10% 11% 100% 1988 32% 10% 37% 5% 4% 9% 13% 100% 1989 32% 9% 36% 6% 5% 11% 12% 100% 1990 32% 8% 36% 7% 4% 10% 13% 100% 1991 32% 8% 36% 7% 4% 10% 14% 100% 1992 33% 8% 36% 6% 4% 10% 14% 100% 1993 33% 7% 37% 6% 4% 10% 13% 100% 1994 33% 7% 36% 5% 4% 9% 14% 100% 1995 33% 7% 36% 6% 3% 10% 14% 100% 1996 33% 7% 36% 7% 3% 10% 14% 100% 1997 33% 6% 38% 7% 3% 10% 13% 100% 1998 32% 6% 38% 6% 3% 9% 14% 100% 1999 32% 6% 38% 6% 3% 9% 15% 100% 2000 32% 6% 38% 5% 3% 8% 15% 100% 2001 32% 6% 38% 4% 3% 7% 16% 100% 2002 33% 6% 38% 5% 3% 8% 15% 100% 2003 32% 6% 38% 5% 3% 8% 15% 100% 2004 32% 6% 38% 5% 3% 8%

468

Numerical Modeling of Geomechanical Processes Related to CO{sub 2} Injection within Generic Reservoirs  

SciTech Connect

In this project generic anticline structures have been used for numerical modeling analyses to study the influence of geometrical parameters, fluid flow boundary conditions, in situ stress regime and inter-bedding friction coefficient on geomechanical risks such as fracture reactivation and fracture generation. The resulting stress states for these structures are also used to determine safe drilling directions and a methodology for wellbore trajection optimization is developed that is applicable for non-Andersonian stress states. The results of the fluid flow simulation show that the type of fluid flow boundary condition is of utmost importance and has significant impact on all injection related parameters. It is recommended that further research is conducted to establish a method to quantify the fluid flow boundary conditions for injection applications. The results of the geomechanical simulation show that in situ stress regime is a crucial, if not the most important, factor determining geomechanical risks. For extension and strike slip stress regimes anticline structures should be favored over horizontally layered basin as they feature higher ?P{sub c} magnitudes. If sedimentary basins are tectonically relaxed and their state of stress is characterized by the uni-axial strain model the basin is in exact frictional equilibrium and fluids should not be injected. The results also show that low inter bedding friction coefficients effectively decouple layers resulting in lower ?P{sub c} magnitudes, especially for the compressional stress regime.

Eckert, Andreas

2013-05-31T23:59:59.000Z

469

Power management of hybrid micro-grid system by a generic centralized supervisory control scheme  

Science Journals Connector (OSTI)

Abstract This paper presents a generic centralized supervisory control scheme for the power management of multiple power converters based hybrid micro-grid system. The system consists of wind generators, photovoltaic system, multiple parallel connected power converters, utility grid, ac and dc loads. Power management of the micro-grid is performed under two cases: grid mode and local mode. Central supervisory unit (CSU) generates command signal to ensure the power management during the two modes. In local mode, the dc loads in the ac–dc hybrid system can be controlled. In the case of grid mode operation, power flow between the utility grid and micro-grid is controlled. A novel feature of this paper is the incorporation of the multiple power converters. The generated command signal from the CSU can also control the operation of the multiple power converters in both grid and local modes. An additional feature is the incorporation of sodium sulfur battery energy storage system (NAS BESS) which is used to smooth the output power fluctuation of the wind farm. The effectiveness of the control scheme is also verified using real time load pattern. The simulation is performed in PSCAD/EMTDC.

Mir Nahidul Ambia; Ahmed Al-Durra; Cedric Caruana; S.M. Muyeen

2014-01-01T23:59:59.000Z

470

PHYSICS AND SAFETY ANALYSIS FOR THE NIST RESEARCH REACTOR.  

SciTech Connect

Detailed reactor physics and safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analyses provide an update to the Final Safety Analysis Report (FSAR) and employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron and photon transport calculations were performed with the MCNP code to determine the safety parameters for the NBSR. The core depletion and determination of the fuel compositions were performed with MONTEBURNS. MCNP calculations were performed to determine the beginning, middle, and end-of-cycle power distributions, moderator temperature coefficient, and shim safety arm, beam tube and void reactivity worths. The calculational model included a plate-by-plate description of each fuel assembly, axial mid-plane water gap, beam tubes and the tubular geometry of the shim safety arms. The time-dependent analysis of the primary loop was determined with a RELAP5 transient analysis model that includes the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. The statistical analysis used to assure protection from critical heat flux (CHF) was performed using a Monte Carlo simulation of the uncertainties contributing to the CHF calculation. The power distributions used to determine the local fuel conditions and margin to CHF were determined with MCNP. Evaluations were performed for the following accidents: (1) the control rod withdrawal startup accident, (2) the maximum reactivity insertion accident, (3) loss-of-flow resulting from loss of electrical power, (4) loss-of-flow resulting from a primary pump seizure, (5) loss-of-flow resulting from inadvertent throttling of a flow control valve, (6) loss-of-flow resulting from failure of both shutdown cooling pumps and (7) misloading of a fuel element. In both the startup and maximum reactivity insertion accidents, the core power transient is terminated by a reactor trip at 26 MW. The calculations show that both the peak reactor power and the excursion energy depend on the negative reactivity insertion from reactor trip. In one of the loss-of-flow accidents offsite electrical power is assumed lost to the three operating primary pumps. A slightly delayed reactor scram is initiated as a result of primary flow coast down. The RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur, because of the momentum of the coolant flowing through the fuel channels and the negative scram reactivity insertion. For both the primary pump seizure and inadvertent throttling of a flow control valve, the RELAP5 analyses indicate that the reduction in power following the trip is sufficient to ensure that there is adequate margin to CHF and that the fuel cladding does not fail. The analysis of the loss-of-flow accident in the extremely unlikely case where both shutdown pumps fail, shows that the cooling provided by the D{sub 2}O is sufficient to ensure the cladding does not fail. The power distributions were examined for a set of fuel misloadings in which a fresh fuel element is moved from a peripheral low-reactivity location to a central high-reactivity location. The calculations show that there is adequate margin to CHF and the cladding does not fail. An additional analysis was performed to simulate the operation at low power (500 kW) without forced flow cooling. The result indicates that natural convection cooling is adequate for operation of the NBSR at a power level of 500 kW.

CHENG,L.HANSON,A.DIAMOND,D.XU,J.CAREW,J.RORER,D.

2004-03-31T23:59:59.000Z

471

New Fuel Cycle and Fuel Management Options in Heavy Liquid Metal-Cooled Reactors  

Science Journals Connector (OSTI)

Technical Paper / Advances in Nuclear Fuel Management - Fuel Management of Reactors Other Than Light Water Reactors

Ehud Greenspan; Pavel Hejzlar; Hiroshi Sekimoto; Georgy Toshinsky; David Wade

472

Feasibility of Burning First- and Second-Generation Plutonium in Pebble Bed High-Temperature Reactors  

Science Journals Connector (OSTI)

Technical Paper / Advances in Nuclear Fuel Management - Fuel Management of Reactors Other Than Light Water Reactors

J. B. M. De Haas; J. C. Kuijper

473

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)

2013-07-01T23:59:59.000Z

474

The Ecological Ideal Free Distribution and Distributed Networked Control Systems  

Science Journals Connector (OSTI)

In this paper we first establish an analogy where we view both animals and vehicles as generic agents. We introduce a model of the ecological behavior of a group of agents and ... Finally, we apply this model to ...

Jorge Finke; Kevin M. Passino

2011-01-01T23:59:59.000Z

475

Nuclear reactor internals alignment configuration  

DOE Patents (OSTI)

An alignment system that employs jacking block assemblies and alignment posts around the periphery of the top plate of a nuclear reactor lower internals core shroud to align an upper core plate with the lower internals and the core shroud with the core barrel. The distal ends of the alignment posts are chamfered and are closely received within notches machined in the upper core plate at spaced locations around the outer circumference of the upper core plate. The jacking block assemblies are used to center the core shroud in the core barrel and the alignment posts assure the proper orientation of the upper core plate. The alignment posts may alternately be formed in the upper core plate and the notches may be formed in top plate.

Gilmore, Charles B. (Greensburg, PA); Singleton, Norman R. (Murrysville, PA)

2009-11-10T23:59:59.000Z

476

Gas-cooled nuclear reactor  

DOE Patents (OSTI)

A gas-cooled nuclear reactor includes a central core located in the lower portion of a prestressed concrete reactor vessel. Primary coolant gas flows upward through the core and into four overlying heat-exchangers wherein stream is generated. During normal operation, the return flow of coolant is between the core and the vessel sidewall to a pair of motor-driven circulators located at about the bottom of the concrete pressure vessel. The circulators repressurize the gas coolant and return it back to the core through passageways in the underlying core structure. If during emergency conditions the primary circulators are no longer functioning, the decay heat is effectively removed from the core by means of natural convection circulation. The hot gas rising through the core exits the top of the shroud of the heat-exchangers and flows radially outward to the sidewall of the concrete pressure vessel. A metal liner covers the entire inside concrete surfaces of the concrete pressure vessel, and cooling tubes are welded to the exterior or concrete side of the metal liner. The gas coolant is in direct contact with the interior surface of the metal liner and transfers its heat through the metal liner to the liquid coolant flowing through the cooling tubes. The cooler gas is more dense and creates a downward convection flow in the region between the core and the sidewall until it reaches the bottom of the concrete pressure vessel when it flows radially inward and up into the core for another pass. Water is forced to flow through the cooling tubes to absorb heat from the core at a sufficient rate to remove enough of the decay heat created in the core to prevent overheating of the core or the vessel.

Peinado, Charles O. (La Jolla, CA); Koutz, Stanley L. (San Diego, CA)

1985-01-01T23:59:59.000Z

477

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

SciTech Connect

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

Parma, Edward J., Jr.

2009-06-01T23:59:59.000Z

478

Distributed design tools: Mapping targeted design tools onto a Web-based distributed architecture for high-performance computing  

SciTech Connect

Design Tools use a Web-based Java interface to guide a product designer through the design-to-analysis cycle for a specific, well-constrained design problem. When these Design Tools are mapped onto a Web-based distributed architecture for high-performance computing, the result is a family of Distributed Design Tools (DDTs). The software components that enable this mapping consist of a Task Sequencer, a generic Script Execution Service, and the storage of both data and metadata in an active, object-oriented database called the Product Database Operator (PDO). The benefits of DDTs include improved security, reliability, scalability (in both problem size and computing hardware), robustness, and reusability. In addition, access to the PDO unlocks its wide range of services for distributed components, such as lookup and launch capability, persistent shared memory for communication between cooperating services, state management, event notification, and archival of design-to-analysis session data.

Holmes, V.P.; Linebarger, J.M.; Miller, D.J.; Poore, C.A.

1999-11-30T23:59:59.000Z

479

Flexible Conversion Ratio Fast Reactor Systems Evaluation  

SciTech Connect

Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

Neil Todreas; Pavel Hejzlar

2008-06-30T23:59:59.000Z

480

Summary Notes from 15 November 2007 Generic Technical Issue Discussion on Sensitivity and Uncertainty Analysis and Model Support  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

5 November 2007 Generic Technical Issue Discussion on 5 November 2007 Generic Technical Issue Discussion on Sensitivity and Uncertainty Analysis and Model Support Attendees: Representatives from Department of Energy-Headquarters (DOE-HQ) and the U.S. Nuclear Regulatory Commission (NRC) met at the DOE offices in Germantown, Maryland on 15 November 2007. Representatives from Department of Energy-Savannah River (DOE-SR) and the South Carolina Department of Health and Environmental Control (SCDHEC) participated in the meeting via a teleconference link. Discussion: DOE believes that based on the position papers provided prior to the meeting, DOE and NRC staff have many areas of agreement and no significant areas of disagreement with respect to the specific sensitivity and uncertainty analysis requirements articulated in the respective DOE and NRC requirements. The NRC

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481

Summary Notes from 22 July 2008 Generic Technical Issue Discussion on Long-Term Engineered Cap Performance  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Summary Notes from 22 July 2008 Generic Technical Issue Discussion on Long-Term Summary Notes from 22 July 2008 Generic Technical Issue Discussion on Long-Term Engineered Cap Performance Attendees: Representatives from the U.S. Department of Energy (DOE)-Headquarters and the U.S. Nuclear Regulatory Commission (NRC) staff met at the DOE offices in Germantown, Maryland on 22 July 2008. Representatives from South Carolina Department of Health and Environmental Control, DOE-Savannah River, and DOE- Office of River Protection participated in the meeting via a teleconference link. Discussion: NRC staff prepared and disseminated agenda topics (listed in the next section) summarizing issues and considerations relative to estimating long-term engineered cover or cap performance. A summary of the discussion regarding each agenda topic is provided below. The purpose of this meeting was for DOE and NRC staff

482

Parton distributions for the LHC Run II  

E-Print Network (OSTI)

We present NNPDF3.0, the first set of parton distribution functions (PDFs) determined with a methodology validated by a closure test. NNPDF3.0 uses a global dataset including HERA-II deep-inelastic inclusive cross-sections, the combined HERA charm data, jet production from ATLAS and CMS, vector boson rapidity and transverse momentum distributions from ATLAS, CMS and LHCb, W+c data from CMS and top quark pair production total cross sections from ATLAS and CMS. Results are based on LO, NLO and NNLO QCD theory and also include electroweak corrections. To validate our methodology, we show that PDFs determined from pseudo-data generated from a known underlying law correctly reproduce the statistical distributions expected on the basis of the assumed experimental uncertainties. This closure test ensures that our methodological uncertainties are negligible in comparison to the generic theoretical and experimental uncertainties of PDF determination. This enables us to determine with confidence PDFs at different perturbative orders and using a variety of experimental datasets ranging from HERA-only up to a global set including the latest LHC results, all using precisely the same validated methodology. We explore some of the phenomenological implications of our results for the upcoming 13 TeV Run of the LHC, in particular for Higgs production cross-sections.

The NNPDF Collaboration; Richard D. Ball; Valerio Bertone; Stefano Carrazza; Christopher S. Deans; Luigi Del Debbio; Stefano Forte; Alberto Guffanti; Nathan P. Hartland; Jose I. Latorre; Juan Rojo; Maria Ubiali

2014-11-06T23:59:59.000Z

483

Scaling properties in the adsorption of ionic polymeric surfactants on generic nanoparticles of metallic oxides by mesoscopic simulation  

E-Print Network (OSTI)

We study the scaling of adsorption isotherms of polyacrylic dispersants on generic surfaces of metallic oxides $XnOm$ as a function of the number of monomeric units, using Electrostatic Dissipative Particle Dynamics simulations. The simulations show how the scaling properties in these systems emerge and how the isotherms rescale to a universal curve, reproducing reported experimental results. The critical exponent for these systems is also obtained, in perfect agreement with the scaling theory of deGennes. Some important applications are mentioned.

E. Mayoral; E. Nahmad-Achar

2014-02-11T23:59:59.000Z

484

Nested off-diagonal Bethe ansatz and exact solutions of the su(n) spin chain with generic integrable boundaries  

E-Print Network (OSTI)

The nested off-diagonal Bethe ansatz method is proposed to diagonalize multi-component integrable models with generic integrable boundaries. As an example, the exact solutions of the su(n)-invariant spin chain model with both periodic and non-diagonal boundaries are derived by constructing the nested T-Q relations based on the operator product identities among the fused transfer matrices and the asymptotic behavior of the transfer matrices.

Junpeng Cao; Wen-Li Yang; Kangjie Shi; Yupeng Wang

2014-05-11T23:59:59.000Z

485

Measurements of Film Flow Rate in Heated Tubes with Various Axial Power Distributions  

E-Print Network (OSTI)

Measurements of Film Flow Rate in Heated Tubes with Various Axial Power Distributions by Carl, Measurements of Film Flow Rate in Heated Tubes with Various Axial Power Distributions KTH Nuclear Reactor power is limited by a phenomenon called critical heat flux (CHF). It appears as a sudden detoriation

Haviland, David

486

Solid oxide fuel cell systems with hot zones having improved reactant distribution  

DOE Patents (OSTI)

A Solid Oxide Fuel Cell (SOFC) system having a hot zone with a center cathode air feed tube for improved reactant distribution, a CPOX reactor attached at the anode feed end of the hot zone with a tail gas combustor at the opposing end for more uniform heat distribution, and a counter-flow heat exchanger for efficient heat retention.

Poshusta, Joseph C.; Booten, Charles W.; Martin, Jerry L.

2012-11-06T23:59:59.000Z

487

Nuclear reactor fissile isotopes antineutrino spectra  

E-Print Network (OSTI)

Positron spectrum from inverse beta decay reaction on proton was measured in 1988-1990 as a result of neutrino exploration experiment. The measured spectrum has the largest statistics and lowest energy threshold between other neutrino experiments made that time at nuclear reactors. On base of the positron spectrum the standard antineutrino spectrum for typical reactor fuel composition was restored. In presented analysis the partial spectra forming this standard spectrum were extracted using specific method. They could be used for neutrino experiments data analysis made at any fuel composition of reactor core.

V. Sinev

2012-07-30T23:59:59.000Z

488

Neutronic Assessment of Transmutation Target Compositions in Heterogeneous Sodium Fast Reactor Geometries  

SciTech Connect

The sodium fast reactor is under consideration for consuming the transuranic waste in the spent nuclear fuel generated by light water reactors. This work is concerned with specialized target assemblies for an oxide-fueled sodium fast reactor that are designed exclusively for burning the americium and higher mass actinide component of light water reactor spent nuclear fuel (SNF). The associated gamma and neutron radioactivity, as well as thermal heat, associated with decay of these actinides may significantly complicate fuel handling and fabrication of recycled fast reactor fuel. The objective of using targets is to isolate in a smaller number of assemblies these concentrations of higher actinides, thus reducing the volume of fuel having more rigorous handling requirements or a more complicated fabrication process. This is in contrast to homogeneous recycle where all recycled actinides are distributed among all fuel assemblies. Several heterogeneous core geometries were evaluated to determine the fewest target assemblies required to burn these actinides without violating a set of established fuel performance criteria. The DIF3D/REBUS code from Argonne National Laboratory was used to perform the core physics and accompanying fuel cycle calculations in support of this work. Using the REBUS code, each core design was evaluated at the equilibrium cycle condition.

Samuel E. Bays; Rodolfo M. Ferrer; Michael A. Pope; Benoit Forget; Mehdi Asgari

2008-02-01T23:59:59.000Z

489

SCALE-4 Analysis of LaSalle Unit 1 BWR Commercial Reactor Critical Configuration  

SciTech Connect

Five commercial reactor criticals (CRCs) for the LaSalle Unit 1 boiling-water reactor have been analyzed using KENO V.a, the Monte Carlo criticality code of the SCALE 4 code system. The irradiated fuel assembly isotopics for the criticality analyses were provided by the Waste Package Design team at the Yucca Mountain Project in the United States, who performed the depletion calculations using the SAS2H sequence of SCALE 4. The reactor critical measurements involved two beginning-of-cycle and three middle-of-cycle configurations. The CRCs involved relatively low-cycle burnups, and therefore contained a relatively high gadolinium poison content in the reactor assemblies. This report summarizes the data and methods used in analyzing the critical configurations and assesses the sensitivity of the results to some of the modeling approximations used to represent the gadolinium poison distribution within the assemblies. The KENO V.a calculations, performed using the SCALE 44GROUPNDF5 ENDF/B-V cross-section library, yield predicted k{sub eff} values within about 1% {Delta}k/k relative to reactor measurements for the five CRCs using general 8-pin and 9-pin heterogeneous gadolinium poison pin assembly models.

Gauld, I.C.

2000-03-16T23:59:59.000Z

490

Use of sensitivity and uncertainty analysis in reactor design verification Part II: Flux measurement analysis  

Science Journals Connector (OSTI)

Abstract As for any new reactor design, the ACR-1000® design has to go through a comprehensive design verification process. One of the activities for supporting the ACR physics design calculations using the ACR physics code toolset, namely WIMS-AECL/DRAGON/RFSP, is to compare the flux distributions resulting from the calculation using this toolset at various power calibration monitor (PCM) detector locations against the flux measurement data from the Japanese Advanced Thermal Reactor (ATR) FUGEN. The discussion of this particular design verification exercise will be presented in a two-part paper. The usage of data from the FUGEN reactor qualifies this exercise as design verification by alternate analysis. In order to have meaningful results at the end of the design verification process, the similarity between the ACR-1000 and FUGEN reactors has to be demonstrated. It is accomplished through the sensitivity and uncertainty analysis using the TSUNAMI (Tools for Sensitivity and Uncertainty Analysis Methodology Implementation) methodology. The results from the similarity comparison have been presented in Part I of the paper. In Part II, results from flux distribution comparison will be presented. Favourable results from this design verification exercise give a high level of confidence that using the same physics toolset in calculating the flux distribution for ACR-1000 reactor will produce results with acceptable fidelity. In addition, the results will also give an indication of expected margins in the design calculations, not only at the locations of the PCM detectors but also at the derived bundle and channel powers obtained through the flux mapping calculation.

Doddy Kastanya; Mohamed Dahmani

2013-01-01T23:59:59.000Z

491

A methodology for the evaluation of the turbine jet engine fragment threat to generic air transportable containers  

SciTech Connect

Uncontained, high-energy gas turbine engine fragments are a potential threat to air-transportable containers carried aboard jet aircraft. The threat to a generic example container is evaluated by probability analyses and penetration testing to demonstrate the methodology to be used in the evaluation of a specific container/aircraft/engine combination. Fragment/container impact probability is the product of the uncontained fragment release rate and the geometric probability that a container is in the path of this fragment. The probability of a high-energy rotor burst fragment from four generic aircraft engines striking one of the containment vessels aboard a transport aircraft is approximately 1.2 {times} 10{sup {minus}9} strikes/hour. Finite element penetration analyses and tests can be performed to identify specific fragments which have the potential to penetrate a generic or specific containment vessel. The relatively low probability of engine fragment/container impacts is primarily due to the low release rate of uncontained, hazardous jet engine fragments.

Harding, D.C.; Pierce, J.D.

1993-06-01T23:59:59.000Z

492