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Note: This page contains sample records for the topic "reactor generic distributed" from the National Library of EnergyBeta (NLEBeta).
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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.


1

Generic small modular reactor plant design.  

SciTech Connect (OSTI)

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

Performance and Safety Analysis of a Generic Small Modular Reactor  

E-Print Network [OSTI]

renewable energy sources such as, wind and solar, along with their limitations on the areas of applicability and the energy output calls for a renaissance in nuclear energy. In this second nuclear era, deliberately small reactors are poised to play a major...

Kitcher, Evans Damenortey, 1987-

2012-11-07T23:59:59.000Z

3

Distributed expert systems for nuclear reactor control  

SciTech Connect (OSTI)

A network of distributed expert systems is the heart of a prototype supervisory control architecture developed at the Oak Ridge National Laboratory (ORNL) for an advanced multimodular reactor. Eight expert systems encode knowledge on signal acquisition, diagnostics, safeguards, and control strategies in a hybrid rule-based, multiprocessing and object-oriented distributed computing environment. An interactive simulation of a power block consisting of three reactors and one turbine provides a realistic, testbed for performance analysis of the integrated control system in real-time. Implementation details and representative reactor transients are discussed.

Otaduy, P.J.

1992-12-01T23:59:59.000Z

4

Distributed expert systems for nuclear reactor control  

SciTech Connect (OSTI)

A network of distributed expert systems is the heart of a prototype supervisory control architecture developed at the Oak Ridge National Laboratory (ORNL) for an advanced multimodular reactor. Eight expert systems encode knowledge on signal acquisition, diagnostics, safeguards, and control strategies in a hybrid rule-based, multiprocessing and object-oriented distributed computing environment. An interactive simulation of a power block consisting of three reactors and one turbine provides a realistic, testbed for performance analysis of the integrated control system in real-time. Implementation details and representative reactor transients are discussed.

Otaduy, P.J.

1992-01-01T23:59:59.000Z

5

Distributed computing and nuclear reactor analysis  

SciTech Connect (OSTI)

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

6

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

7

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

8

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, John T. (Los Alamos, NM); Kunz, Walter E. (Santa Fe, NM); Atencio, James D. (Los Alamos, NM)

1984-01-01T23:59:59.000Z

9

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

10

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

SciTech Connect (OSTI)

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

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

2012-04-04T23:59:59.000Z

11

Savannah River Site generic data base development  

SciTech Connect (OSTI)

This report describes the results of a project to improve the generic component failure database for the Savannah River Site (SRS). Additionally, guidelines were developed further for more advanced applications of database values. A representative list of components and failure modes for SRS risk models was generated by reviewing existing safety analyses and component failure data bases and from suggestions from SRS safety analysts. Then sources of data or failure rate estimates were identified and reviewed for applicability. A major source of information was the Nuclear Computerized Library for Assessing Reactor Reliability, or NUCLARR. This source includes an extensive collection of failure data and failure rate estimates for commercial nuclear power plants. A recent Idaho National Engineering Laboratory report on failure data from the Idaho Chemical Processing Plant was also reviewed. From these and other recent sources, failure data and failure rate estimates were collected for the components and failure modes of interest. For each component failure mode, this information was aggregated to obtain a recommended generic failure rate distribution (mean and error factor based on a lognormal distribution). Results are presented in a table in this report. A major difference between generic database and previous efforts is that this effort estimates failure rates based on actual data (failure events) rather than on existing failure rate estimates. This effort was successful in that over 75% of the results are now based on actual data. Also included is a section on guidelines for more advanced applications of failure rate data. This report describes the results of a project to improve the generic component failure database for the Savannah River site (SRS). Additionally, guidelines were developed further for more advanced applications of database values.

Blanchard , A.

2000-01-04T23:59:59.000Z

12

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

SciTech Connect (OSTI)

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

13

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.

14

Seasoning of plasma etching reactors: Ion energy distributions to walls and real-time and run-to-run control strategies  

E-Print Network [OSTI]

Seasoning of plasma etching reactors: Ion energy distributions to walls and real-time and run-to-run control strategies Ankur Agarwala Department of Chemical and Biomolecular Engineering, University etching often depends on the conditioning of the inside surfaces of the reactor. Passivation of reactor

Kushner, Mark

15

Friction pressure drop measurements and flow distribution analysis for LEU conversion study of MIT Research Reactor  

E-Print Network [OSTI]

The MIT Nuclear Research Reactor (MITR) is the only research reactor in the United States that utilizes plate-type fuel elements with longitudinal fins to augment heat transfer. Recent studies on the conversion to low-enriched ...

Wong, Susanna Yuen-Ting

2008-01-01T23:59:59.000Z

16

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

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

Method for operating a nuclear reactor to accommodate load follow while maintaining a substantially constant axial power distribution  

SciTech Connect (OSTI)

This invention provides a method of operating a nuclear reactor having a negative reactivity moderator temperature coefficient with the object of maintaining a uniform and symmetric xenon distribution above and below substantially the center of the core over a substantial axial length of the core during normal reactor operation including load follow. In one embodiment variations in the xenon distribution are controlled by maintaining a substantially symmetric axial power distribution. The axial offset, which is employed as an indication of the axial power distribution, is maintained substantially equal to a target value , which is modified periodically to account for core burnup. A neutron absorbing element within the core coolant, or moderator, is employed to assist control of reactivity changes associated with changes in power, with the full-length control rods mainly employed to adjust variations in the axial power distribution while the part-length rodsremain completely withdrawn from the fuel region of the core. Rapid changes in reactivity are implemented, to accommodate corresponding changes in load, by a controlled reduction of the core coolant temperature. Thus, active core coolant temperature control is employed to control the reactivity of the core during load follow operation and effectively increase the spinning reserve capability of a power plant without altering the axial power distribution.

Mueller, N.P.; Rossi, C.E.; Scherpereel, L.R.

1980-09-16T23:59:59.000Z

19

Generic robot architecture  

DOE Patents [OSTI]

The present invention provides methods, computer readable media, and apparatuses for a generic robot architecture providing a framework that is easily portable to a variety of robot platforms and is configured to provide hardware abstractions, abstractions for generic robot attributes, environment abstractions, and robot behaviors. The generic robot architecture includes a hardware abstraction level and a robot abstraction level. The hardware abstraction level is configured for developing hardware abstractions that define, monitor, and control hardware modules available on a robot platform. The robot abstraction level is configured for defining robot attributes and provides a software framework for building robot behaviors from the robot attributes. Each of the robot attributes includes hardware information from at least one hardware abstraction. In addition, each robot attribute is configured to substantially isolate the robot behaviors from the at least one hardware abstraction.

Bruemmer, David J. (Idaho Falls, ID) [Idaho Falls, ID; Few, Douglas A. (Idaho Falls, ID) [Idaho Falls, ID

2010-09-21T23:59:59.000Z

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 (OSTI)

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

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

23

The power distribution and neutron fluence measurements and calculations in the VVER-1000 Mock-Up on the LR-0 research reactor  

SciTech Connect (OSTI)

The power density distribution in a reactor has significant influence on core structures and pressure vessel mechanical resistance, as well as on the physical characteristics of nuclear fuel. This quantity also has an effect on the leakage neutron and photon field. This issue has become of increasing importance, as it touches on actual questions of the VVER nuclear power plant life time extension. This paper shows the comparison of calculated and experimentally determined pin by pin power distributions. The calculation has been performed with deterministic and Monte Carlo approaches. This quantity is accompanied by the neutron and photon flux density calculation and measurements at different points of the light water zero-power (LR-0) research reactor mock-up core, reactor built-in component (core barrel), and reactor pressure vessel and model. The effect of the different data libraries used for calculation is discussed. (authors)

Kostal, M.; Juricek, V.; Rypar, V.; Svadlenkova, M. [Research Center Rez Ltd., 250 68 Husinec-Rez 130 (Czech Republic); Cvachovec, F. [Univ. of Defence, Kounicova 65, 662 10 Brno (Czech Republic)

2011-07-01T23:59:59.000Z

24

FHR Generic Design Criteria  

SciTech Connect (OSTI)

The purpose of this document is to provide an initial, focused reference to the safety characteristics of and a licensing approach for Fluoride-Salt-Cooled High-Temperature Reactors (FHRs). The document does not contain details of particular reactor designs nor does it attempt to identify or classify either design basis or beyond design basis accidents. Further, this document is an initial attempt by a small set of subject matter experts to document the safety and licensing characteristics of FHRs for a larger audience. The document is intended to help in setting the safety and licensing research, development, and demonstration path forward. Input from a wider audience, further technical developments, and additional study will be required to develop a consensus position on the safety and licensing characteristics of FHRs. This document begins with a brief overview of the attributes of FHRs and then a general description of their anticipated safety performance. Following this, an overview of the US nuclear power plant approval process is provided that includes both test and power reactors, as well as the role of safety standards in the approval process. The document next describes a General Design Criteria (GDC)–based approach to licensing an FHR and provides an initial draft set of FHR GDCs. The document concludes with a description of a path forward toward developing an FHR safety standard that can support both a test and power reactor licensing process.

Flanagan, G.F.; Holcomb, D.E.; Cetiner, S.M.

2012-06-15T23:59:59.000Z

25

Generic refinements for behavioral specifications   

E-Print Network [OSTI]

This thesis investigates the properties of generic refinements of behavioral specifications. At the base of this investigation stands the view from algebraic specification that abstract data types can be modeled as ...

Petria, Marius

2011-01-01T23:59:59.000Z

26

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

27

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

28

Descriptive Model of Generic WAMS  

SciTech Connect (OSTI)

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

29

Object and Reference Immutability using Java Generics  

E-Print Network [OSTI]

A compiler-checked immutability guarantee provides useful documentation, facilitates reasoning, and enables optimizations. This paper presents Immutability Generic Java (IGJ), a novel language extension that expresses ...

Zibin, Yoav

2007-03-16T23:59:59.000Z

30

Plasma instrumentation for fusion power reactor control  

SciTech Connect (OSTI)

Feedback control will be implemented in fusion power reactors to guard against unpredicted behavior of the plant and to assure desirable operation. In this study, plasma state feedback requirements for plasma control by systems strongly coupled to the plasma (magnet sets, RF, and neutral beam heating systems, and refueling systems) are estimated. Generic considerations regarding the impact of the power reactor environment on plasma instrumentation are outlined. Solutions are proposed to minimize the impact of the power reactor environment on plasma instrumentation. Key plasma diagnostics are evaluated with respect to their potential for upgrade and implementation as power reactor instruments.

Sager, G.T.; Bauer, J.F.; Maya, I.; Miley, G.H.

1985-07-01T23:59:59.000Z

31

Reactor Physics  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Reactor Physics Reactor and nuclear physics is a key area of research at INL. Much of the research done in reactor physics can be separated into one of three categories:...

32

Generic physical protection logic trees  

SciTech Connect (OSTI)

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

33

Comparative analysis of peat gasification reactor configuration  

SciTech Connect (OSTI)

A comparative analysis of two generic (fluidized and entrained beds) and two specific (PEATGAS and the Rockwell International) hydrogasifiers involved data-base assessment, regression analysis, a chemical-engineering evaluation of down- and upstream equipment needs, and computer simulation. The study concluded that the PEATGAS reactor is closer to commercialization than the Rockwell.

Not Available

1981-07-01T23:59:59.000Z

34

Evaluation of Generic EBS Design Concepts and Process Models...  

Broader source: Energy.gov (indexed) [DOE]

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...

35

Adaptive control of a generic hypersonic vehicle  

E-Print Network [OSTI]

This thesis presents a an adaptive augmented, gain-scheduled baseline LQR-PI controller applied to the Road Runner six-degree-of-freedom generic hypersonic vehicle model. Uncertainty in control effectiveness, longitudinal ...

Wiese, Daniel Philip

2013-01-01T23:59:59.000Z

36

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

SciTech Connect (OSTI)

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

37

Generic solar photovoltaic system dynamic simulation model specification.  

SciTech Connect (OSTI)

This document is intended to serve as a specification for generic solar photovoltaic (PV) system positive-sequence dynamic models to be implemented by software developers and approved by the WECC MVWG for use in bulk system dynamic simulations in accordance with NERC MOD standards. Two specific dynamic models are included in the scope of this document. The first, a Central Station PV System model, is intended to capture the most important dynamic characteristics of large scale (> 10 MW) PV systems with a central Point of Interconnection (POI) at the transmission level. The second, a Distributed PV System model, is intended to represent an aggregation of smaller, distribution-connected systems that comprise a portion of a composite load that might be modeled at a transmission load bus.

Ellis, Abraham; Behnke, Michael Robert; Elliott, Ryan Thomas

2013-10-01T23:59:59.000Z

38

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

39

Distribution of Correspondence  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

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

40

A Generic Distributed Broadcast Scheme in Ad Hoc Wireless Networks  

E-Print Network [OSTI]

-to-date") before the broadcast process or a reactive process (i.e., "on-the- fly") during the broadcast process

Wu, Jie

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

Sudoplatov S. V. ON GENERIC GROUP TRIGONOMETRIES  

E-Print Network [OSTI]

­447. 6. Sudoplatov S. V. On type identifications in trigonometrical theories // Materials of InternatSudoplatov S. V. ON GENERIC GROUP TRIGONOMETRIES The positive solution of known problem projective closure (the construction is defined in [5, theorem 8]); 2) an amalgamation (type identification

Sudoplatov, Sergey Vladimirovich

42

Geniaal programmeren Generic programming at Utrecht  

E-Print Network [OSTI]

Geniaal programmeren ­ Generic programming at Utrecht ­ Johan Jeuring Lambert Meertens Technical Report UU-CS-2009-001 January 2009 Department of Information and Computing Sciences Utrecht University, Utrecht, The Netherlands www.cs.uu.nl #12;ISSN: 0924-3275 Department of Information and Computing Sciences

Meertens, Lambert

43

Imaging Fukushima Daiichi reactors with muons  

SciTech Connect (OSTI)

A study of imaging the Fukushima Daiichi reactors with cosmic-ray muons to assess the damage to the reactors is presented. Muon scattering imaging has high sensitivity for detecting uranium fuel and debris even through thick concrete walls and a reactor pressure vessel. Technical demonstrations using a reactor mockup, detector radiation test at Fukushima Daiichi, and simulation studies have been carried out. These studies establish feasibility for the reactor imaging. A few months of measurement will reveal the spatial distribution of the reactor fuel. The muon scattering technique would be the best and probably the only way for Fukushima Daiichi to make this determination in the near future.

Miyadera, Haruo; Borozdin, Konstantin N.; Greene, Steve J.; Milner, Edward C.; Morris, Christopher L. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Lukic, Zarija [Computational Cosmology Center, Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States); Masuda, Koji [University of New Mexico, Albuquerque, NM 87131 (United States); Perry, John O. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); University of New Mexico, Albuquerque, NM 87131 (United States)

2013-05-15T23:59:59.000Z

44

Evolution of the Generic Lock System at Jefferson Lab  

SciTech Connect (OSTI)

The Generic Lock system is a software framework that allows highly flexible feedback control of large distributed systems. It allows system operators to implement new feedback loops between arbitrary process variables quickly and with no disturbance to the underlying control system. Several different types of feedback loops are provided and more are being added. This paper describes the further evolution of the system since it was first presented at ICALEPCS 2001 and reports on two years of successful use in accelerator operations. The framework has been enhanced in several key ways. Multiple-input, multiple-output (MIMO) lock types have been added for accelerator orbit and energy stabilization. The general purpose Proportional-Integral-Derivative (PID) locks can now be tuned automatically. The generic lock server now makes use of the Proxy IOC (PIOC) developed at Jefferson Lab to allow the locks to be monitored from any EPICS Channel Access aware client. (Previously clients had to be Cdev aware.) The dependency on the Qt XML parser has been replaced with the freely available Xerces DOM parser from the Apache project.

Brian Bevins; Yves Roblin

2003-10-13T23:59:59.000Z

45

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

46

Fusion reactor control  

SciTech Connect (OSTI)

The plasma kinetic temperature and density changes, each per an injected fuel density rate increment, control the energy supplied by a thermonuclear fusion reactor in a power production cycle. This could include simultaneously coupled control objectives for plasma current, horizontal and vertical position, shape and burn control. The minimum number of measurements required, use of indirect (not plasma parameters) system measurements, and distributed control procedures for burn control are to be verifiable in a time dependent systems code. The International Thermonuclear Experimental Reactor (ITER) has the need to feedback control both the fusion output power and the driven plasma current, while avoiding damage to diverter plates. The system engineering of fusion reactors must be performed to assure their development expeditiously and effectively by considering reliability, availability, maintainability, environmental impact, health and safety, and cost.

Plummer, D.A.

1995-12-31T23:59:59.000Z

47

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 (OSTI)

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

48

Automatic reactor power control for a pressurized water reactor  

SciTech Connect (OSTI)

An automatic reactor power control system is presented for a pressurized water reactor (PWR). The associated reactor control strategy is called mode K.' The new system implements a heavy-worth bank dedicated to axial shape control, independent of the existing regulating banks. The heavy bank provides a monotonic relationship between its motion and the axial shape change, which allows automatic control of the axial power distribution. Thus, the mode K enables precise regulation of both the reactivity and the power distribution, by using double closed-loop control of the reactor coolant temperature and the axial power difference. Automatic reactor power control permits the nuclear power plant to accommodate the load-follow operations, including frequency control, to respond to the grid requirements. The mode K reactor control concepts were tested using simulation responses of a Korean standardized 1,000-MW (electric) PWR. The simulation results illustrate that the mode K would be a practical reactor power control strategy for the increased automation of nuclear plants.

Jungin Choi (Kyungwon Univ. (Korea, Republic of)); Yungjoon Hah (Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of)); Unchul Lee (Seoul National Univ. (Korea, Republic of))

1993-05-01T23:59:59.000Z

49

Generic isolated horizons in loop quantum gravity  

E-Print Network [OSTI]

Isolated horizons model equilibrium states of classical black holes. A detailed quantization, starting from a classical phase space restricted to spherically symmetric horizons, exists in the literature and has since been extended to axisymmetry. This paper extends the quantum theory to horizons of arbitrary shape. Surprisingly, the Hilbert space obtained by quantizing the full phase space of \\textit{all} generic horizons with a fixed area is identical to that originally found in spherical symmetry. The entropy of a large horizon remains one quarter its area, with the Barbero-Immirzi parameter retaining its value from symmetric analyses. These results suggest a reinterpretation of the intrinsic quantum geometry of the horizon surface.

Christopher Beetle; Jonathan Engle

2010-07-16T23:59:59.000Z

50

Generic specificity of Aeromonas extracellular antigens  

E-Print Network [OSTI]

should be given a generic place in the family Pseudomonadaceae; and Habs and Schubert (9) hgg tdth t f fA. ~A'*ll 'd t th goyl'oo. A. ~tlt'h ly*tly been described (18, 19) and is classified according to two schools of thought: a separate species, A..., urine, bile, pus, spleen, pleural fluid, placenta, skin lesion, throat, and gall bladder (7). Nygaard, et. al. (21) have reported an increase in the number of Aeromonads submitted to the California State Department of Public Health...

Brinkley, Allen Wayne

1972-01-01T23:59:59.000Z

51

Generic turbine design study. Final report  

SciTech Connect (OSTI)

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

52

Nuclear reactor engineering  

SciTech Connect (OSTI)

Chapters are presented concerning energy from nuclear fission; nuclear reactions and radiations; diffusion and slowing-down of neutrons; principles of reactor analysis; nuclear reactor kinetics and control; energy removal; non-fuel reactor materials; the reactor fuel system; radiation protection and environmental effects; nuclear reactor shielding; nuclear reactor safety; and power reactor systems.

Glasstone, S.; Sesonske, A.

1981-01-01T23:59:59.000Z

53

Nuclear power plant Generic Aging Lessons Learned (GALL). Main report and appendix A  

SciTech Connect (OSTI)

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 document is Volume 1, consisting of the executive summary, summary and observations, and an appendix listing the GALL literature review tables.

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

1996-12-01T23:59:59.000Z

54

Research reactors - an overview  

SciTech Connect (OSTI)

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

55

Light Water Reactor Sustainability  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

3 Light Water Reactor Sustainability Program ACCOMPLISHMENTS REPORT 2013 Accomplishments Report | Light Water Reactor Sustainability 2 T he mission of the Light Water Reactor...

56

Light Water Reactor Sustainability  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (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...

57

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

58

Bioconversion reactor  

DOE Patents [OSTI]

A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

McCarty, Perry L. (Stanford, CA); Bachmann, Andre (Palo Alto, CA)

1992-01-01T23:59:59.000Z

59

Mitigative techniques and analysis of generic site conditions for ground-water contamination associated with severe accidents  

SciTech Connect (OSTI)

The purpose of this study is to evaluate the feasibility of using ground-water contaminant mitigation techniques to control radionuclide migration following a severe commercial nuclear power reactor accident. The two types of severe commercial reactor accidents investigated are: (1) containment basemat penetration of core melt debris which slowly cools and leaches radionuclides to the subsurface environment, and (2) containment basemat penetration of sump water without full penetration of the core mass. Six generic hydrogeologic site classifications are developed from an evaluation of reported data pertaining to the hydrogeologic properties of all existing and proposed commercial reactor sites. One-dimensional radionuclide transport analyses are conducted on each of the individual reactor sites to determine the generic characteristics of a radionuclide discharge to an accessible environment. Ground-water contaminant mitigation techniques that may be suitable, depending on specific site and accident conditions, for severe power plant accidents are identified and evaluated. Feasible mitigative techniques and associated constraints on feasibility are determined for each of the six hydrogeologic site classifications. The first of three case studies is conducted on a site located on the Texas Gulf Coastal Plain. Mitigative strategies are evaluated for their impact on contaminant transport and results show that the techniques evaluated significantly increased ground-water travel times. 31 references, 118 figures, 62 tables.

Shafer, J.M.; Oberlander, P.L.; Skaggs, R.L.

1984-04-01T23:59:59.000Z

60

Experiments with a Generic Dantzig-Wolfe Decomposition for ...  

E-Print Network [OSTI]

and-cut algorithms are essentially “blind.” While both ..... 2009. 6. G. Gamrath, Generic branch-cut-and-price, Master's thesis, Institut für Mathe- matik, Technische ...

2010-03-19T23:59:59.000Z

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.


61

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...

62

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 (OSTI)

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

63

Neutronic reactor  

DOE Patents [OSTI]

A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

Wende, Charles W. J. (Augusta, GA); Babcock, Dale F. (Wilmington, DE); Menegus, Robert L. (Wilmington, DE)

1983-01-01T23:59:59.000Z

64

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

65

australian generics policy: Topics by E-print Network  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

australian generics policy First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 A Generic Metamodel For...

66

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, this approach tries to provide a generic Modelica interface for embedding an FMU to be imported into a Modelica

Zhao, Yuxiao

67

Generic Argillite/Shale Disposal Reference Case  

SciTech Connect (OSTI)

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

68

Reactor pressure vessel. Status report  

SciTech Connect (OSTI)

This report describes the issues raised as a result of the staffs review of Generic Letter (GL) 92-01, Revision 1, responses and plant-specific reactor pressure vessel (RPV) assessments and the actions taken or work in progress to address these issues. In addition, the report describes actions taken by the staff and the nuclear industry to develop a thermal annealing process for use at U.S. commercial nuclear power plants. This process is intended to be used as a means of mitigating the effects of neutron radiation on the fracture toughness of RPV materials. The Nuclear Regulatory Commission (NRC) issued GL 92-01, Revision 1, Supplement 1, to obtain information needed to assess compliance with regulatory requirements and licensee commitments regarding RPV integrity. GL 92-01, Revision 1, Supplement 1, was issued as a result of generic issues that were raised in the NRC staff`s reviews of licensee responses to GL 92-01, Revision 1, and plant-specific RPV evaluations. In particular, an integrated review of all data submitted in response to GL 92-01, Revision 1, indicated that licensees may not have considered all relevant data in their RPV assessments. This report is representative of submittals to and evaluations by the staff as of September 30, 1996. An update of this report will be issued at a later date.

Elliot, B.J.; Hackett, E.M.; Lee, A.D. [and others

1996-10-01T23:59:59.000Z

69

Risk analysis of highly combustible gas storage, supply, and distribution systems in PWR plants  

SciTech Connect (OSTI)

This report presents the evaluation of the potential safety concerns for pressurized water reactors (PWRs) identified in Generic Safety Issue 106, Piping and the Use of Highly Combustible Gases in Vital Areas. A Westinghouse four-loop PWR plant was analyzed for the risk due to the use of combustible gases (predominantly hydrogen) within the plant. The analysis evaluated an actual hydrogen distribution configuration and conducted several sensitivity studies to determine the potential variability among PWRs. The sensitivity studies were based on hydrogen and safety-related equipment configurations observed at other PWRs within the United States. Several options for improving the hydrogen distribution system design were identified and evaluated for their effect on risk and core damage frequency. A cost/benefit analysis was performed to determine whether alternatives considered were justifiable based on the safety improvement and economics of each possible improvement.

Simion, G.P. [Science Applications International Corp., Albuquerque, NM (United States); VanHorn, R.L.; Smith, C.L.; Bickel, J.H.; Sattison, M.B. [EG and G Idaho, Inc., Idaho Falls, ID (United States); Bulmahn, K.D. [SCIENTECH, Inc., Idaho Falls, ID (United States)

1993-06-01T23:59:59.000Z

70

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

71

Bacterial Community Structure in Geographically Distributed Biological  

E-Print Network [OSTI]

Bacterial Community Structure in Geographically Distributed Biological Wastewater Treatment of the microbial communities within biological wastewater treatment reactors is incomplete due to limitations microbial community composition in five biological wastewater treatment reactors in China and the United

Alvarez-Cohen, Lisa

72

Knowledge and abilities catalog for nuclear power plant operators: Pressurized water reactors. Revision 1  

SciTech Connect (OSTI)

This document provides the basis for the development of content-valid licensing examinations for reactor operators and senior reactor operators. The examinations developed using the PWR catalog will cover those topics listed under Title 10, (ode of Federal Regulations Part 55. The PWR catalog contains approximately 5100 knowledge and ability (K/A) statements for reactor operators and senior reactor operators. The catalog is organized into six major sections: Catalog Organization; Generic Knowledge and Abilities; Plant Systems; Emergency and Abnormal Plant Evolutions; Components and Theory.

NONE

1995-08-01T23:59:59.000Z

73

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

74

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

75

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

76

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

77

An assessment of space reactor technology needs and recommendations for development  

SciTech Connect (OSTI)

In order to provide a strategy for space reactor technology development, the Defense Nuclear Agency (DNA) has authorized a brief review of potential national needs that may be addressed by space reactor systems. a systematic approach was used to explore needs at several levels that are increasingly specific. Level 0 -- general trends and issues; Level 1 -- generic space capabilities to address trends; Level 2 -- requirements to support capabilities; Level 3 -- system types capable of meeting requirements; Level 4 --generic reactor system types; and Level 5 -- specific baseline systems. Using these findings, a strategy was developed to support important space reactor technologies within a limited budget. A preliminary evaluation identified key technical issues and provide a prioritized set of candidate research projects. The evaluation of issues and the recommended research projects are presented in a companion paper.

Marshall, A.C. [Sandia National Labs., Albuquerque, NM (United States); Wiley, R.L. [Consultant, Columbia, MD (United States)

1995-11-01T23:59:59.000Z

78

Nuclear reactor engineering  

SciTech Connect (OSTI)

A book is reviewed which emphasizes topics directly related to the light water reactor power plant and the fast reactor power system. Current real-world problems are addressed throughout the text, and a chapter on safety includes much of the postThree Mile Island impact on operating systems. Topics covered include Doppler broadening, neutron resonances, multigroup diffusion theory, reactor kinetics, reactor control, energy removal, nonfuel materials, reactor fuel, radiation protection, environmental effects, and reactor safety.

Glasstone, S.; Sesonske, A.

1982-07-01T23:59:59.000Z

79

Testing of SP-100 reactor control approaches in the NAT  

SciTech Connect (OSTI)

The Generic Flight System (GFS) reactor control approach will be tested at the SP-100 Ground Engineering System (GES) test site as part of the Nuclear Assembly Test (NAT) program. A control scheme for the NAT has been developed to emulate the GFS conditions for testing of the GFS reactor control approach in the NAT. Comparisons between the GFS simulation results and the NAT simulation results show that a reasonably good emulation of the GFS conditions can be achieved in the NAT using the control scheme described in this paper.

Rhow, S.K.; Meyer, R.A.; Wong, K.K.; Halfen, F.J. (General Electric Company, SP-100 Programs, Astro Space Division, San Jose, California 95153-5354 (US))

1991-01-01T23:59:59.000Z

80

Reactor safety method  

DOE Patents [OSTI]

This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

Vachon, Lawrence J. (Clairton, PA)

1980-03-11T23:59:59.000Z

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.


81

SRS Small Modular Reactors  

SciTech Connect (OSTI)

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

None

2012-04-27T23:59:59.000Z

82

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

83

Radiation effects on reactor pressure vessel supports  

SciTech Connect (OSTI)

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

84

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

85

Parallel Monte Carlo reactor neutronics  

SciTech Connect (OSTI)

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

86

Spectral shift reactor control method  

SciTech Connect (OSTI)

A method of operating a pressurized water nuclear reactor is described which comprises the determining of the present core power and reactivity levels and predicting the change in such levels due to displacer rod movements. Groups or single clusters of displacer rods can be inserted or withdrawn based on the predicted core power and reactivity levels to change the core power level and power distribution thereby providing load follow capability, without changing control rod positions or coolant boron concentrations.

Impink, A.J. Jr.

1984-02-21T23:59:59.000Z

87

Generic repository design concepts and thermal analysis (FY11).  

SciTech Connect (OSTI)

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

88

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

SciTech Connect (OSTI)

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

89

Principles of Distributed Data Management in 2020?  

E-Print Network [OSTI]

With the advents of high-speed networks, fast commodity hardware, and the web, distributed data sources have become ubiquitous. The third edition of the \\"Ozsu-Valduriez textbook Principles of Distributed Database Systems [10] reflects the evolution of distributed data management and distributed database systems. In this new edition, the fundamental principles of distributed data management could be still presented based on the three dimensions of earlier editions: distribution, heterogeneity and autonomy of the data sources. In retrospect, the focus on fundamental principles and generic techniques has been useful not only to understand and teach the material, but also to enable an infinite number of variations. The primary application of these generic techniques has been obviously for distributed and parallel DBMS versions. Today, to support the requirements of important data-intensive applications (e.g. social networks, web data analytics, scientific applications, etc.), new distributed data management tech...

Valduriez, Patrick

2011-01-01T23:59:59.000Z

90

Development and application of the dynamic system doctor to nuclear reactor probabilistic risk assessments.  

SciTech Connect (OSTI)

This LDRD project has produced a tool that makes probabilistic risk assessments (PRAs) of nuclear reactors - analyses which are very resource intensive - more efficient. PRAs of nuclear reactors are being increasingly relied on by the United States Nuclear Regulatory Commission (U.S.N.R.C.) for licensing decisions for current and advanced reactors. Yet, PRAs are produced much as they were 20 years ago. The work here applied a modern systems analysis technique to the accident progression analysis portion of the PRA; the technique was a system-independent multi-task computer driver routine. Initially, the objective of the work was to fuse the accident progression event tree (APET) portion of a PRA to the dynamic system doctor (DSD) created by Ohio State University. Instead, during the initial efforts, it was found that the DSD could be linked directly to a detailed accident progression phenomenological simulation code - the type on which APET construction and analysis relies, albeit indirectly - and thereby directly create and analyze the APET. The expanded DSD computational architecture and infrastructure that was created during this effort is called ADAPT (Analysis of Dynamic Accident Progression Trees). ADAPT is a system software infrastructure that supports execution and analysis of multiple dynamic event-tree simulations on distributed environments. A simulator abstraction layer was developed, and a generic driver was implemented for executing simulators on a distributed environment. As a demonstration of the use of the methodological tool, ADAPT was applied to quantify the likelihood of competing accident progression pathways occurring for a particular accident scenario in a particular reactor type using MELCOR, an integrated severe accident analysis code developed at Sandia. (ADAPT was intentionally created with flexibility, however, and is not limited to interacting with only one code. With minor coding changes to input files, ADAPT can be linked to other such codes.) The results of this demonstration indicate that the approach can significantly reduce the resources required for Level 2 PRAs. From the phenomenological viewpoint, ADAPT can also treat the associated epistemic and aleatory uncertainties. This methodology can also be used for analyses of other complex systems. Any complex system can be analyzed using ADAPT if the workings of that system can be displayed as an event tree, there is a computer code that simulates how those events could progress, and that simulator code has switches to turn on and off system events, phenomena, etc. Using and applying ADAPT to particular problems is not human independent. While the human resources for the creation and analysis of the accident progression are significantly decreased, knowledgeable analysts are still necessary for a given project to apply ADAPT successfully. This research and development effort has met its original goals and then exceeded them.

Kunsman, David Marvin; Aldemir, Tunc (Ohio State University); Rutt, Benjamin (Ohio State University); Metzroth, Kyle (Ohio State University); Catalyurek, Umit (Ohio State University); Denning, Richard (Ohio State University); Hakobyan, Aram (Ohio State University); Dunagan, Sean C.

2008-05-01T23:59:59.000Z

91

A Generic Database Benchmarking Service Martin Kaufmann #1  

E-Print Network [OSTI]

of a particular kind of benchmark. Examples of such frameworks include SIMS [4] (local resources), OLTP- BenchmarkA Generic Database Benchmarking Service Martin Kaufmann #1 , Peter M. Fischer2 , Donald Kossmann #3, Walldorf, Germany 4 norman.may@sap.com Abstract--Benchmarks are widely applied for the development

Mannheim, Universität

92

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

93

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

94

A Generic Approach to Simplification of Geodata for Mobile Applications  

E-Print Network [OSTI]

, tourism, security and disaster management. All mobile applications have characteristics in commonA Generic Approach to Simplification of Geodata for Mobile Applications Theodor Foerster¹, Jantien,stoter,kobben}@itc.nl ² Delft University of Technology, OTB, Section GIS-technology P.O. Box 5030, 2600 GA Delft

Köbben, Barend

95

A Generic Arc-Consistency Algorithm and its Specializations1  

E-Print Network [OSTI]

not take into account the semantics of constraints. In this paper, we present a new generic arc satisfaction problems (CSP). A CSP is de ned by a nite set of variables taking values from nite domains of applications include graph-coloring, warehouse locations, car-sequencing and cutting stock (see for instance 4

Deville, Yves

96

A Generic ArcConsistency Algorithm and its Specializations 1  

E-Print Network [OSTI]

, these algorithms do not take into account the semantics of constraints. In this paper, we present a new generic arc as constraint satisfaction problems (CSP). A CSP is defined by a finite set of variables taking values from. Examples of applications include graph­coloring, warehouse locations, car­sequencing and cutting stock (see

Deville, Yves

97

A Generic Framework for Interprocedural Analysis of Numerical Properties  

E-Print Network [OSTI]

A Generic Framework for Interprocedural Analysis of Numerical Properties Markus M¨uller-Olm1, Germany markus.mueller-olm@cs.uni-dortmund.de 2 TU M¨unchen, Institut f¨ur Informatik, I2 80333 M of the abstracted collecting semantics is also used by M¨uller-Olm and Seidl in [12] where (in absence of equality

Müller-Olm, Markus

98

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.

99

Supporting Generic Cost Models for Wide-Area Stream Processing  

E-Print Network [OSTI]

Supporting Generic Cost Models for Wide-Area Stream Processing Olga Papaemmanouil #1 , Ugur C¸ etintemel 2 , John Jannotti 2 # Deparment of Computer Science, Brandeis University, Waltham, MA, USA 1 olga and increased availability of receptors that report physical or software events has led to the emergence

Cetintemel, Ugur

100

A Generic WebBased Network Management Application  

E-Print Network [OSTI]

of network manage­ ment applications that eliminates most of the tedious work, and reduces development costsA Generic Web­Based Network Management Application Yehuda Afek \\Lambda Yoram Cohen y Computer to network management is introduced. In traditional network management each management application

Afek, Yehuda

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

Modular and Generic Control Software System for Scalable Automation  

E-Print Network [OSTI]

Modular and Generic Control Software System for Scalable Automation Christian Brecher, Martin.freundt@ipt.fraunhofer.de Abstract. The development of automated production systems is subdivided in two mayor tasks. One production with a high rate of changes, this is why fully automated solutions don't pay off and manual

Paris-Sud XI, Université de

102

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

103

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.

104

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

105

A Generic Grid Interface for Parallel and Adaptive Scientific Computing.  

E-Print Network [OSTI]

A Generic Grid Interface for Parallel and Adaptive Scientific Computing. Part I: Abstract Framework definition of a grid for al- gorithms solving partial differential equations. Unlike previous ap- proaches [2, 3], our grids have a hierarchical structure. This makes them suitable for geometric multigrid

Kornhuber, Ralf

106

Basis for NGNP Reactor Design Down-Selection  

SciTech Connect (OSTI)

The purpose of this paper is to identify the extent of technology development, design and licensing maturity anticipated to be required to credibly identify differences that could make a technical choice practical between the prismatic and pebble bed reactor designs. This paper does not address a business decision based on the economics, business model and resulting business case since these will vary based on the reactor application. The selection of the type of reactor, the module ratings, the number of modules, the configuration of the balance of plant and other design selections will be made on the basis of optimizing the Business Case for the application. These are not decisions that can be made on a generic basis.

L.E. Demick

2010-08-01T23:59:59.000Z

107

Basis for NGNP Reactor Design Down-Selection  

SciTech Connect (OSTI)

The purpose of this paper is to identify the extent of technology development, design and licensing maturity anticipated to be required to credibly identify differences that could make a technical choice practical between the prismatic and pebble bed reactor designs. This paper does not address a business decision based on the economics, business model and resulting business case since these will vary based on the reactor application. The selection of the type of reactor, the module ratings, the number of modules, the configuration of the balance of plant and other design selections will be made on the basis of optimizing the Business Case for the application. These are not decisions that can be made on a generic basis.

L.E. Demick

2011-11-01T23:59:59.000Z

108

Undergraduate reactor control experiment  

SciTech Connect (OSTI)

A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise.

Edwards, R.M.; Power, M.A.; Bryan, M. (Pennsylvania State Univ., University Park (United States))

1992-01-01T23:59:59.000Z

109

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.

Scott, Charles D. (Oak Ridge, TN); Davison, Brian H. (Knoxvile, TN)

1993-01-01T23:59:59.000Z

110

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

111

Reactor Sharing Program  

SciTech Connect (OSTI)

Progress achieved at the University of Florida Training Reactor (UFTR) facility through the US Department of Energy's University Reactor Sharing Program is reported for the period of 1991--1992.

Vernetson, W.G.

1993-01-01T23:59:59.000Z

112

Assessment of industrial attitudes toward generic research needs in tribology  

SciTech Connect (OSTI)

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

Savannah River Site production reactor technical specifications. K Production Reactor  

SciTech Connect (OSTI)

These technical specifications are explicit restrictions on the operation of the Savannah River Site K Production Reactor. They are designed to preserve the validity of the plant safety analysis by ensuring that the plant is operated within the required conditions bounded by the analysis, and with the operable equipment that is assumed to mitigate the consequences of an accident. Technical specifications preserve the primary success path relied upon to detect and respond to accidents. This report describes requirements on thermal-hydraulic limits; limiting conditions for operation and surveillance for the reactor, power distribution control, instrumentation, process water system, emergency cooling and emergency shutdown systems, confinement systems, plant systems, electrical systems, components handling, and special test exceptions; design features; and administrative controls.

NONE

1996-02-01T23:59:59.000Z

114

High solids fermentation reactor  

DOE Patents [OSTI]

A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

1993-03-02T23:59:59.000Z

115

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

116

Advanced Test Reactor Tour  

SciTech Connect (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

2011-01-01T23:59:59.000Z

117

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

118

High solids fermentation reactor  

DOE Patents [OSTI]

A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

Wyman, Charles E. (Lakewood, CO); Grohmann, Karel (Littleton, CO); Himmel, Michael E. (Littleton, CO); Richard, Christopher J. (Lakewood, CO)

1993-01-01T23:59:59.000Z

119

Hypothetical Reactor Accident Study  

E-Print Network [OSTI]

- W 4 DfcSkoollo Rise-R-427 CARNSORE: Hypothetical Reactor Accident Study O. Walmod-Larsen, N. O: HYPOTHETICAL REACTOR ACCIDENT STUDY O. Walmod-Larsen, N.O. Jensen, L. Kristensen, A. Heide, K.L. Nedergård, P-basis accident and a series of hypothetical core-melt accidents to a 600 MWe reactor are de- scribed

120

Reduced-Enrichment Research and Test Reactor Program: Environmental assessment  

SciTech Connect (OSTI)

The principal program objective and principal part of the proposed action is to improve the proliferation resistance of nuclear fuels used in research and test reactors by providing the technical means (through technical development, design, and testing) for reducing the uranium enrichment requirements of these fuels to substantially less than the 90 to 93% enrichment currently used. Operator acceptance of the reduced-enrichment-uranium (REU) fuel alternative will require minimizing of reactor performance reduction, fuel cycle cost increases, the number of new safety and licensing issues raised, and reactor and facility modifications. The other part of the proposed action is to assure the capability for commercial production and supply of REU fuel for use both in the US and abroad. The RERTR Program scope is limited to generic design studies, technical support to reactor operating organizations in preparing for conversions to REU fuels, fuel development, fuel demonstrations, and technical support for commercialization of REU fuels. This environmental assessment addresses the environmental consequences of RERTR Program activities and of specific conversions of typical reactors (the Ford Nuclear Reactor and one or two other to-be-designated demonstrations) to REU-fuel cycles, including domestic and international shipments of enriched uranium pertinent to the conduct of RERTR Program activities.

Not Available

1980-05-01T23:59:59.000Z

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

Neutron behavior, reactor control, and reactor heat transfer. Volume four  

SciTech Connect (OSTI)

Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant).

Not Available

1986-01-01T23:59:59.000Z

122

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

123

Spinning fluids reactor  

DOE Patents [OSTI]

A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

Miller, Jan D; Hupka, Jan; Aranowski, Robert

2012-11-20T23:59:59.000Z

124

Determining Reactor Neutrino Flux  

E-Print Network [OSTI]

Flux is an important source of uncertainties for a reactor neutrino experiment. It is determined from thermal power measurements, reactor core simulation, and knowledge of neutrino spectra of fuel isotopes. Past reactor neutrino experiments have determined the flux to (2-3)% precision. Precision measurements of mixing angle $\\theta_{13}$ by reactor neutrino experiments in the coming years will use near-far detector configurations. Most uncertainties from reactor will be canceled out. Understanding of the correlation of uncertainties is required for $\\theta_{13}$ experiments. Precise determination of reactor neutrino flux will also improve the sensitivity of the non-proliferation monitoring and future reactor experiments. We will discuss the flux calculation and recent progresses.

Jun Cao

2012-03-08T23:59:59.000Z

125

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

126

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

127

E-Print Network 3.0 - ajustable para reactores Sample Search...  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Summary: diagnostic measurements of acetylene absorption in hot-filament and DC-arc plasma jet reactors operating... deposition. The acetylene was widely4 2 distributed...

128

Generic Structures in Parameter Space and Ratchet Transport  

E-Print Network [OSTI]

This work reports the existence of Isoperiodic Stable Ratchet Transport Structures in the parameter spaces dissipation versus spatial asymmetry and versus phase of a ratchet model. Such structures were found [Phys. Rev. Lett. 106, 234101 (2011)] in the parameter space dissipation versus amplitude of the ratchet potential and they appear to have generic shapes and to align themselves along preferred directions in the parameter space. Since the ratchet current is usually larger inside these structures, this allows us to make general statements about the relevant parameters combination to obtain an efficient ratchet current. Results of the present work give further evidences of the suggested generic properties of the isoperiodic stable structures in the context of ratchet transport.

Alan Celestino; Cesar Manchein; Holokx A. Albuquerque; Marcus W. Beims

2011-11-06T23:59:59.000Z

129

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

130

A Generic Framework for Interprocedural Analysis of Numerical Properties  

E-Print Network [OSTI]

A Generic Framework for Interprocedural Analysis of Numerical Properties Markus MË?uller­Olm 1, Germany markus.mueller­olm@cs.uni­dortmund.de 2 TU MË?unchen, Institut fË?ur Informatik, I2 80333 M of the abstracted collecting semantics is also used by MË?uller­Olm and Seidl in [12] where (in absence of equality

Seidl, Helmut

131

Developing Generic Dynamic Models for the 2030 Eastern Interconnection Grid  

SciTech Connect (OSTI)

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

132

An introduction to InP-based generic integration technology  

E-Print Network [OSTI]

, ring lasers and, when used in combination with a wavelength tunable reflec- tor, also tunable lasers. By using a short SOA section in reverse bias as a saturable absorber we can make picosecond pulse lasers. And a SOA in reverse bias can also be used... multiplexers. And by making use of the nonlinear properties of SOAs integrated in an MZI we can make ultrafast switches. An advantage of generic integration technologies is that because they serve many different applications, they justify the investments...

Smit, Meint; Leijtens, Xaveer; Ambrosius, Huub; Bente, Erwin; van der Tol, Jos; Smalbrugge, Barry; de Vries, Tjibbe; Geluk, Erik-Jan; Bolk, Jeroen; van Veldhoven, Rene; Augustin, Luc; Thijs, Peter; D’Agostino, Domenico; Rabbani, Hadi; Lawniczuk, Katarzyna; Stopinski, Stanislaw; Tahvili, Saeed; Corradi, Antonio; Kleijn, Emil; Dzibrou, Dzmitry; Felicetti, Manuela; Bitincka, Elton; Moskalenko, Valentina; Zhao, Jing; Santos, Rui; Gilardi, Giovanni; Yao, Weiming; Williams, Kevin; Stabile, Patty; Kuindersma, Piet; Pello, Josselin; Bhat, Srivathsa; Jiao, Yuqing; Heiss, Dominik; Roelkens, Gunther; Wale, Mike; Firth, Paul; Soares, Francisco; Grote, Norbert; Schell, Martin; Debregeas, Helene; Achouche, Mohand; Gentner, Jean-Louis; Bakker, Arjen; Korthorst, Twan; Gallagher, Dominic; Dabbs, Andrew; Melloni, Andrea; Morichetti, Francesco; Melati, Daniele; Wonfor, Adrian; Penty, Richard; Broeke, Ronald; Musk, Bob; Robbins, Dave

2014-06-13T23:59:59.000Z

133

Cooling water distribution system  

DOE Patents [OSTI]

A passive containment cooling system for a nuclear reactor containment vessel. Disclosed is a cooling water distribution system for introducing cooling water by gravity uniformly over the outer surface of a steel containment vessel using an interconnected series of radial guide elements, a plurality of circumferential collector elements and collector boxes to collect and feed the cooling water into distribution channels extending along the curved surface of the steel containment vessel. The cooling water is uniformly distributed over the curved surface by a plurality of weirs in the distribution channels.

Orr, Richard (Pittsburgh, PA)

1994-01-01T23:59:59.000Z

134

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

SciTech Connect (OSTI)

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

135

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

136

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

137

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

138

Pressurized fluidized bed reactor  

DOE Patents [OSTI]

A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

Isaksson, J.

1996-03-19T23:59:59.000Z

139

Pressurized fluidized bed reactor  

DOE Patents [OSTI]

A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.

Isaksson, Juhani (Karhula, FI)

1996-01-01T23:59:59.000Z

140

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

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

Next Generation Reactors  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Nuclear Advances We are coordinating the Generation IV Nuclear Systems Initiative - an international effort to develop the next generation of nuclear power reactors. Skip...

142

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

143

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....

144

Brookhaven Graphite Research Reactor Workshop  

Broader source: Energy.gov [DOE]

The Brookhaven Graphite Research Reactor (BGRR) was the first reactor built in the U.S. for peacetime atomic research following World War II.  Construction began in 1947 and the reactor started...

145

Portfolio for fast reactor collaboration  

SciTech Connect (OSTI)

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

146

Generic Quantum Ratchet Accelerator with Full Classical Chaos  

E-Print Network [OSTI]

A simple model of quantum ratchet transport that can generate unbounded linear acceleration of the quantum ratchet current is proposed, with the underlying classical dynamics fully chaotic. The results demonstrate that generic acceleration of quantum ratchet transport can occur with any type of classical phase space structure. The quantum ratchet transport with full classical chaos is also shown to be very robust to noise due to the large linear acceleration afforded by the quantum dynamics. One possible experiment allowing observation of these predictions is suggested.

Jiangbin Gong; Paul Brumer

2006-09-05T23:59:59.000Z

147

Generic protocol for environmental audits at federal facilities  

SciTech Connect (OSTI)

The Generic Protocol is a supplement to earlier efforts by EPA to provide technical assistance to Federal agencies on environmental auditing. EPA`s Office of Federal Activities (OFA) issued the Federal Facility Compliance Strategy (referred to as the `Yellow Book`), prepared in 1984 and revised in 1988, as a framework for EPA`s media programs to follow to ensure that Federal facilities are fully integrated into Federal and State compliance monitoring and enforcement activities. The policy encourages Federal agencies to voluntarily develop comprehensive auditing programs to ensure environmental compliance.

NONE

1989-08-01T23:59:59.000Z

148

Generic copy of DOE's IDIQ ESPC contract | Department of Energy  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergyENERGY TAX POLICIES7.pdfFuel Cell Vehicle BasicsWashers |DepartmentGeishaGeneralDeepGeneric

149

REACTOR OPERATIONS AND CONTROL  

E-Print Network [OSTI]

REACTOR OPERATIONS AND CONTROL KEYWORDS: core calculations, neural networks, control rod elevation of a control rod, or a group of control rods, is an important parameter from the viewpoint of reactor control DETERMINATION OF PWR CONTROL ROD POSITION BY CORE PHYSICS AND NEURAL NETWORK METHODS NINOS S. GARIS* and IMRE

Pázsit, Imre

150

NPR (New Production Reactor) capacity cost evaluation  

SciTech Connect (OSTI)

The ORNL Cost Evaluation Technical Support Group (CETSG) has been assigned by DOE-HQ Defense Programs (DP) the task defining, obtaining, and evaluating the capital and life-cycle costs for each of the technology/proponent/site/revenue possibilities envisioned for the New Production Reactor (NPR). The first part of this exercise is largely one of accounting, since all NPR proponents use different accounting methodologies in preparing their costs. In order to address this problem of comparing ''apples and oranges,'' the proponent-provided costs must be partitioned into a framework suitable for all proponents and concepts. If this is done, major cost categories can then be compared between concepts and major cost differences identified. Since the technologies proposed for the NPR and its needed fuel and target support facilities vary considerably in level of technical and operational maturity, considerable care must be taken to evaluate the proponent-derived costs in an equitable manner. The use of cost-risk analysis along with derivation of single point or deterministic estimates allows one to take into account these very real differences in technical and operational maturity. Chapter 2 summarizes the results of this study in tabular and bar graph form. The remaining chapters discuss each generic reactor type as follows: Chapter 3, LWR concepts (SWR and WNP-1); Chapter 4, HWR concepts; Chapter 5, HTGR concept; and Chapter 6, LMR concept. Each of these chapters could be a stand-alone report. 39 refs., 36 figs., 115 tabs.

none,

1988-07-01T23:59:59.000Z

151

Reactor & Nuclear Systems Publications | ORNL  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

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

152

Reed Reactor Facility. Final report  

SciTech Connect (OSTI)

This report discusses the operation and maintenance of the Reed Reactor Facility. The Reed reactor is mostly used for education and train purposes.

Frantz, S.G.

1994-12-31T23:59:59.000Z

153

Evaluation of proposed German safety criteria for high-temperature gas-cooled reactors  

SciTech Connect (OSTI)

This work reviews proposed safety criteria prepared by the German Bundesministerium des Innern (BMI) for future licensing of gas-cooled high-temperature reactor (HTR) concepts in the Federal Republic of Germany. Comparison is made with US General Design Criteria (GDCs) in 10CFR50 Appendix A and with German light water reactor (LWR) criteria. Implications for the HTR design relative to the US design and safety approach are indicated. Both inherent characteristics and design features of the steam cycle, gas turbine, and process heat concepts are taken into account as well as generic design options such as a pebble bed or prismatic core.

Barsell, A.W.

1980-05-01T23:59:59.000Z

154

Monte Carlo simulations for generic granite repository studies  

SciTech Connect (OSTI)

In a collaborative study between Los Alamos National Laboratory (LANL) and Sandia National Laboratories (SNL) for the DOE-NE Office of Fuel Cycle Technologies Used Fuel Disposition (UFD) Campaign project, we have conducted preliminary system-level analyses to support the development of a long-term strategy for geologic disposal of high-level radioactive waste. A general modeling framework consisting of a near- and a far-field submodel for a granite GDSE was developed. A representative far-field transport model for a generic granite repository was merged with an integrated systems (GoldSim) near-field model. Integrated Monte Carlo model runs with the combined near- and farfield transport models were performed, and the parameter sensitivities were evaluated for the combined system. In addition, a sub-set of radionuclides that are potentially important to repository performance were identified and evaluated for a series of model runs. The analyses were conducted with different waste inventory scenarios. Analyses were also conducted for different repository radionuelide release scenarios. While the results to date are for a generic granite repository, the work establishes the method to be used in the future to provide guidance on the development of strategy for long-term disposal of high-level radioactive waste in a granite repository.

Chu, Shaoping [Los Alamos National Laboratory; Lee, Joon H [SNL; Wang, Yifeng [SNL

2010-12-08T23:59:59.000Z

155

Regulatory analysis for the resolution of Generic Issue 143: Availability of chilled water system and room cooling  

SciTech Connect (OSTI)

This report presents the regulatory analysis for Generic Issue (GI-143), {open_quotes}Availability of Chilled Water System and Room Cooling.{close_quotes} The heating, ventilating, and air conditioning (HVAC) systems and related auxiliaries are required to provide control of environmental conditions in areas in light water reactor (LWR) plants that contain safety-related equipment. In some plants, the HVAC and chilled water systems serve to maintain a suitable environment for both safety and non-safety-related areas. Although some plants have an independent chilled water system for the safety-related areas, the heat removal capability often depends on the operability of other supporting systems such as the service water system or the component cooling water system. The operability of safety-related components depends upon operation of the HVAC and chilled water systems to remove heat from areas containing the equipment. If cooling to dissipate the heat generated is unavailable, the ability of the safety-related equipment to operate as intended cannot be assured. Typical components or areas in the nuclear power plant that could be affected by the failure of cooling from HVAC or chilled water systems include the (1) emergency switchgear and battery rooms, (2) emergency diesel generator room, (3) pump rooms for residual heat removal, reactor core isolation cooling, high-pressure core spray, and low-pressure core spray, and (4) control room. The unavailability of such safety-related equipment or areas could cause the core damage frequency (CDF) to increase significantly.

Leung, V.T.

1993-12-01T23:59:59.000Z

156

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

SciTech Connect (OSTI)

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

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

158

Nuclear reactor control column  

SciTech Connect (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 crosssectional 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, D.M.

1982-08-10T23:59:59.000Z

159

Reactor Safety Research Programs  

SciTech Connect (OSTI)

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

160

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

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

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. 12 figs.

Hopkins, R.J.; Land, J.T.; Misvel, M.C.

1994-06-07T23:59:59.000Z

162

Fast Breeder Reactor studies  

SciTech Connect (OSTI)

This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

1980-07-01T23:59:59.000Z

163

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

164

Microfluidic electrochemical reactors  

DOE Patents [OSTI]

A microfluidic electrochemical reactor includes an electrode and one or more microfluidic channels on the electrode, where the microfluidic channels are covered with a membrane containing a gas permeable polymer. The distance between the electrode and the membrane is less than 500 micrometers. The microfluidic electrochemical reactor can provide for increased reaction rates in electrochemical reactions using a gaseous reactant, as compared to conventional electrochemical cells. Microfluidic electrochemical reactors can be incorporated into devices for applications such as fuel cells, electrochemical analysis, microfluidic actuation, pH gradient formation.

Nuzzo, Ralph G. (Champaign, IL); Mitrovski, Svetlana M. (Urbana, IL)

2011-03-22T23:59:59.000Z

165

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

166

Stochastic solution of population balance equations for reactor networks  

SciTech Connect (OSTI)

This work presents a sequential modular approach to solve a generic network of reactors with a population balance model using a stochastic numerical method. Full-coupling to the gas-phase is achieved through operator-splitting. The convergence of the stochastic particle algorithm in test networks is evaluated as a function of network size, recycle fraction and numerical parameters. These test cases are used to identify methods through which systematic and statistical error may be reduced, including by use of stochastic weighted algorithms. The optimal algorithm was subsequently used to solve a one-dimensional example of silicon nanoparticle synthesis using a multivariate particle model. This example demonstrated the power of stochastic methods in resolving particle structure by investigating the transient and spatial evolution of primary polydispersity, degree of sintering and TEM-style images. Highlights: •An algorithm is presented to solve reactor networks with a population balance model. •A stochastic method is used to solve the population balance equations. •The convergence and efficiency of the reported algorithms are evaluated. •The algorithm is applied to simulate silicon nanoparticle synthesis in a 1D reactor. •Particle structure is reported as a function of reactor length and time.

Menz, William J.; Akroyd, Jethro; Kraft, Markus

2014-01-01T23:59:59.000Z

167

Shielded fluid stream injector for particle bed reactor  

DOE Patents [OSTI]

A shielded fluid-stream injector assembly is provided for particle bed reactors. The assembly includes a perforated pipe injector disposed across the particle bed region of the reactor and an inverted V-shaped shield placed over the pipe, overlapping it to prevent descending particles from coming into direct contact with the pipe. The pipe and shield are fixedly secured at one end to the reactor wall and slidably secured at the other end to compensate for thermal expansion. An axially extending housing aligned with the pipe and outside the reactor and an in-line reamer are provided for removing deposits from the inside of the pipe. The assembly enables fluid streams to be injected and distributed uniformly into the particle bed with minimized clogging of injector ports. The same design may also be used for extraction of fluid streams from particle bed reactors.

Notestein, John E. (Morgantown, WV)

1993-01-01T23:59:59.000Z

168

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

SciTech Connect (OSTI)

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

169

Generic CSP Performance Model for NREL's System Advisor Model: Preprint  

SciTech Connect (OSTI)

The suite of concentrating solar power (CSP) modeling tools in NREL's System Advisor Model (SAM) includes technology performance models for parabolic troughs, power towers, and dish-Stirling systems. Each model provides the user with unique capabilities that are catered to typical design considerations seen in each technology. Since the scope of the various models is generally limited to common plant configurations, new CSP technologies, component geometries, and subsystem combinations can be difficult to model directly in the existing SAM technology models. To overcome the limitations imposed by representative CSP technology models, NREL has developed a 'Generic Solar System' (GSS) performance model for use in SAM. This paper discusses the formulation and performance considerations included in this model and verifies the model by comparing its results with more detailed models.

Wagner, M. J.; Zhu, G.

2011-08-01T23:59:59.000Z

170

Processing pharus data with the generic SAR processor  

SciTech Connect (OSTI)

The Generic SAR Processor (GSP) is a SAR processing environment created to process airborne and spaceborne SAR data with a maximum amount of flexibility, while at the same time providing a user friendly and powerful environment for handling and analyzing SAR, including polarimetric calibration. PHARUS is an airborne polarimetric C-band SAR, utilizing an active (solid state) phased array. The absence of mechanical antenna stabilization, the use of electronic beam steering, in combination with high PRF, polarimetric operation, under motion condition which can be severe, requires a very large flexibility of the SAR processor. The GSP is designed to handle this type of SAR data with a very flexible motion compensation-, azimuth compression-, and radiometric correction approach. First experiences with the processing of PHARUS data show that this is a valid approach to obtain high quality polarimetric imagery with a phased array SAR. 4 refs., 5 figs.

Otten, M.P.G. [TNO Physics and Electronics Lab., The Hague (Netherlands)

1996-11-01T23:59:59.000Z

171

Asymptotically Flat Wormhole Solutions in a Generic Cosmological Constant Background  

E-Print Network [OSTI]

There are a number of reasons to study wormholes with generic cosmological constant $\\Lambda$. Recent observations indicate that present accelerating expansion of the universe demands $\\Lambda>0$. On the other hand, some extended theories of gravitation such as supergravity and superstring theories posses vacuum states with $\\Lambdaenergy density and pressure profiles which support such a geometry are obtained. It is shown that for having such a geometry, the wormhole throat $r_0$, the cosmological constant $\\Lambda$ and the equation of state parameter $\\omega$ should satisfy two specific conditions. The possibility of setting different values for the parameters of the model helps us to find exact solutions for the metric functions, mass functions and energy-momentum profiles. At last, the volume integral quantifier, which provides useful information about the total amount of energy condition violating matter is discussed briefly.

Y. Heydarzade; N. Riazi; H. Moradpour

2015-01-09T23:59:59.000Z

172

Reactor hot spot analysis  

SciTech Connect (OSTI)

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

P Reactor Grouting  

SciTech Connect (OSTI)

Filling the P Reactor with grout. This seals the radioactive material and reduces the environmental footprint left from the Cold War. Project sponsored by the Recovery Act at the Savannah River Site.

None

2010-01-01T23:59:59.000Z

174

Nuclear reactor control  

SciTech Connect (OSTI)

A liquid metal cooled fast breeder nuclear reactor has power setback means for use in an emergency. On initiation of a trip-signal a control rod is injected into the core in two stages, firstly, by free fall to effect an immediate power-set back to a safe level and, secondly, by controlled insertion. Total shut-down of the reactor under all emergencies is avoided. 4 claims.

Ingham, R.V.

1980-01-01T23:59:59.000Z

175

Polymerization reactor control  

SciTech Connect (OSTI)

The principal difficulties in achieving good control of polymerization reactors are related to inadequate on-line measurement, a lack of understanding of the dynamics of the process, the highly sensitive and nonlinear behavior of these reactors, and the lack of well-developed techniques for the control of nonlinear processes. Some illustrations of these problems and a discussion of potential techniques for overcoming some of these difficulties is provided.

Ray, W.H.

1985-01-01T23:59:59.000Z

176

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

177

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

178

F Reactor Inspection  

SciTech Connect (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

2014-10-29T23:59:59.000Z

179

Reactor Safety Research Programs  

SciTech Connect (OSTI)

This document summarizes the work performed by Pacific Northwest laboratory from October 1 through December 31, 1979, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, lspra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

Dotson, CW

1980-08-01T23:59:59.000Z

180

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

Note: This page contains sample records for the topic "reactor generic distributed" from the National Library of EnergyBeta (NLEBeta).
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181

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

SciTech Connect (OSTI)

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

182

Effects of Spatial Variations in Packing Fraction on Reactor Physics Parameters in Pebble-Bed Reactors  

SciTech Connect (OSTI)

The well-known spatial variation of packing fraction near the outer boundary of a pebble-bed reactor core is cited. The ramifications of this variation are explored with the MCNP computer code. It is found that the variation has negligible effects on the global reactor physics parameters extracted from the MCNP calculations for use in analysis by diffusion-theory codes, but for local reaction rates the effects of the variation are naturally important. Included is some preliminary work in using first-order perturbation theory for estimating the effect of the spatial variation of packing fraction on the core eigenvalue and the fision density distribution.

William K. Terry; A. M. Ougouag; Farzad Rahnema; Michael Scott McKinley

2003-04-01T23:59:59.000Z

183

Solar Thermal Reactor Materials Characterization  

SciTech Connect (OSTI)

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

184

Termination of rewriting strategies: a generic approach Isabelle Gnaedig, Hel`ene Kirchner  

E-Print Network [OSTI]

Termination of rewriting strategies: a generic approach Isabelle Gnaedig, H´el`ene Kirchner LORIA@loria.fr,Helene.Kirchner@loria.fr Abstract. We propose a generic termination proof method for rewriting under strategies, based on an explicit induction on the termination property. Rewriting trees on ground terms are modelized by proof

Paris-Sud XI, Université de

185

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

186

An Algorithm to Learn Read-Once Threshold Formulas, and some Generic Transformations between Learning Models  

E-Print Network [OSTI]

An Algorithm to Learn Read-Once Threshold Formulas, and some Generic Transformations between. Hancocky Siemens Corporate Research, Inc. 755 College Road East Princeton, NJ 08540 e-mail: hancock@learning.scr.siemens of generic transformations that can be used to convert an algorithm in one learning model into an algorithm

Eckmiller, Rolf

187

REACTOR GROUT THERMAL PROPERTIES  

SciTech Connect (OSTI)

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

188

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

189

An assessment of space reactor technology needs and recommendations for development  

SciTech Connect (OSTI)

In order to provide a strategy for space reactor technology development, the Defense Nuclear Agency (DNA) has authorized a brief review of potential national needs that may be addressed by space reactor systems. A systematic approach was used to explore needs at several levels that are increasingly specific. {sm_bullet} Level 0{emdash}General Trends and Issues {sm_bullet} Level 1{emdash}Generic Space Capabilities to Address Trends {sm_bullet} Level 2{emdash}Requirements to Support Capabilities {sm_bullet} Level 3{emdash}System Types Capable of Meeting Requirements {sm_bullet} Level 4{emdash}Generic Reactor System Types {sm_bullet} Level 5{emdash}Specific Baseline Systems Using these findings, a strategy was developed to support important space reactor technologies within a limited budget. A preliminary evaluation identified key technical issues and provide a prioritized set of candidate research projects. The evaluation of issues and the recommended research projects are presented in a companion paper. {copyright} {ital 1996 American Institute of Physics.}

Marshall, A.C. [Sandia National Laboratories, P.O. Box 5800, Albuquerque, New Mexico 87185 (United States); Wiley, R.L. [Consultant, 5998 Camelback Lane, Columbia, Maryland 21045 (United States)

1996-03-01T23:59:59.000Z

190

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

SciTech Connect (OSTI)

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

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

2008-07-15T23:59:59.000Z

191

Methanation assembly using multiple reactors  

DOE Patents [OSTI]

A methanation assembly for use with a water supply and a gas supply containing gas to be methanated in which a reactor assembly has a plurality of methanation reactors each for methanating gas input to the assembly and a gas delivery and cooling assembly adapted to deliver gas from the gas supply to each of said methanation reactors and to combine water from the water supply with the output of each methanation reactor being conveyed to a next methanation reactor and carry the mixture to such next methanation reactor.

Jahnke, Fred C.; Parab, Sanjay C.

2007-07-24T23:59:59.000Z

192

Power Burst Facility (PBF) Reactor Reactor Decommissioning  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 - September 2006PhotovoltaicSeptember 22,Reactor Decommissioning Click here to view

193

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

Office of Scientific and Technical Information (OSTI)

A reactor and associated power plant designed to produce 1.05 Mwh and 3.535 Mwh of steam for heating purposes are described. The total thermal output of the reactor is 10 Mwh....

194

Heat dissipating nuclear reactor  

DOE Patents [OSTI]

Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extends from the metal base plate downwardly and outwardly into the earth.

Hunsbedt, Anstein (Los Gatos, CA); Lazarus, Jonathan D. (Sunnyvale, CA)

1987-01-01T23:59:59.000Z

195

Inexpensive Mini Thermonuclear Reactor  

E-Print Network [OSTI]

This proposed design for a mini thermonuclear reactor uses a method based upon a series of important innovations. A cumulative explosion presses a capsule with nuclear fuel up to 100 thousands of atmospheres, the explosive electric generator heats the capsule/pellet up to 100 million degrees and a special capsule and a special cover which keeps these pressure and temperature in capsule up to 0.001 sec. which is sufficient for Lawson criteria for ignition of thermonuclear fuel. Major advantages of these reactors/bombs is its very low cost, dimension, weight and easy production, which does not require a complex industry. The mini thermonuclear bomb can be delivered as a shell by conventional gun (from 155 mm), small civil aircraft, boat or even by an individual. The same method may be used for thermonuclear engine for electric energy plants, ships, aircrafts, tracks and rockets. Key words: Thermonuclear mini bomb, thermonuclear reactor, nuclear energy, nuclear engine,

Alexander Bolonkin; Alexander Bolonkin

196

Nuclear reactor safety device  

DOE Patents [OSTI]

A safety device is disclosed for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of an upward thermal excursion. Such safety device comprises a laminated helical ribbon configured as a tube-like helical coil having contiguous helical turns with slidably abutting edges. The helical coil is disclosed as a portion of a drive member connected axially to the control rod. The laminated ribbon is formed of outer and inner laminae. The material of the outer lamina has a greater thermal coefficient of expansion than the material of the inner lamina. In the event of an upward thermal excursion, the laminated helical coil curls inwardly to a smaller diameter. Such inward curling causes the total length of the helical coil to increase by a substantial increment, so that the control rod is axially repositioned by a corresponding amount to reduce the power output of the reactor.

Hutter, Ernest (Wilmette, IL)

1986-01-01T23:59:59.000Z

197

Thermionic Reactor Design Studies  

SciTech Connect (OSTI)

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

198

Reactor for exothermic reactions  

DOE Patents [OSTI]

A liquid phase process 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.degree. to 300.degree. 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, Jr., Lawrence A. (Bellaire, TX); Hearn, Dennis (Houston, TX); Jones, Jr., Edward M. (Friendswood, TX)

1993-01-01T23:59:59.000Z

199

Heat dissipating nuclear reactor  

DOE Patents [OSTI]

Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

Hunsbedt, A.; Lazarus, J.D.

1985-11-21T23:59:59.000Z

200

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

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.


201

The B & W Owners` Group Generic License Renewal Program  

SciTech Connect (OSTI)

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

202

Generic Degraded Congiguration Probability Analysis for DOE Codisposal Waste Package  

SciTech Connect (OSTI)

In accordance with the technical work plan, ''Technical Work Plan For: Department of Energy Spent Nuclear Fuel Work Packages'' (CRWMS M&O 2000c), this Analysis/Model Report (AMR) is developed for the purpose of screening out degraded configurations for U.S. Department of Energy (DOE) spent nuclear fuel (SNF) types. It performs the degraded configuration parameter and probability evaluations of the overall methodology specified in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000, Section 3) to qualifying configurations. Degradation analyses are performed to assess realizable parameter ranges and physical regimes for configurations. Probability calculations are then performed for configurations characterized by k{sub eff} in excess of the Critical Limit (CL). The scope of this document is to develop a generic set of screening criteria or models to screen out degraded configurations having potential for exceeding a criticality limit. The developed screening criteria include arguments based on physical/chemical processes and probability calculations and apply to DOE SNF types when codisposed with the high-level waste (HLW) glass inside a waste package. The degradation takes place inside the waste package and is long after repository licensing has expired. The emphasis of this AMR is on degraded configuration screening and the probability analysis is one of the approaches used for screening. The intended use of the model is to apply the developed screening criteria to each DOE SNF type following the completion of the degraded mode criticality analysis internal to the waste package.

S.F.A. Deng; M. Saglam; L.J. Gratton

2001-05-23T23:59:59.000Z

203

Reactor dosimetry and RPV life management  

SciTech Connect (OSTI)

Reactor dosimetry (RD) is a tool that provides data for neutron fluence accumulated over the reactor pressure vessel (RPV) during the reactor operation. This information, however, is not sufficient for RPV lifetime assessment. The life management of RPV is a multidisciplinary task. To assess whether the RPV steel properties at the current stage (for actual accumulated neutron fluence) of reactor operation are still 'safe enough,' the dependence of material properties on the fluence must be known; this is a task for material science (MS). Moreover, the mechanical loading over the RPV during normal operation and accidence have to be known, as well, for evaluation, if the RPV material integrity in this loading condition and existing cracks is provided. The crack loading path in terms of stress intensity factor is carried out by structural analyses (SA). Pressure and temperature distribution over RPV used in these analyses are obtained from a thermal hydraulic (TH) calculation. The conjunction of RD and other disciplines in RPV integrity assessment is analyzed in accordance with the FFP (fitness for purpose) approach. It could help to improve the efficiency in multi-disciplinary tasks solutions. (authors)

Belousov, S.; Ilieva, K.; Mitev, M. [Inst. for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences, Tsarigradsko 72, 1784 Sofia (Bulgaria)

2011-07-01T23:59:59.000Z

204

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.

Jassby, Daniel L. (Princeton, NJ)

1988-01-01T23:59:59.000Z

205

Fast quench reactor method  

DOE Patents [OSTI]

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream.

Detering, Brent A. (Idaho Falls, ID); Donaldson, Alan D. (Idaho Falls, ID); Fincke, James R. (Idaho Falls, ID); Kong, Peter C. (Idaho Falls, ID); Berry, Ray A. (Idaho Falls, ID)

1999-01-01T23:59:59.000Z

206

Fast quench reactor method  

DOE Patents [OSTI]

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream. 8 figs.

Detering, B.A.; Donaldson, A.D.; Fincke, J.R.; Kong, P.C.; Berry, R.A.

1999-08-10T23:59:59.000Z

207

Diagnostics for hybrid reactors  

SciTech Connect (OSTI)

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

208

Perspectives on reactor safety  

SciTech Connect (OSTI)

The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)

1994-03-01T23:59:59.000Z

209

Innovative design of uranium startup fast reactors  

E-Print Network [OSTI]

Sodium Fast Reactors are one of the three candidates of GEN-IV fast reactors. Fast reactors play an important role in saving uranium resources and reducing nuclear wastes. Conventional fast reactors rely on transuranic ...

Fei, Tingzhou

2012-01-01T23:59:59.000Z

210

Reactor operation environmental information document  

SciTech Connect (OSTI)

The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

1989-12-01T23:59:59.000Z

211

Generic configuration management system for documentation and on-line beam optics applications in the CERN PS complex  

E-Print Network [OSTI]

Generic configuration management system for documentation and on-line beam optics applications in the CERN PS complex

Autin, Bruno; Jensen, O; Kazymov, A; Lindroos, M; Schinzel, Josi; Wildner, E

1996-01-01T23:59:59.000Z

212

Generic vehicle speed models based on traffic simulation: Development and application  

SciTech Connect (OSTI)

This paper summarizes the findings of a research project to develop new methods of estimating speeds for inclusion in the Highway Performance Monitoring System (HPMS) Analytical Process. The paper focuses on the effects of traffic conditions excluding incidents (recurring congestion) on daily average ed and excess fuel consumption. A review of the literature revealed that many techniques have been used to predict speeds as a function of congestion but most fail to address the effects of queuing. However, the method of Dowling and Skabardonis avoids this limitation and was adapted to the research. The methodology used the FRESIM and NETSIM microscopic traffic simulation models to develop uncongested speed functions and as a calibration base for the congested flow functions. The chief contributions of the new speed models are the simplicity of application and their explicit accounting for the effects of queuing. Specific enhancements include: (1) the inclusion of a queue discharge rate for freeways; (2) use of newly defined uncongested flow speed functions; (3) use of generic temporal distributions that account for peak spreading; and (4) a final model form that allows incorporation of other factors that influence speed, such as grades and curves. The main limitation of the new speed models is the fact that they are based on simulation results and not on field observations. They also do not account for the effect of incidents on speed. While appropriate for estimating average national conditions, the use of fixed temporal distributions may not be suitable for analyzing specific facilities, depending on observed traffic patterns. Finally, it is recommended that these and all future speed models be validated against field data where incidents can be adequately identified in the data.

Margiotta, R.; Cohen, H.; Elkins, G.; Rathi, A.; Venigalla, M.

1994-12-15T23:59:59.000Z

213

Reactor operation safety information document  

SciTech Connect (OSTI)

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

214

Reed Reactor Facility Annual Report  

SciTech Connect (OSTI)

This is the report of the operations, experiments, modifications, and other aspects of the Reed Reactor Facility for the year.

Frantz, Stephen G.

2000-09-01T23:59:59.000Z

215

Thermal Reactor Safety  

SciTech Connect (OSTI)

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

216

Nuclear reactor building  

DOE Patents [OSTI]

A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed thereabove. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define therebetween an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin.

Gou, Perng-Fei (Saratoga, CA); Townsend, Harold E. (Campbell, CA); Barbanti, Giancarlo (Sirtori, IT)

1994-01-01T23:59:59.000Z

217

Nuclear reactor building  

DOE Patents [OSTI]

A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed there above. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define there between an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin. 4 figures.

Gou, P.F.; Townsend, H.E.; Barbanti, G.

1994-04-05T23:59:59.000Z

218

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

219

Nuclear Reactors and Technology  

SciTech Connect (OSTI)

This publication 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 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 the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, 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.

1992-01-01T23:59:59.000Z

220

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

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

Aerosol engineering: design and stability of aerosol reactors  

SciTech Connect (OSTI)

A theoretical study of the performance of aerosol reactors is presented. The goals of this study are (1) to identify the appropriate reactor types (batch, CSTR, and tubular) for production of aerosol with specific properties (for example, uniform size particles, high aerosol surface area, etc.) and (2) to investigate the effect of various process parameters on product aerosol characteristics and on the stability of operation of aerosol reactors. In all the reactors considered, the aerosol dynamics were detemined by chemical reaction, nucleation, and aerosol growth in the free molecule regime in the absence of coagulation at isothermal conditions. Formulation of the aerosol dynamics in terms of moments of the aerosol size distribution facilitated the numerical solution of the resulting systems of ordinary or partial differential equations. The stability characteristics of a continuous stirred tank aerosol reactor (CSTAR) were investigated since experimental data in the literature indicate that under certain conditions this reactor exhibits oscillatory behavior with respect to product aerosol concentration and size distribution.

Pratsinis, S.E.

1985-01-01T23:59:59.000Z

222

Generic effluent monitoring system certification for salt well portable exhauster  

SciTech Connect (OSTI)

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

223

Generic effluent monitoring system certification for AP-40 exhauster stack  

SciTech Connect (OSTI)

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

224

Baseline Concept Description of a Small Modular High Temperature Reactor  

SciTech Connect (OSTI)

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

225

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

226

Neutronic analysis of a fusion hybrid reactor  

SciTech Connect (OSTI)

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

227

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

228

Cooperative Monitoring Center Occasional Paper/7: A Generic Model for Cooperative Border Security  

SciTech Connect (OSTI)

This paper presents a generic model for dealing with security problems along borders between countries. It presents descriptions and characteristics of various borders and identifies the threats to border security, while emphasizing cooperative monitoring solutions.

Netzer, Colonel Gideon

1999-03-01T23:59:59.000Z

229

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

230

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

SciTech Connect (OSTI)

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

231

Predicting Reactor Antineutrino Emissions Using New Precision Beta Spectroscopy  

SciTech Connect (OSTI)

Neutrino experiments at nuclear reactors are currently vital to the study of neutrino oscillations. The observed antineutrino rates at reactors are typically lower than model expectations. This observed deficit is called the “reactor neutrino anomaly”. A new understanding of neutrino physics may be required to explain this deficit, though model estimation uncertainties may also play a role in the apparent discrepancy. PNNL is currently investigating an experimental technique that promises reduced uncertainties for measured data to support these hypotheses and interpret reactor antineutrino measurements. The experimental approach is to 1) direct a proton accelerator beam on a metal target to produce a source of neutrons, 2) use spectral tailoring to modify the neutron spectrum to closely simulate the energy distribution of a power reactor neutron spectrum, 3) irradiate isotopic fission foils (235U, 238U, 239Pu, 241Pu) in this neutron spectrum so that fissions occur at energies representative of a reactor, 4) transport the beta particles released by the fission products in the foils to a beta spectrometer, 5) measure the beta energy spectrum, and 6) invert the measured beta energy spectrum to an antineutrino energy spectrum. A similar technique using a beta spectrometer and isotopic fission foils was pioneered in the 1980’s at the ILL thermal reactor. Those measurements have been the basis for interpreting all subsequent antineutrino measurements at reactors. A basic constraint in efforts to reduce uncertainties in predicting the antineutrino emission from reactor cores is any underlying limitation of the original measurements. This may include beta spectrum energy resolution, the absolute normalization of beta emission to number of fission, statistical counting uncertainties, lack of 238U data, the purely thermal nature of the IIL reactor neutrons used, etc. An accelerator-based neutron source that can be tailored to match various reactor neutron spectra provides an advantage for control in studying how changes in the neutron spectra (i.e. "in the reactor core") affects the resulting fission product beta spectrum. Furthermore, the 238U antineutrino spectrum, which has not been measured, can be studied directly because of the enhanced 1 MeV fast neutron flux available at the accelerator source. A facility such as the Project X Injector Experiment (PXIE) 30 MeV proton linear accelerator at Fermilab is being considered for this experiment. The hypothesis is that a new approach utilizing the flexibility of an accelerator neutron source with spectral tailoring coupled with a careful design of an isotopic fission target and beta spectrometer and the inversion of the beta spectrum to the neutrino spectrum will allow further reduction in the uncertainties associated with prediction of the reactor antineutrino spectrum.

Asner, David M.; Burns, Kimberly A.; Greenfield, Bryce A.; Kos, Marek S.; Orrell, John L.; Schram, Malachi; VanDevender, Brent A.; Wootan, David W.

2013-05-01T23:59:59.000Z

232

OXIDATIVE COUPLING OF METHANE USING INORGANIC MEMBRANE REACTORS  

SciTech Connect (OSTI)

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

233

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

234

Nuclear divisional reactor  

SciTech Connect (OSTI)

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

235

Nuclear reactor safety device  

DOE Patents [OSTI]

A safety device is described for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of a thermal excursion. It comprises a laminated strip helically configured to form a tube, said tube being in operative relation to said control rod. The laminated strip is formed of at least two materials having different thermal coefficients of expansion, and is helically configured such that the material forming the outer lamina of the tube has a greater thermal coefficient of expansion than the material forming the inner lamina of said tube. In the event of a thermal excursion the laminated strip will tend to curl inwardly so that said tube will increase in length, whereby as said tube increases in length it exerts a force on said control rod to axially reposition said control rod with respect to said core.

Hutter, E.

1983-08-15T23:59:59.000Z

236

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

237

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

238

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

SciTech Connect (OSTI)

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

239

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

SciTech Connect (OSTI)

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

240

Thermionic Reactor Design Studies  

SciTech Connect (OSTI)

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

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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241

Nuclear reactor engineering: Reactor systems engineering. Fourth edition, Volume Two  

SciTech Connect (OSTI)

This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in the design and operation of nuclear power plants. Extensively updated, the fourth edition includes new materials on reactor safety and risk analysis, regulation, fuel management, waste management and operational aspects of nuclear power. This volume contains the following: the systems concept, design decisions, and information tools; energy transport; reactor fuel management and energy cost considerations; environmental effects of nuclear power and waste management; nuclear reactor safety and regulation; power reactor systems; plant operations; and advanced plants and the future.

Glasstone, S.; Sesonske, A.

1994-12-31T23:59:59.000Z

242

Designing Reactors to Facilitate Decommissioning  

SciTech Connect (OSTI)

Critics of nuclear power often cite issues with tail-end-of-the-fuel-cycle activities as reasons to oppose the building of new reactors. In fact, waste disposal and the decommissioning of large nuclear reactors have proven more challenging than anticipated. In the early days of the nuclear power industry the design and operation of various reactor systems was given a great deal of attention. Little effort, however, was expended on end-of-the-cycle activities, such as decommissioning and disposal of wastes. As early power and test reactors have been decommissioned difficulties with end-of-the-fuel-cycle activities have become evident. Even the small test reactors common at the INEEL were not designed to facilitate their eventual decontamination, decommissioning, and dismantlement. The results are that decommissioning of these facilities is expensive, time consuming, relatively hazardous, and generates large volumes of waste. This situation clearly supports critics concerns about building a new generation of power reactors.

Richard H. Meservey

2006-06-01T23:59:59.000Z

243

Solid oxide electrochemical reactor science.  

SciTech Connect (OSTI)

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

244

Progress Update: Reactor Disassembly Grouting  

SciTech Connect (OSTI)

Grouting the P&R reactors in order to remove these basins as an environmental threat. This will end the Cold War legacy and end the environmental footprint.

Cody, Tom

2010-01-01T23:59:59.000Z

245

Neutrino Oscillation Studies with Reactors  

E-Print Network [OSTI]

Nuclear reactors are one of the most intense, pure, controllable, cost-effective, and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavors are quantum mechanical mixtures. Over the past several decades reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle $\\theta_{13}$. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

Petr Vogel; Liangjian Wen; Chao Zhang

2015-03-03T23:59:59.000Z

246

Neutrino Oscillation Studies with Reactors  

E-Print Network [OSTI]

Nuclear reactors are one of the most intense, pure, controllable, cost-effective, and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavors are quantum mechanical mixtures. Over the past several decades reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle $\\theta_{13}$. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

Vogel, Petr; Zhang, Chao

2015-01-01T23:59:59.000Z

247

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

248

Light Water Reactor Sustainability Newsletter  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (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)...

249

Light Water Reactor Sustainability Newsletter  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

30-35, August 2012. Clayton, D. A. and M. S. Hileman, 2012, Light Water Reactor Sustainability Non-Destructive Evaluation for Concrete Research and Development Roadmap, ORNLTM-...

250

Progress Update: Reactor Disassembly Grouting  

ScienceCinema (OSTI)

Grouting the P&R reactors in order to remove these basins as an environmental threat. This will end the Cold War legacy and end the environmental footprint.

Cody, Tom

2012-06-14T23:59:59.000Z

251

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

252

Reactor refueling containment system  

DOE Patents [OSTI]

A method of refueling a nuclear reactor is disclosed whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced. 2 figs.

Gillett, J.E.; Meuschke, R.E.

1995-05-02T23:59:59.000Z

253

Reactor refueling containment system  

DOE Patents [OSTI]

A method of refueling a nuclear reactor whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced.

Gillett, James E. (Greensburg, PA); Meuschke, Robert E. (Pittsburgh, PA)

1995-01-01T23:59:59.000Z

254

Nuclear reactor control assembly  

SciTech Connect (OSTI)

This patent describes an assembly for providing global power control in a nuclear reactor having the core split into two halves. It comprises a disk assembly formed from at least two disks each machined with an identical surface hole pattern such that rotation of one disk relative to the other causes the hole pattern to open or close, the disk assembly being positioned substantially at the longitudinal center of and coaxial with the core halves; and means for rotating at least one of the disks relative to the other.

Negron, S.B.

1991-06-11T23:59:59.000Z

255

Nuclear reactor control apparatus  

SciTech Connect (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 additonal magnetically-operated latch for releasing the control rod under two different conditions, one involving seismic shock.

Sridhar, B.N.

1981-08-28T23:59:59.000Z

256

Biparticle fluidized bed reactor  

DOE Patents [OSTI]

A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase is described. 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 figures.

Scott, C.D.

1993-12-14T23:59:59.000Z

257

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

258

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

259

Licensed reactor nuclear safety criteria applicable to DOE reactors  

SciTech Connect (OSTI)

The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC (Nuclear Regulatory Commission) licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor.

Not Available

1991-04-01T23:59:59.000Z

260

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

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

Fast Reactor Fuel Type and Reactor Safety Performance  

SciTech Connect (OSTI)

Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of ‘inherent safety’ concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and thermophysical properties of the fuel and their compatibility with the reactor coolant, with corresponding differences in the challenges presented to the reactor developers. Accident phenomena are discussed for the sodium-cooled fast reactor based on the mechanistic progression of conditions from accident initiation to accident termination, whether a benign state is achieved or more severe consequences are expected. General principles connecting accident phenomena and fuel properties are developed from the oxide and metal fuel safety analyses, providing guidelines that can be used as part of the evaluation for selection of fuel type for the sodium-cooled fast reactor.

R. Wigeland; J. Cahalan

2009-09-01T23:59:59.000Z

262

Gaseous reactor control system  

SciTech Connect (OSTI)

This paper describes a nuclear reactor control system for controlling the reactivity of the core of a nuclear reactor. It includes a control gas having a high neutron cross-section; a first tank containing a first supply of the control gas; a first conduit providing a first fluid passage extending into the core, the first conduit being operatively connected to communicate with the first tank; a first valve operatively connected to regulate the flow of the control gas between the first tank and the first conduit; a second conduit concentrically disposed around the first conduit such that a second fluid passage is defined between the outer surface of the first conduit and the inner surface of the second conduit; a second tank containing a second supply of the control gas, the second tank being operatively connected to communicate with the second fluid passage; a second supply valve operatively connected to regulate the flow of the control gas between the second tank and the second fluid passage.

Abdel-Khalik, S.

1991-09-03T23:59:59.000Z

263

Overview of the US stellarator reactor study  

SciTech Connect (OSTI)

This study, which uses a cost-minimization code that incorporates the ARIES costing and reactor component models with a I-D energy transport calculation, shows that a torsatron reactor could be competitive with a tokamak reactor.

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)

1993-12-31T23:59:59.000Z

264

Adaptive Distributed Parameter and Input Estimation in Plasma Tokamak Heat  

E-Print Network [OSTI]

. Keywords: Thermonuclear fusion, distributed parameter systems, input state and parameter estimation, adaptive infinite-dimensional estimation, Galerkin method 1. INTRODUCTION In a controlled thermonuclear fusion reactor, the plasma thermal diffusivity and heating energy play an important role

Boyer, Edmond

265

Reactor Cost Analysis Brian James  

E-Print Network [OSTI]

Reactor Cost Analysis Brian James Directed Technologies, Inc. 6-7 November 2007 This presentation specification & optimization · Capital cost estimation · Projected hydrogen $/kg #12;Directed Technologies, Inc/WGS Membrane Reactor OTM/ Water-Splitting ANL With WGS #12;Directed Technologies, Inc. 6-7 November 2007 BILIWG

266

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.

Thorogood, Robert M. (Macungie, PA)

1983-01-01T23:59:59.000Z

267

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

268

Iterative methods for solving nonlinear problems of nuclear reactor criticality  

SciTech Connect (OSTI)

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

269

Fast reactors and nuclear nonproliferation  

SciTech Connect (OSTI)

Problems are discussed with regard to nuclear fuel cycle resistance in fast reactors to nuclear proliferation risk due to the potential for use in military programs of the knowledge, technologies and materials gained from peaceful nuclear power applications. Advantages are addressed for fast reactors in the creation of a more reliable mode of nonproliferation in the closed nuclear fuel cycle in comparison with the existing fully open and partially closed fuel cycles of thermal reactors. Advantages and shortcomings are also discussed from the point of view of nonproliferation from the start with fast reactors using plutonium of thermal reactor spent fuel and enriched uranium fuel to the gradual transition using their own plutonium as fuel. (authors)

Avrorin, E.N. [Russian Federal Nuclear Center - Zababakhin Institute of Applied Physics, Snezhinsk (Russian Federation); Rachkov, V.I.; Chebeskov, A.N. [State Scientific Center of the Russian Federation - Institute for Physics and Power Engineering, Bondarenko Square, 1, Obninsk, Kaluga region, 249033 (Russian Federation)

2013-07-01T23:59:59.000Z

270

Knowledge base expert system control of spatial xenon oscillations in pressurized water reactors  

SciTech Connect (OSTI)

Nuclear reactor operators are required to pay special attention to spatial xenon oscillations during the load-follow operation of pressurized water reactors. They are expected to observe the axial offset of the core, and to estimate the correct time and amount of necessary control action based on heuristic rules given in axial xenon oscillations are knowledge intensive, and heuristic in nature. An expert system, ACES (Axial offset Control using Expert Systems) is developed to implement a heuristic constant axial offset control procedure to aid reactor operators in increasing the plant reliability by reducing the human error component of the failure probability. ACES is written in a production system language, OPS5, based on the forward chaining algorithm. It samples reactor data with a certain time interval in terms of measurable parameters, such as the power, period, and the axial offset of the core. It then processes the core status utilizing a set of equations which are used in a back of the envelope calculations by domain experts. Heuristic rules of ACES identify the control variable to be used among the full and part length control rods and boron concentration, while a knowledge base is used to determine the amount of control. ACES is designed as a set of generic rules to avoid reducing the system into a set of patterns. Instead ACES evaluates the system, determines the necessary corrective actions in terms of reactivity insertion, and provides this reactivity insertion using the control variables. The amount of control action is determined using a knowledge base which consists of the differential rod worth curves, and the boron reactivity worth of a given reactor. Having the reactor dependent parameters in its knowledge base, ACES is applicable to an arbitrary reactor for axial offset control purposes.

Alten, S.

1992-01-01T23:59:59.000Z

271

MOOSE simulating nuclear reactor CRUD buildup  

SciTech Connect (OSTI)

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

272

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

273

MOOSE simulating nuclear reactor CRUD buildup  

ScienceCinema (OSTI)

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

2014-07-21T23:59:59.000Z

274

LIGHT WATER REACTOR SUSTAINABILITY PROGRAM: INTRODUCTION  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

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

275

Granular Dynamics in Pebble Bed Reactor Cores  

E-Print Network [OSTI]

gas reactors with the effective heat transfer of a molten salt coolant and the passive natural circulation safety systems of sodium fast reactors.

Laufer, Michael Robert

2013-01-01T23:59:59.000Z

276

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

277

HTGR Generic Technology Program. Semiannual report for the period ending September 30, 1980  

SciTech Connect (OSTI)

This document reports the technical accomplishments on the HTGR Generic Technology Program at General Atomic during the second half of FY-80. It covers a period when the design direction of the National HTGR Program is in the process of an overall review. The HTGR Generic Technology Program activities have continued so as to provide the basic technology required for all HTGR applications. The activities include the need to develop an LEU fuel and the need to qualify materials and components for the higher temperatures of the gas turbines and process heat plants.

Not Available

1980-11-01T23:59:59.000Z

278

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

SciTech Connect (OSTI)

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

279

COMSOL-based Nuclear Reactor Kinetics Studies at the HFIR  

SciTech Connect (OSTI)

The computational ability to accurately predict the dynamic behavior of a nuclear reactor core in response to reactivity-induced perturbations is an important subject in reactor physics. Space-time and point kinetics methodologies were developed for the purpose of studying the transient-induced behavior of the High Flux Isotope Reactor s (HFIR) compact core. The space-time simulations employed the three-energy-group neutron diffusion equations, and transients initiated by control cylinder and hydraulic tube rabbit ejections were studied. The work presented here is the first step towards creating a comprehensive multiphysics methodology for studying the dynamic behavior of the HFIR core during reactivity perturbations. The results of these studies show that point kinetics is adequate for small perturbations in which the power distribution is assumed to be time-independent, but space-time methods must be utilized to determine localized effects.

Chandler, David [ORNL] [ORNL; Freels, James D [ORNL] [ORNL; Maldonado, G Ivan [ORNL] [ORNL; Primm, Trent [ORNL] [ORNL

2011-01-01T23:59:59.000Z

280

Nuclear reactor control apparatus  

DOE Patents [OSTI]

Nuclear reactor safety rod release apparatus comprises a ring which carries detents normally positioned in an annular recess in outer side of the rod, the ring being held against the lower end of a drive shaft by magnetic force exerted by a solenoid carried by the drive shaft. When the solenoid is de-energized, the detent-carrying ring drops until the detents contact a cam surface associated with the lower end of the drive shaft, at which point the detents are cammed out of the recess in the safety rod to release the rod from the drive shaft. In preferred embodiments of the invention, an additional latch is provided to release a lower portion of a safety rod under conditions that may interfere with movement of the entire rod.

Sridhar, Bettadapur N. (Cupertino, CA)

1983-10-25T23:59:59.000Z

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

Nuclear reactor control  

DOE Patents [OSTI]

1. In a nuclear reactor incorporating a plurality of columns of tubular fuel elements disposed in horizontal tubes in a mass of graphite wherein water flows through the tubes to cool the fuel elements, the improvement comprising at least one control column disposed in a horizontal tube including fewer fuel elements than in a normal column of fuel elements and tubular control elements disposed at both ends of said control column, and means for varying the horizontal displacement of the control column comprising a winch at the upstream end of the control column and a cable extending through the fuel and control elements and attached to the element at the downstream end of the column.

Cawley, William E. (Phoenix, AZ); Warnick, Robert F. (Pasco, WA)

1982-01-01T23:59:59.000Z

282

Nuclear reactor control  

SciTech Connect (OSTI)

In a nuclear reactor incorporating a plurality of columns of tubular fuel elements disposed in horizontal tubes in a mass of graphite wherein water flows through the tubes to cool the fuel elements, the improvement comprising at least one control column disposed in a horizontal tube including fewer fuel elements than in a normal column of fuel elements and tubular control elements disposed at both ends of said control column, and means for varying the horizontal displacement of the control column comprising a winch at the upstream end of the control column and a cable extending through the fuel and control elements and attached to the element at the downstream end of the column.

Cawley, W.E.; Warnick, R.F.

1982-03-30T23:59:59.000Z

283

Nuclear reactor engineering: Reactor design basics. Fourth edition, Volume One  

SciTech Connect (OSTI)

This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in design and operation of nuclear power plants. Extensively updated, the fourth edition includes new material on reactor safety and risk analysis, regulation, fuel management, waste management, and operational aspects of nuclear power. This volume contains the following: energy from nuclear fission; nuclear reactions and radiations; neutron transport; nuclear design basics; nuclear reactor kinetics and control; radiation protection and shielding; and reactor materials.

Glasstone, S.; Sesonske, A.

1994-12-31T23:59:59.000Z

284

University Reactor Conversion Lessons Learned Workshop for Purdue University Reactor  

SciTech Connect (OSTI)

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

285

Licensed reactor nuclear safety criteria applicable to DOE reactors  

SciTech Connect (OSTI)

This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards.

Not Available

1993-11-01T23:59:59.000Z

286

Formulation and experimental evaluation of closed-form control laws for the rapid maneuvering of reactor neutronic power  

SciTech Connect (OSTI)

This report describes both the theoretical development and the experimental evaluation of a novel, robust methodology for the time-optimal adjustment of a reactor's neutronic power under conditions of closed-loop digital control. Central to the approach are the MIT-SNL Period-Generated Minimum Time Control Laws' which determine the rate at which reactivity should be changed in order to cause a reactor's neutronic power to conform to a specified trajectory. Using these laws, reactor power can be safely raised by five to seven orders of magnitude in a few seconds. The MIT-SNL laws were developed to facilitate rapid increases of neutronic power on spacecraft reactors operating in an SDI environment. However, these laws are generic and have other applications including the rapid recovery of research and test reactors subsequent to an unanticipated shutdown, power increases following the achievement of criticality on commercial reactors, power adjustments on commercial reactors so as to minimize thermal stress, and automated startups. The work reported here was performed by the Massachusetts Institute of Technology under contract to the Sandia National Laboratories. Support was also provided by the US Department of Energy's Division of University and Industry Programs. The work described in this report is significant in that a novel solution to the problem of time-optimal control of neutronic power was identified, in that a rigorous description of a reactor's dynamics was derived in that the rate of change of reactivity was recognized as the proper control signal, and in that extensive experimental trials were conducted of these newly developed concepts on actual nuclear reactors. 43 refs., 118 figs., 11 tabs.

Bernard, J.A. (Massachusetts Inst. of Tech., Cambridge, MA (USA). Nuclear Reactor Lab.)

1989-09-01T23:59:59.000Z

287

Benchmark Evaluation of the Medium-Power Reactor Experiment Program Critical Configurations  

SciTech Connect (OSTI)

A series of small, compact critical assembly (SCCA) experiments were performed in 1962-1965 at the Oak Ridge National Laboratory Critical Experiments Facility (ORCEF) for the Medium-Power Reactor Experiment (MPRE) program. The MPRE was a stainless-steel clad, highly enriched uranium (HEU)-O2 fuelled, BeO reflected reactor design to provide electrical power to space vehicles. Cooling and heat transfer were to be achieved by boiling potassium in the reactor core and passing vapor directly through a turbine. Graphite- and beryllium-reflected assemblies were constructed at ORCEF to verify the critical mass, power distribution, and other reactor physics measurements needed to validate reactor calculations and reactor physics methods. The experimental series was broken into three parts, with the third portion of the experiments representing the beryllium-reflected measurements. The latter experiments are of interest for validating current reactor design efforts for a fission surface power reactor. The entire series has been evaluated as acceptable benchmark experiments and submitted for publication in the International Handbook of Evaluated Criticality Safety Benchmark Experiments and in the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

Margaret A. Marshall; John D. Bess

2013-02-01T23:59:59.000Z

288

Reactor physics design of supercritical CO?-cooled fast reactors  

E-Print Network [OSTI]

Gas-Cooled Fast Reactors (GFRs) are among the GEN-IV designs proposed for future deployment. Driven by anticipated plant cost reduction, the use of supercritical CO? (S-CO?) as a Brayton cycle working fluid in a direct ...

Pope, Michael A. (Michael Alexander)

2004-01-01T23:59:59.000Z

289

Reactor protection system design alternatives for sodium fast reactors  

E-Print Network [OSTI]

Historically, unprotected transients have been viewed as design basis events that can significantly challenge sodium-cooled fast reactors. The perceived potential consequences of a severe unprotected transient in a ...

DeWitte, Jacob D. (Jacob Dominic)

2011-01-01T23:59:59.000Z

290

United States Domestic Research Reactor Infrastrucutre TRIGA Reactor Fuel Support  

SciTech Connect (OSTI)

The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

Douglas Morrell

2011-03-01T23:59:59.000Z

291

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

292

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.

293

A generic knowledge Model for SME Supply Chain Based on Multiagent Paradigm  

E-Print Network [OSTI]

A generic knowledge Model for SME Supply Chain Based on Multiagent Paradigm Jihene Tounsi1 Enterprises (SME). These companies evolve in an unstable and complex network. In order to guarantee its role in a supply chain, SME must be able to support the inherent requirements of the supply chain (low lead times

Paris-Sud XI, Université de

294

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

295

An Algorithm to Learn ReadOnce Threshold Formulas, and some Generic Transformations between Learning Models  

E-Print Network [OSTI]

An Algorithm to Learn Read­Once Threshold Formulas, and some Generic Transformations between@cpsc.ucalgary.ca Thomas R. Hancock y Siemens Corporate Research, Inc. 755 College Road East Princeton, NJ 08540 e­mail: hancock@learning.scr.siemens.com Lisa Hellerstein z Department of EECS Northwestern University 2145

Eckmiller, Rolf

296

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

297

Generic Design Methodology for Symmetrically Coupled Line Structures Sarmad Al-Taei, Phil Lane, George Passiopoulos  

E-Print Network [OSTI]

Generic Design Methodology for Symmetrically Coupled Line Structures Sarmad Al-Taei, Phil Lane, Surrey, GU15 3DT Abstract: High performance miniaturised and cost effective passive components are needed systems. In recent years there has been increasing interest in utilising passive multi-layer IC

Haddadi, Hamed

298

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

299

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

300

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

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

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.

302

Wigner spacing distribution in classical mechanics  

E-Print Network [OSTI]

The Wigner spacing distribution has a long and illustrious history in nuclear physics and in the quantum mechanics of classically chaotic systems. In this paper, a long-overlooked connection between the Wigner distribution and classical chaos in two-degree-of-freedom (2D) conservative systems is introduced. In the specific context of fully chaotic 2D systems, the hypothesis that typical pseudotrajectories of a canonical Poincar\\'{e} map have a Wignerian nearest-neighbor spacing distribution (NNSD), is put forward and tested. Employing the 2D circular stadium billiard as a generic test case, the NNSD of a typical pseudotrajectory of the Birkhoff map is shown to be in excellent agreement with the Wigner distribution. The relevance of the higher-order Wigner surmises from random matrix theory are also illustrated.

Jamal Sakhr

2014-07-08T23:59:59.000Z

303

2012 Annual Report Research Reactor Infrastructure Program  

SciTech Connect (OSTI)

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

Douglas Morrell

2012-11-01T23:59:59.000Z

304

Unique features of space reactors  

SciTech Connect (OSTI)

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

305

Nuclear Reactors and Technology; (USA)  

SciTech Connect (OSTI)

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

306

Spectral shift reactor control method  

SciTech Connect (OSTI)

The method is described of closely controlling the reactor water coolant temperature of an operating spectral-shift nuclear reactor, the reactor comprising a core formed of fuel assemblies through which the reactor water coolant flows; different types of elongated elements operable to be controllably moved into and out of the core; one type of the elongated elements comprising control rods formed of neutron absorbing material and operable to decrease reactivity through neutron absorption when inserted into the core; another of the types of elongated elements comprising displacer rods formed of material which has a low absorption for neutrons and which have overall neutron-absorbing and moderating characteristics essentially not exceeding those of hollow tubular Zircaloy members with a filling zirconium oxide or aluminum oxide, the displacer rods operating to displace an equivalent volume of water coolant fluid from the core when inserted therein to decrease reactivity and to increase reactivity when moved from the core.

Impink, A.J. Jr.

1987-08-18T23:59:59.000Z

307

Reactor core isolation cooling system  

DOE Patents [OSTI]

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

Cooke, F.E.

1992-12-08T23:59:59.000Z

308

University Reactor Matching Grants Program  

SciTech Connect (OSTI)

During the 2002 Fiscal year, funds from the DOE matching grant program, along with matching funds from the industrial sponsors, have been used to support research in the area of thermal-hydraulics. Both experimental and numerical research projects have been performed. Experimental research focused on two areas: (1) Identification of the root cause mechanism for axial offset anomaly in pressurized water reactors under prototypical reactor conditions, and (2) Fluid dynamic aspects of thin liquid film protection schemes for inertial fusion reactor chambers. Numerical research focused on two areas: (1) Multi-fluid modeling of both two-phase and two-component flows for steam conditioning and mist cooling applications, and (2) Modeling of bounded Rayleigh-Taylor instability with interfacial mass transfer and fluid injection through a porous wall simulating the ''wetted wall'' protection scheme in inertial fusion reactor chambers. Details of activities in these areas are given.

John Valentine; Farzad Rahnema; Said Abdel-Khalik

2003-02-14T23:59:59.000Z

309

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

310

Teaching About Nature's Nuclear Reactors  

E-Print Network [OSTI]

Naturally occurring nuclear reactors existed in uranium deposits on Earth long before Enrico Fermi built the first man-made nuclear reactor beneath Staggs Field in 1942. In the story of their discovery, there are important lessons to be learned about scientific inquiry and scientific discovery. Now, there is evidence to suggest that the Earth's magnetic field and Jupiter's atmospheric turbulence are driven by planetary-scale nuclear reactors. The subject of planetocentric nuclear fission reactors can be a jumping off point for stimulating classroom discussions about the nature and implications of planetary energy sources and about the geomagnetic field. But more importantly, the subject can help to bring into focus the importance of discussing, debating, and challenging current thinking in a variety of areas.

Herndon, J M

2005-01-01T23:59:59.000Z

311

Reactor core isolation cooling system  

DOE Patents [OSTI]

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

Cooke, Franklin E. (San Jose, CA)

1992-01-01T23:59:59.000Z

312

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

313

Automatic safety rod for reactors  

DOE Patents [OSTI]

An automatic safety rod for a nuclear reactor containing neutron absorbing material and designed to be inserted into a reactor core after a loss-of-core flow. Actuation is based upon either a sudden decrease in core pressure drop or the pressure drop decreases below a predetermined minimum value. The automatic control rod includes a pressure regulating device whereby a controlled decrease in operating pressure due to reduced coolant flow does not cause the rod to drop into the core.

Germer, John H. (San Jose, CA)

1988-01-01T23:59:59.000Z

314

Computer aided nuclear reactor modeling  

E-Print Network [OSTI]

CHAPTER Page IV ALPHA ARCHITECTURE Design Philosophy Abstract Data Type Based Modules Grouping by Functions Miscellaneous Design Influences Architecture . . X Window System . Editor Library Model Library User Interface Library . V CONCLUSIONS... Connected Model . . . . , . . . 31 12 13 Header Section Editor Editing a "Choice" Attribute A Table of Vectors . 32 33 . 34 14 15 16 Current Reactor Modeling Schematic Reactor Modeling Schematic with Alpha Public Header File of Vertex Module...

Warraich, Khalid Sarwar

1995-01-01T23:59:59.000Z

315

Fast quench reactor and method  

DOE Patents [OSTI]

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

Detering, Brent A. (Idaho Falls, ID); Donaldson, Alan D. (Idaho Falls, ID); Fincke, James R. (Idaho Falls, ID); Kong, Peter C. (Idaho Falls, ID)

1998-01-01T23:59:59.000Z

316

Novel Catalytic Membrane Reactors  

SciTech Connect (OSTI)

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

317

Nuclear reactor control rod  

SciTech Connect (OSTI)

This patent describes a vertically oriented bottom entry control rod from a nuclear reactor: a frame including an elongated central spine of cruciform cross section connected between an upper support member and a lower support member both of cruciform shape having four laterally extending arms. The arms are in alignment with the arms of the lower support member and each aligned upper and lower support members has a sheath extending between; absorber plates of neutron absorber material, different from the material of the frame, one of the absorber plates is positioned within a sheath beneath each of the arms; attachment means suspends the absorber plates from the arms of the upper support member within a sheath; elongated absorber members positioned within a sheath between each of the suspended absorber plates and an arm of the lower support member; and joint means between the upper ends of the absorber members and the lower ends of the suspended absorber plates for minimizing gaps; the sheath means encloses the suspended absorber plates and the absorber members extending between aligned arms of the upper and lower support members and secured.

Cearley, J.E.; Izzo, K.R.

1987-06-30T23:59:59.000Z

318

Reactor pressure vessel nozzle  

DOE Patents [OSTI]

A nozzle for joining a pool of water to a nuclear reactor pressure vessel includes a tubular body having a proximal end joinable to the pressure vessel and a distal end joinable in flow communication with the pool. The body includes a flow passage therethrough having in serial flow communication a first port at the distal end, a throat spaced axially from the first port, a conical channel extending axially from the throat, and a second port at the proximal end which is joinable in flow communication with the pressure vessel. The inner diameter of the flow passage decreases from the first port to the throat and then increases along the conical channel to the second port. In this way, the conical channel acts as a diverging channel or diffuser in the forward flow direction from the first port to the second port for recovering pressure due to the flow restriction provided by the throat. In the backflow direction from the second port to the first port, the conical channel is a converging channel and with the abrupt increase in flow area from the throat to the first port collectively increase resistance to flow therethrough. 2 figs.

Challberg, R.C.; Upton, H.A.

1994-10-04T23:59:59.000Z

319

Propellant actuated nuclear reactor steam depressurization valve  

DOE Patents [OSTI]

A nuclear fission reactor combined with a propellant actuated depressurization and/or water injection valve is disclosed. The depressurization valve releases pressure from a water cooled, steam producing nuclear reactor when required to insure the safety of the reactor. Depressurization of the reactor pressure vessel enables gravity feeding of supplementary coolant water through the water injection valve to the reactor pressure vessel to prevent damage to the fuel core.

Ehrke, Alan C. (San Jose, CA); Knepp, John B. (San Jose, CA); Skoda, George I. (Santa Clara, CA)

1992-01-01T23:59:59.000Z

320

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

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

Passive containment cooling water distribution device  

DOE Patents [OSTI]

A passive containment cooling system for a nuclear reactor containment vessel. Disclosed is a cooling water distribution system for introducing cooling water by gravity uniformly over the outer surface of a steel containment vessel using a series of radial guide elements and cascading weir boxes to collect and then distribute the cooling water into a series of distribution areas through a plurality of cascading weirs. The cooling water is then uniformly distributed over the curved surface by a plurality of weir notches in the face plate of the weir box.

Conway, Lawrence E. (Hookstown, PA); Fanto, Susan V. (Plum Borough, PA)

1994-01-01T23:59:59.000Z

322

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

323

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

324

Comments on Americium Volatilization during Fuel Fabrication for Fast Reactors  

SciTech Connect (OSTI)

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

325

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

SciTech Connect (OSTI)

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

326

Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency  

SciTech Connect (OSTI)

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

327

Development of refined MCNPX-PARET multi-channel model for transient analysis in research reactors  

SciTech Connect (OSTI)

Reactivity insertion transients are often analyzed (RELAP, PARET) using a two-channel model, representing the hot assembly with specified power distribution and an average assembly representing the remainder of the core. For the analysis of protected by the reactor safety system transients and zero reactivity feedback coefficients this approximation proves to give adequate results. However, a more refined multi-channel model representing the various assemblies, coupled through the reactivity feedback effects to the whole reactor core is needed for the analysis of unprotected transients with excluded over power and period trips. In the present paper a detailed multi-channel PARET model has been developed which describes the reactor core in different clusters representing typical BR2 fuel assemblies. The distribution of power and reactivity feedback in each cluster of the reactor core is obtained from a best-estimate MCNPX calculation using the whole core geometry model of the BR2 reactor. The sensitivity of the reactor response to power, temperature and energy distributions is studied for protected and unprotected reactivity insertion transients, with zero and non-zero reactivity feedback coefficients. The detailed multi-channel model is compared vs. simplified fewer-channel models. The sensitivities of transient characteristics derived from the different models are tested on a few reactivity insertion transients with reactivity feedback from coolant temperature and density change. (authors)

Kalcheva, S.; Koonen, E. [SCK-CEN, BR2 Reactor Dept., Boeretang 200, 2400 Mol (Belgium); Olson, A. P. [RERTR Program, Nuclear Engineering Div., Argonne National Laboratory, Cass Avenue, Argonne, IL 60439 (United States)

2012-07-01T23:59:59.000Z

328

Supercell Depletion Studies for Prismatic High Temperature Reactors  

SciTech Connect (OSTI)

The traditional two-step method of analysis is not accurate enough to represent the neutronic effects present in the prismatic high temperature reactor concept. The long range coupling of the various regions in high temperature reactors poses a set of challenges that are not seen in either LWRs or fast reactors. Unlike LWRs, which exhibit large, localized effects, the dominant effects in PMRs are, for the most part, distributed over larger regions, but with lower magnitude. The 1-D in-line treatment currently used in pebble bed reactor analysis is not sufficient because of the 2-D nature of the prismatic blocks. Considerable challenges exist in the modeling of blocks in the vicinity of reflectors, which, for current small modular reactor designs with thin annular cores, include the majority of the blocks. Additional challenges involve the treatment of burnable poisons, operational and shutdown control rods. The use of a large domain for cross section preparation provides a better representation of the neutron spectrum, enables the proper modeling of BPs and CRs, allows the calculation of generalized equivalence theory parameters, and generates a relative power distribution that can be used in compact power reconstruction. The purpose of this paper is to quantify the effects of the reflector, burnable poison, and operational control rods on an LEU design and to delineate an analysis approach for the Idaho National Laboratory. This work concludes that the use of supercells should capture these long-range effects in the preparation of cross sections and along with a set of triangular meshes to treat BPs, and CRs a high fidelity neutronics computation is attainable.

J. Ortensi

2012-10-01T23:59:59.000Z

329

Uncertainties in the Anti-neutrino Production at Nuclear Reactors  

E-Print Network [OSTI]

neutrino Production at Nuclear Reactors Z. Djurcic 1 , ?emission rates from nuclear reactors are determined fromlarge commercial nuclear reactors are playing an important

Djurcic, Zelimir

2009-01-01T23:59:59.000Z

330

Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors  

E-Print Network [OSTI]

Fundamental aspects of nuclear reactor fuel elements.Unlike permanent nuclear reactor core components, nuclearof the first nuclear reactors, commercial nuclear fuel still

Terrani, Kurt Amir

2010-01-01T23:59:59.000Z

331

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

Broader source: Energy.gov (indexed) [DOE]

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...

332

Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation  

E-Print Network [OSTI]

selected as part of the Generation IV reactors .. - 4 -The development of Generation IV fast reactors can make aconcepts selected for the Generation IV reactors, three,

Heidet, Florent

2010-01-01T23:59:59.000Z

333

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

334

UCLA program in reactor studies: The ARIES tokamak reactor study  

SciTech Connect (OSTI)

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

335

Refined Topological Vertex and Duality of Gauge Theories in Generic Omega Backgrounds  

E-Print Network [OSTI]

The partition functions of refined topological strings(A-models) are computed, which give rise to the circle-compactified five-dimensional supersymmetric linear quiver gauge theories in generic (not necessarily self-dual) Omega backgrounds. Based on the slicing independence conjecture of refined topological string partition functions, it is demonstrated explicitly that the duality exists between $SU(N)^{M-1}$ and $SU(M)^{N-1}$ supersymmetric linear quiver gauge theories, even in generic Omega backgrounds. It is found that the relations between string moduli and gauge moduli are deformed from the self-dual case. However if the duality map which preserves the ratio of the Omega background parameters q and t, is considered, duality maps of the gauge moduli are not changed from the self-dual case.

Kei Ito

2012-11-29T23:59:59.000Z

336

Generic behaviour of nonlinear sound waves near the surface of a star: smooth solutions  

E-Print Network [OSTI]

We are interested in the generic behaviour of nonlinear sound waves as they approach the surface of a star, here assumed to have the polytropic equation of state $P=K\\rho^\\Gamma$. Restricting to spherical symmetry, and considering only the region near the surface, we generalise the methods of Carrier and Greenspan (1958) for the shallow water equations on a sloping beach to this problem. We give a semi-quantitative criterion for a shock to form near the surface during the evolution of generic initial data with support away from the surface. We show that in smooth solutions the velocity and the square of the sound speed remain regular functions of Eulerian radius at the surface.

Carsten Gundlach; Colin Please

2009-04-10T23:59:59.000Z

337

International Research Reactor Decommissioning Project  

SciTech Connect (OSTI)

Many research reactors have been or will be shut down and are candidates for decommissioning. Most of the respective countries neither have a decommissioning policy nor the required expertise and funds to effectively implement a decommissioning project. The IAEA established the Research Reactor Decommissioning Demonstration Project (R{sup 2}D{sup 2}P) to help answer this need. It was agreed to involve the Philippine Research Reactor (PRR-1) as model reactor to demonstrate 'hands-on' experience as it is just starting the decommissioning process. Other facilities may be included in the project as they fit into the scope of R{sup 2}D{sup 2}P and complement to the PRR-1 decommissioning activities. The key outcome of the R{sup 2}D{sup 2}P will be the decommissioning of the PRR-1 reactor. On the way to this final goal the preparation of safety related documents (i.e., decommissioning plan, environmental impact assessment, safety analysis report, health and safety plan, cost estimate, etc.) and the licensing process as well as the actual dismantling activities could provide a model to other countries involved in the project. It is expected that the R{sup 2}D{sup 2}P would initiate activities related to planning and funding of decommissioning activities in the participating countries if that has not yet been done.

Leopando, Leonardo [Philippine Nuclear Research Institute, Quezon City (Philippines); Warnecke, Ernst [International Atomic Energy Agency, Vienna (Austria)

2008-01-15T23:59:59.000Z

338

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

339

Distributed Generation  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

come a long way in addressing interconnection standards for distributed generation, utilities have largely overlooked the untapped potential of these resources. Under certain...

340

Reactor control rod timing system  

SciTech Connect (OSTI)

A fluid driven jet-edge whistle timing system for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (Above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

Wu, P.T.

1982-02-09T23:59:59.000Z

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.


341

Reactor control rod timing system  

DOE Patents [OSTI]

A fluid driven jet-edge whistle timing system for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

Wu, Peter T. K. (Clifton Park, NY)

1982-01-01T23:59:59.000Z

342

Computational Fluid Dynamics Simulation of Fluidized Bed Polymerization Reactors  

SciTech Connect (OSTI)

Fluidized beds (FB) reactors are widely used in the polymerization industry due to their superior heat- and mass-transfer characteristics. Nevertheless, problems associated with local overheating of polymer particles and excessive agglomeration leading to FB reactors defluidization still persist and limit the range of operating temperatures that can be safely achieved in plant-scale reactors. Many people have been worked on the modeling of FB polymerization reactors, and quite a few models are available in the open literature, such as the well-mixed model developed by McAuley, Talbot, and Harris (1994), the constant bubble size model (Choi and Ray, 1985) and the heterogeneous three phase model (Fernandes and Lona, 2002). Most these research works focus on the kinetic aspects, but from industrial viewpoint, the behavior of FB reactors should be modeled by considering the particle and fluid dynamics in the reactor. Computational fluid dynamics (CFD) is a powerful tool for understanding the effect of fluid dynamics on chemical reactor performance. For single-phase flows, CFD models for turbulent reacting flows are now well understood and routinely applied to investigate complex flows with detailed chemistry. For multiphase flows, the state-of-the-art in CFD models is changing rapidly and it is now possible to predict reasonably well the flow characteristics of gas-solid FB reactors with mono-dispersed, non-cohesive solids. This thesis is organized into seven chapters. In Chapter 2, an overview of fluidized bed polymerization reactors is given, and a simplified two-site kinetic mechanism are discussed. Some basic theories used in our work are given in detail in Chapter 3. First, the governing equations and other constitutive equations for the multi-fluid model are summarized, and the kinetic theory for describing the solid stress tensor is discussed. The detailed derivation of DQMOM for the population balance equation is given as the second section. In this section, monovariate population balance, bivariate population balance, aggregation and breakage equation and DQMOM-Multi-Fluid model are described. In the last section of Chapter 3, numerical methods involved in the multi-fluid model and time-splitting method are presented. Chapter 4 is based on a paper about application of DQMOM to polydisperse gas-solid fluidized beds. Results for a constant aggregation and breakage kernel and a kernel developed from kinetic theory are shown. The effect of the aggregation success factor and the fragment distribution function are investigated. Chapter 5 shows the work on validation of mixing and segregation phenomena in gas-solid fluidized beds with a binary mixture or a continuous size distribution. The simulation results are compared with available experiment data and discrete-particle simulation. Chapter 6 presents the project with Univation Technologies on CFD simulation of a Polyethylene pilot-scale FB reactor, The fluid dynamics, mass/heat transfer and particle size distribution are investigated through CFD simulation and validated with available experimental data. The conclusions of this study and future work are discussed in Chapter 7.

Rong Fan

2006-08-09T23:59:59.000Z

343

Calculation of Heating Values for the High Flux Isotope Reactor  

SciTech Connect (OSTI)

Calculating the amount of energy released by a fission reaction (fission Q value) and the heating rate distribution in a nuclear reactor is an important part of the safety analysis. However, these calculations can become very complex. One of the codes that can be used for this type of analyses is the Monte Carlo transport code MCNP5. Currently it is impossible to calculate the Q value and heating rate disposition for delayed beta and delayed gamma particles directly from MCNP5. The purpose of this paper is to outline a rigorous method for indirectly calculating the Q values and heating rates in the High Flux Isotope Reactor (HFIR), based on previous similar studies carried out for very high-temperature reactor configurations. This method has been applied in this study to calculate heating rates for the beginning of cycle (BOC) and end-of-cycle (EOC) states of HFIR. In addition, the BOC results obtained for HFIR are compared with corresponding results for the Advanced Test Reactor. The fission Q value for HFIR was calculated as 200.2 MeV for the BOC and 201.3 MeV for the EOC. It was also determined that 95.1% and 95.4% of the heat was deposited within the HFIR fuel plates for the BOC and EOC models, respectively. This methodology can also be used for heating rate calculations for HFIR experiments.

Peterson, Joshua L [ORNL] [ORNL; Ilas, Germina [ORNL] [ORNL

2012-01-01T23:59:59.000Z

344

Development of Improved Burnable Poisons for Commercial Nuclear Power Reactors  

SciTech Connect (OSTI)

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

345

Level III probabilistic risk assessment for N Reactor  

SciTech Connect (OSTI)

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

346

DOE's way-out reactors  

SciTech Connect (OSTI)

The SP-100 reactor, envisioned long before Star Wars, was to power civilian structures such as the space station and orbiting commercial labs. According to the SDI Organization, it will be the cornerstone for SDI, used as a no-maintenance, general source of energy for the military's infrastructure - weapons scale power will come later. DOE wants to spend $72 in FY 1977 to design and build these reactors. Funding problems with Congress, as well as some of the technology and timetables are discussed here.

Marshall, E.

1986-03-21T23:59:59.000Z

347

Nuclear reactor safety heat transfer  

SciTech Connect (OSTI)

Reviewed is a book which has 5 parts: Overview, Fundamental Concepts, Design Basis Accident-Light Water Reactors (LWRs), Design Basis Accident-Liquid-Metal Fast Breeder Reactors (LMFBRs), and Special Topics. It combines a historical overview, textbook material, handbook information, and the editor's personal philosophy on safety of nuclear power plants. Topics include thermal-hydraulic considerations; transient response of LWRs and LMFBRs following initiating events; various accident scenarios; single- and two-phase flow; single- and two-phase heat transfer; nuclear systems safety modeling; startup and shutdown; transient response during normal and upset conditions; vapor explosions, natural convection cooling; blockages in LMFBR subassemblies; sodium boiling; and Three Mile Island.

Jones, O.C.

1982-07-01T23:59:59.000Z

348

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

SciTech Connect (OSTI)

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

349

Comparative analysis of peat gasification reactor configuration. Final report  

SciTech Connect (OSTI)

This report summarizes the comparative analysis of two generic gasifiers (fluidized bed and entrained bed) and two specific hydrogasifiers (IGT's Peat-gas and Rockwell International's hydrogasifier). The objective is to establish a basis for recommending a peat gasification reactor configuration for testing in a DOE peat gasification pilot plant project. The approach involved the following four steps: data base evaluation, regression analysis, a chemical engineering evaluation of upstream and downstream equipment requirement, and computer simulation. Mathematical models and computer programs were developed to simulate the entrained-bed and fluidized-bed reactors. Parametric analyses were made, using these computer programs, to test the sensitivity and effects of significant operating variables (temperature, pressure and feed gas composition, for instance) on the product gas composition in the IGT gasifier and Rockwell International gasifier. This review and analysis concentrates primarily upon the Peatgas process of the Institute of Gas Technology and the Rockwell International Hydrogasification Process. The two-stage Peatgas process appears to have advantages in higher thermal efficiency, smaller capital investment, and its application of existing commercial operations and proven design practices from other types of fluidized solids processing units. There are two problems concerning operability of the Peatgas process: (1) the design of a fluidized solids unit for quite low standpipe densities (approximately 11 to 15 pounds per cubic foot), which may become even lower upon operating upsets, and (2) the potential problem of ash sintering in the very hot combustion zone of the steam-oxygen gasifier. On a relative time scale of development, the Peatgas would seem to be much closer to possible early commercialization than is the Rockwell system.

Not Available

1981-07-01T23:59:59.000Z

350

Uncertainties in the Anti-neutrino Production at Nuclear Reactors  

E-Print Network [OSTI]

reactors are determined from thermal power measure- ments and ?ssion rate calculations.of a reactor’s ther- mal power is given by a calculation ofCALCULATIONS During the power cycle of a nuclear reactor,

Djurcic, Zelimir

2009-01-01T23:59:59.000Z

351

Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation  

E-Print Network [OSTI]

Gas Expansion Module Gas-cooled Fast Reactor High Enrichedfast reactors: gas-cooled fast reactor (GFR), sodium-cooledderived from the Gas cooled Fast Reactor (GFR). This core

Heidet, Florent

2010-01-01T23:59:59.000Z

352

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

353

Knowledge and abilities catalog for nuclear power plant operators: boiling water reactors  

SciTech Connect (OSTI)

The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWR) (NUREG-1123) 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 and Examiners' Handbook for Developing Operator Licensing Examinations (NUREG-1121) will cover those topics listed under Title 10, Code of Federal Regulations, Part 55. The BWR Catalog contains approximately 7000 knowledge and ability (K/A) statements for ROs and SROs at boiling water reactors. Each K/A statement has been rated for its importance to the safe operation of the plant in a manner ensuring personnel and public health and safety. The BWR K/A Catalog is organized into five major sections: Plant-wide Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Function, Emergency and Abnormal Plant Evolutions, Components, and Theory. The BWR Catalog represents a modification of the form and content of the K/A Catalog for Nuclear Power Plant Operators: Pressurized Water Reactors (NUREG-1122). First, categories of knowledge and ability statements have been redefined. Second, the scope of the definition of emergency and abnormal plant evolutions has been revised in line with a symptom-based approach. Third, K/As related to the operational applications of theory have been incorporated into the delineations for both plant systems and emergency and abnormal plant evolutions, while K/As pertaining to theory fundamental to plant operation have been delineated in a separate theory section. Finally, the components section has been revised.

Not Available

1986-09-01T23:59:59.000Z

354

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

355

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

356

New fast-reactor approach. [LMFBR  

SciTech Connect (OSTI)

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

357

Stability analysis of supercritical water cooled reactors  

E-Print Network [OSTI]

The Supercritical Water-Cooled Reactor (SCWR) is a concept for an advanced reactor that will operate at high pressure (25MPa) and high temperature (500°C average core exit). The high coolant temperature as it leaves the ...

Zhao, Jiyun, Ph. D. Massachusetts Institute of Technology

2005-01-01T23:59:59.000Z

358

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

359

Things to Consider When Upgrading a Non-Power Reactor to a Digital I&C System  

SciTech Connect (OSTI)

Non-Power Reactor (NPR) licensees are increasing their use of state-of-the-art digital technology in instrumentation and control (I&C) systems because digital systems offer improved reactor control, information processing, and information storage. In Generic Letter GL 95-02, the NRC recognized that the design characteristics specific to the new digital electronics could result in failure modes and system malfunctions that either were not considered during the initial plant design or not evaluated in sufficient detail in the safety analysis report. These concerns include potential common mode failures. A conversion from analog to digital I&C systems in NPRs solves some problems while potentially introducing others. Good design, engineering, review, and testing can identify and minimize these risks.

Muhlheim, Michael David [ORNL] [ORNL; Hardin, LeRoy A [U.S. Nuclear Regulatory Commission] [U.S. Nuclear Regulatory Commission; Hardesty, Duane [U.S. Nuclear Regulatory Commission] [U.S. Nuclear Regulatory Commission; Wilson, Thomas L [ORNL] [ORNL

2011-01-01T23:59:59.000Z

360

LandScape Command Set: Local Area Network Distributed Supervisory Control and Programming Environment  

SciTech Connect (OSTI)

This paper presents the Local Area Network Distributed Supervisory Control and Programming Environment (LandScape) commands set that provides a Generic Device Subsystem Application Programmers Interface (API). These commands are implemented using the Common Object Request Broker Architecture (CORBA) specification with Orbix from Iona Technologies.

Burchard, R.L.; Small, D.E.

1999-01-01T23:59:59.000Z

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.


361

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

362

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

363

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

364

Westinghouse Advanced Reactors Division Plutonium Fuel Laboratories  

Office of Legacy Management (LM)

Radiological Condition of the Westinghouse Advanced Reactors Division Plutonium Fuel Laboratories Cheswick, Pennsylvania -. -, -- AGENCY: Office of Operational Safety, Department...

365

Light Water Reactor Sustainability (LWRS) Program  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

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

366

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

367

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

368

MULTIPHASE REACTOR MODELING FOR ZINC CHLORIDE CATALYZED COAL LIQUEFACTION  

E-Print Network [OSTI]

for the Coal Slurry Reactor Calculations are shown here for= Total reactor pressure, psi. The calculation is iterative,

Joyce, Peter James

2011-01-01T23:59:59.000Z

369

D Ris-R-406 Department of Reactor  

E-Print Network [OSTI]

of a Nuclear District Heating Reactor ... 17 3. REACTOR PHYSICS AND DYNAMICS 13 3.1. Core Follow Studies: Bti

370

Interdisciplinary Institute for Innovation Nuclear reactors' construction  

E-Print Network [OSTI]

Interdisciplinary Institute for Innovation Nuclear reactors' construction costs: The role of lead@mines-paristech.fr hal-00956292,version1-6Mar2014 #12;hal-00956292,version1-6Mar2014 #12;Nuclear reactors' construction reactor construction costs in France and the United States. Studying the cost of nuclear power has often

Paris-Sud XI, Université de

371

The Reactor An ObjectOriented Framework  

E-Print Network [OSTI]

The Reactor An Object­Oriented Framework for Event Demultiplexing and Event Handler Dispatching Douglas C. Schmidt 1 Overview ffl The Reactor is an object­oriented frame­ work that encapsulates OS event demul­ tiplexing mechanisms -- e.g., the Reactor API runs transparently atop both Wait

Schmidt, Douglas C.

372

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

373

Laminar Entrained Flow Reactor (Fact Sheet)  

SciTech Connect (OSTI)

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

374

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.

1995-11-07T23:59:59.000Z

375

Nozzle for electric dispersion reactor  

DOE Patents [OSTI]

A nozzle for an electric dispersion reactor includes two coaxial cylindrical bodies, 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. 5 figs.

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

1998-06-02T23:59:59.000Z

376

Nozzle for electric dispersion reactor  

DOE Patents [OSTI]

A nozzle for an electric dispersion reactor includes two coaxial cylindrical bodies, 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. 5 figs.

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

1996-04-02T23:59:59.000Z

377

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

378

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)

1995-01-01T23:59:59.000Z

379

Nozzle for electric dispersion reactor  

DOE Patents [OSTI]

A nozzle for an electric dispersion reactor includes two coaxial cylindrical bodies, 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); Harris, Michael T. (Knoxville, TN); Scott, Timothy C. (Knoxville, TN); Basaran, Osman A. (Oak Ridge, TN)

1998-01-01T23:59:59.000Z

380

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

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

Nozzle for electric dispersion reactor  

DOE Patents [OSTI]

A nozzle for an electric dispersion reactor includes two coaxial cylindrical bodies, 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); Harris, Michael T. (Knoxville, TN); Scott, Timothy C. (Knoxville, TN); Basaran, Osman A. (Oak Ridge, TN)

1996-01-01T23:59:59.000Z

382

Nuclear Reactor Safety Design Criteria  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

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

383

Starfire - a commercial tokamak reactor  

SciTech Connect (OSTI)

The basic objective of the STARFIRE Project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. 10 refs.

Baker, C.C.; Abdou, M.A.; DeFreece, D.A.; Trachsel, C.A.; Kokoszenski, J.; Graumann, D.

1981-01-01T23:59:59.000Z

384

Advanced Test Reactor probabilistic risk assessment methodology and results summary  

SciTech Connect (OSTI)

The Advanced Test Reactor (ATR) probabilistic risk assessment (PRA) Level 1 report documents a comprehensive and state-of-the-art study to establish and reduce the risk associated with operation of the ATR, expressed as a mean frequency of fuel damage. The ATR Level 1 PRA effort is unique and outstanding because of its consistent and state-of-the-art treatment of all facets of the risk study, its comprehensive and cost-effective risk reduction effort while the risk baseline was being established, and its thorough and comprehensive documentation. The PRA includes many improvements to the state-of-the-art, including the following: establishment of a comprehensive generic data base for component failures, treatment of initiating event frequencies given significant plant improvements in recent years, performance of efficient identification and screening of fire and flood events using code-assisted vital area analysis, identification and treatment of significant seismic-fire-flood-wind interactions, and modeling of large loss-of-coolant accidents (LOCAs) and experiment loop ruptures leading to direct damage of the ATR core. 18 refs.

Eide, S.A.; Atkinson, S.A.; Thatcher, T.A.

1992-01-01T23:59:59.000Z

385

Nuclear reactor safeguards and monitoring with antineutrino detectors A. Bernsteina)  

E-Print Network [OSTI]

Nuclear reactor safeguards and monitoring with antineutrino detectors A. Bernsteina) Sandia of nuclear reactor types, including power reactors, research reactors, and plutonium production reactors-understood principles that govern the core's evolution in time, can be used to determine whether the reactor is being

Gratta, Giorgio

386

Spectral shift reactor control method  

SciTech Connect (OSTI)

This patent describes the method of operating a pressurized-water fissile-material-fueled spectral-shift nuclear reactor in such manner that short-term reactivity requirement variations can be satisfied without making control rod or chemical shim changes. The reactor includes a pressure vessel enclosing a reactor core and having an inlet and an outlet for circulating a water coolant moderator in heat transfer relationship with the core. The core comprises fuel assemblies disposed therein for generating heat by nuclear fission. The reactor provided with neutron-absorbing control rods which are vertically movable into and out of the core so that movement of the control rods into the core will substantially decrease reactivity and withdrawal of the control rods from the core will substantially increase reactivity. The control rods when inserted into the core displace an equivalent volume of the water coolant moderator. The reactor also provides neutron-spectral-shift rods which have a lower absorptivity for neutrons than the control rods, the neutron-spectral shift rods when inserted into the core displacing an equaivalent volume of the water coolant moderator. The neutron-spectral-shift rods comprises two different types of rods, a first of the different types of the neutron-spectral-shift rods comprising displacer rods which have a low absorptivity for neutrons, the remainder of the neutron-spectral-shift rods comprising gray rods which have an absorption for neutrons which is intermediate the neutron absorption of the control rods and the low neutron absorption of the displacer rods. Each neutron-spectral-shift displacer rod comprises a hollow thin-walled Zircaloy member containing a filling of solid or annular zirconium- or aluminum-containing material for providing internal support and mass for the thin-walled tubular member.

Impink, A.J. Jr.

1987-12-29T23:59:59.000Z

387

Heterogeneous Recycling in Fast Reactors  

SciTech Connect (OSTI)

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

388

Distributed DBMS Introduction  

E-Print Network [OSTI]

Distributed DBMS Outline Introduction What is a distributed DBMS Problems Current state-of-affairs Background Distributed DBMS Architecture Distributed Database Design Semantic Data Control Distributed Query Processing Distributed Transaction Management Parallel Database Systems Distributed Object DBMS

Chen, Yangjun

389

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

390

Neurocontrol of Pressurized Water Reactors in Load-Follow Operations  

SciTech Connect (OSTI)

The neurocontrol technique was applied to control a pressurized water reactor (PWR) in load-follow operations. Generalized learning or direct inverse control architecture was adopted in which the neural network was trained off-line to learn the inverse model of the PWR. Two neural network controllers were designed: One provided control rod position, which controlled the axial power distribution, and the other provided the change in boron concentration, which adjusted core total power. An additional feedback controller was designed so that power tracking capability was improved. The time duration between control actions was 15 min; thus, the xenon effect is limited and can be neglected. Therefore, the xenon concentration was not considered as a controller input variable, which simplified controller design. Center target strategy and minimum boron strategy were used to operate the reactor, and the simulation results demonstrated the effectiveness and performance of the proposed controller.

Lin Chaung; Shen Chihming

2000-12-15T23:59:59.000Z

391

Diamond neutral particle spectrometer for fusion reactor ITER  

SciTech Connect (OSTI)

A compact diamond neutral particle spectrometer with digital signal processing has been developed for fast charge-exchange atoms and neutrons measurements at ITER fusion reactor conditions. This spectrometer will play supplementary role for Neutral Particle Analyzer providing 10 ms time and 30 keV energy resolutions for fast particle spectra in non-tritium ITER phase. These data will also be implemented for independent studies of fast ions distribution function evolution in various plasma scenarios with the formation of a single fraction of high-energy ions. In tritium ITER phase the DNPS will measure 14 MeV neutrons spectra. The spectrometer with digital signal processing can operate at peak counting rates reaching a value of 10{sup 6} cps. Diamond neutral particle spectrometer is applicable to future fusion reactors due to its high radiation hardness, fast response and high energy resolution.

Krasilnikov, V.; Amosov, V.; Kaschuck, Yu.; Skopintsev, D. [Institution PROJECT CENTER ITER, 1, Akademik Kurchatov Sq., Moscow (Russian Federation)

2014-08-21T23:59:59.000Z

392

Characterization of the TRIGA Mark II reactor full-power steady state  

E-Print Network [OSTI]

In this work, the characterization of the full-power steady state of the TRIGA Mark II nuclear reactor of the University of Pavia is performed by coupling Monte Carlo (MC) simulation for neutronics with "Multiphysics" model for thermal-hydraulics. Neutronic analyses have been performed starting from a MC model of the entire reactor system, based on the MCNP5 code, that was already validated in fresh fuel and zero-power configuration (in which thermal effects are negligible) using the available experimental data as benchmark. In order to describe the full-power reactor configuration, the temperature distribution in the core is necessary. To evaluate it, a thermal-hydraulic model has been developed, using the power distribution results from MC simulation as input. The thermal-hydraulic model is focused on the core active region and takes into account sub-cooled boiling effects present at full reactor power. The obtained temperature distribution is then introduced in the MC model and a benchmark analysis is carried out to validate the model in fresh fuel and full-power configuration. The good agreement between experimental data and simulation results concerning full-power reactor criticality, proves the reliability of the adopted methodology of analysis, both from neutronics and thermal-hydraulics perspective.

Antonio Cammi; Matteo Zanetti; Davide Chiesa; Massimiliano Clemenza; Stefano Pozzi; Ezio Previtali; Monica Sisti; Giovanni Magrotti; Michele Prata; Andrea Salvini

2015-03-03T23:59:59.000Z

393

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

394

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. 8 figs.

Groh, E.F.; Olson, A.P.; Wade, D.C.; Robinson, B.W.

1984-06-05T23:59:59.000Z

395

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

396

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

397

2007 DOE Bio-Derived Liquids to Hydrogen Distributed Reforming Working Group (BILIWG) Meeting  

E-Print Network [OSTI]

Biogas reformation Shift reactorShift reactor Selective oxidation of COSelective oxidation of CO Fuel cellFuel cell2007 DOE Bio-Derived Liquids to Hydrogen Distributed Reforming Working Group (BILIWG) Meeting Investigation of Bio-Ethanol Steam Reforming over Cobalt-based Catalysts Hua Song Lingzhi Zhang Umit S

398

Synthesis and Manipulation of Semiconductor Nanocrystals inMicrofluidic Reactors  

SciTech Connect (OSTI)

Microfluidic reactors are investigated as a mechanism tocontrol the growth of semiconductor nanocrystals and characterize thestructural evolution of colloidal quantum dots. Due to their shortdiffusion lengths, low thermal masses, and predictable fluid dynamics,microfluidic devices can be used to quickly and reproducibly alterreaction conditions such as concentration, temperature, and reactiontime, while allowing for rapid reagent mixing and productcharacterization. These features are particularly useful for colloidalnanocrystal reactions, which scale poorly and are difficult to controland characterize in bulk fluids. To demonstrate the capabilities ofnanoparticle microreactors, a size series of spherical CdSe nanocrystalswas synthesized at high temperature in a continuous-flow, microfabricatedglass reactor. Nanocrystal diameters are reproducibly controlled bysystematically altering reaction parameters such as the temperature,concentration, and reaction time. Microreactors with finer control overtemperature and reagent mixing were designed to synthesize nanoparticlesof different shapes, such as rods, tetrapods, and hollow shells. The twomajor challenges observed with continuous flow reactors are thedeposition of particles on channel walls and the broad distribution ofresidence times that result from laminar flow. To alleviate theseproblems, I designed and fabricated liquid-liquid segmented flowmicroreactors in which the reaction precursors are encapsulated inflowing droplets suspended in an immiscible carrier fluid. The synthesisof CdSe nanocrystals in such microreactors exhibited reduced depositionand residence time distributions while enabling the rapid screening aseries of samples isolated in nL droplets. Microfluidic reactors werealso designed to modify the composition of existing nanocrystals andcharacterize the kinetics of such reactions. The millisecond kinetics ofthe CdSe-to-Ag2Se nanocrystal cation exchange reaction are measured insitu with micro-X-ray Absorption Spectroscopy in silicon microreactorsspecifically designed for rapid mixing and time-resolved X-rayspectroscopy. These results demonstrate that microreactors are valuablefor controlling and characterizing a wide range of reactions in nLvolumes even when nanoscale particles, high temperatures, causticreagents, and rapid time scales are involved. These experiments providethe foundation for future microfluidic investigations into the mechanismsof nanocrystal growth, crystal phase evolution, and heterostructureassembly.

Chan, Emory Ming-Yue

2006-12-19T23:59:59.000Z

399

Research Program of a Super Fast Reactor  

SciTech Connect (OSTI)

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

400

Analysis of Effectiveness of Lyapunov Control for Non-generic Quantum States  

E-Print Network [OSTI]

A Lyapunov-based control design for natural trajectory-tracking problems is analyzed for quantum states where the analysis in the generic case is not applicable. Using dynamical systems tools we show almost global asymptotic stability for stationary target states subject to certain conditions on the Hamiltonians, and discuss effectiveness of the design when these conditions are not satisfied. For pseudo-pure target states the effectiveness of the design is studied further for both stationary and non-stationary states using alternative tools.

Xiaoting Wang; Sonia Schirmer

2009-10-01T23:59:59.000Z

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.


401

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

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth7-1D: Vegetation ProposedUsingFun with Big Sky Learning Fun withGenepoolCrystals. |Generic TriBITS PRoject,

402

Safety control circuit for a neutronic reactor  

DOE Patents [OSTI]

A neutronic reactor comprising an active portion containing material fissionable by neutrons of thermal energy, means to control a neutronic chain reaction within the reactor comprising a safety device and a regulating device, a safety device including means defining a vertical channel extending into the reactor from an aperture in the upper surface of the reactor, a rod containing neutron-absorbing materials slidably disposed within the channel, means for maintaining the safety rod in a withdrawn position relative to the active portion of the reactor including means for releasing said rod on actuation thereof, a hopper mounted above the active portion of the reactor having a door disposed at the bottom of the hopper opening into the vertical channel, a plurality of bodies of neutron-absorbing materials disposed within the hopper, and means responsive to the failure of the safety rod on actuation thereof to enter the active portion of the reactor for opening the door in the hopper.

Ellsworth, Howard C. (Richland, WA)

2004-04-27T23:59:59.000Z

403

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

404

Fast-acting nuclear reactor control device  

SciTech Connect (OSTI)

A fast-acting nuclear reactor control device is described for controlling a safety control rod within the core of a nuclear reactor, the reactor controlled by a reactor control system, the device comprising: a safety control rod drive shaft and an electromagnetic clutch co-axial with the drive shaft operatively connected to the safety control rod for driving and positioning the safety control rod within or without the reactor core during reactor operation, the safety rod being oriented in a substantially vertical position to allow the rod to fall into the reactor core under the influence of gravity during shutdown of the reactor; the safety control rod drive shaft further operatively connected to a hydraulic pump such that operation of the drive shaft simultaneously drives and positions the safety control rod and operates the hydraulic pump such that a hydraulic fluid is forced into an accumulator, filling the accumulator with oil for the storage and supply of primary potential energy for safety control rod insertion such that the release of potential energy in the accumulator causes hydraulic fluid to flow through the hydraulic pump, converting the hydraulic pump to a hydraulic motor having speed and power capable of full length insertion and high speed driving of the safety control rod into the reactor core; a solenoid valve interposed between the hydraulic pump and the accumulator, said solenoid valve being a normally open valve, actuated to close when the safety control rod is out of the reactor during reactor operation; and further wherein said solenoid opens in response to a signal from the reactor control system calling for shutdown of the reactor and rapid insertion of the safety control rod into the reactor core, such that the opening of the solenoid releases the potential energy in the accumulator to place the safety control rod in a safe shutdown position.

Kotlyar, O.M.; West, P.B.

1993-08-03T23:59:59.000Z

405

Grid integrated distributed PV (GridPV).  

SciTech Connect (OSTI)

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

406

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

407

Fourth annual progress report on special-purpose materials for magnetically confined fusion reactors  

SciTech Connect (OSTI)

The scope of Special Purpose Materials covers fusion reactor materials problems other than the first-wall and blanket structural materials, which are under the purview of the ADIP, DAFS, and PMI task groups. Components that are considered as special purpose materials include breeding materials, coolants, neutron multipliers, barriers for tritium control, materials for compression and OH coils and waveguides, graphite and SiC, heat-sink materials, ceramics, and materials for high-field (>10-T) superconducting magnets. The Task Group on Special Purpose Materials has limited its concern to crucial and generic materials problems that must be resolved if magnetic-fusion devices are to succeed. Important areas specifically excluded include low-field (8-T) superconductors, fuels for hybrids, and materials for inertial-confinement devices. These areas may be added in the future when funding permits.

Not Available

1982-08-01T23:59:59.000Z

408

Reactor safeguards against insider sabotage  

SciTech Connect (OSTI)

A conceptual safeguards system is structured to show how both reactor operations and physical protection resources could be integrated to prevent release of radioactive material caused by insider sabotage. Operational recovery capabilities are addressed from the viewpoint of both detection of and response to disabled components. Physical protection capabilities for preventing insider sabotage through the application of work rules are analyzed. Recommendations for further development of safeguards system structures, operational recovery, and sabotage prevention are suggested.

Bennett, H.A.

1982-03-01T23:59:59.000Z

409

Nuclear power reactor education and training at the Ford nuclear reactor  

SciTech Connect (OSTI)

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

410

Updating reactor control: mini-computers  

SciTech Connect (OSTI)

An aging reactor control console and a limited operating budget have impeded many research projects in the TRIGA reactor facility at the University of Utah. The, University's present console is Circa 1959 vintage and repairs to the console are frequently required which present many electronic problems to a staff with little electronic training. As an alternative to a single function control console we are developing a TRIGA control system based upon a mini-computer. The system hardware has been specified and the hardware is currently being acquired. The software will be programmed by the staff to customize the system to the reactor's physical systems and technical specifications. The software will be designed to monitor and control all reactor functions, control a pneumatic sample transfer system, acquire and analyze neutron activation data, provide reactor facility security surveillance, provide reactor documentation including online logging of physical parameters, and record regularly scheduled reactor calibrations and laboratory accounting procedures. The problem of hardware rewiring and changing technical specifications and changing safety system characteristics can be easily handled in the software. Our TRIGA reactor also functions as a major educational resource using available reactor based software. The computer control system can be employed to provide on-line training in reactor physics and kinetics. (author)

Crawford, K.C.; Sandquist, G.M. [University of Utah, Salt Lake City, UT (United States)

1984-07-01T23:59:59.000Z

411

Advanced Safeguards Approaches for New Fast Reactors  

SciTech Connect (OSTI)

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

412

Reference worldwide model for antineutrinos from reactors  

E-Print Network [OSTI]

Antineutrinos produced at nuclear reactors constitute a severe source of background for the detection of geoneutrinos, which bring to the Earth's surface information about natural radioactivity in the whole planet. In this framework we provide a reference worldwide model for antineutrinos from reactors, in view of reactors operational records yearly published by the International Atomic Energy Agency (IAEA). We evaluate the expected signal from commercial reactors for ongoing (KamLAND and Borexino), planned (SNO+) and proposed (Juno, RENO-50, LENA and Hanohano) experimental sites. Uncertainties related to reactor antineutrino production, propagation and detection processes are estimated using a Monte Carlo based approach, which provides an overall site dependent uncertainty on the signal in the geoneutrino energy window on the order of 3%. We also implement the off-equilibrium correction to the reference reactor spectra associated with the long-lived isotopes and we estimate a 2.4% increase of the unoscillated event rate in the geoneutrino energy window due to the storage of spent nuclear fuels in the cooling pools. We predict that the research reactors contribute to less than 0.2% to the commercial reactor signal in the investigated 14 sites. We perform a multitemporal analysis of the expected reactor signal over a time lapse of 10 years using reactor operational records collected in a comprehensive database published at www.fe.infn.it/antineutrino.

Marica Baldoncini; Ivan Callegari; Giovanni Fiorentini; Fabio Mantovani; Barbara Ricci; Virginia Strati; Gerti Xhixha

2015-02-16T23:59:59.000Z

413

Fabrication Flaws in Reactor Pressure Vessel Repair Welds  

SciTech Connect (OSTI)

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

414

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

415

E-Print Network 3.0 - advanced reactors division Sample Search...  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Historical Context of the Omega Reactor Facility, Technical Area 2 Volume 1 "Water... Boiler" Reactor (SUPO) Schematic Omega West Reactor Clementine Reactor ......

416

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

Broader source: All U.S. Department of Energy (DOE) Office Webpages (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...

417

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 (OSTI)

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

418

Molten-Salt Depleted-Uranium Reactor  

E-Print Network [OSTI]

The supercritical, reactor core melting and nuclear fuel leaking accidents have troubled fission reactors for decades, and greatly limit their extensive applications. Now these troubles are still open. Here we first show a possible perfect reactor, Molten-Salt Depleted-Uranium Reactor which is no above accident trouble. We found this reactor could be realized in practical applications in terms of all of the scientific principle, principle of operation, technology, and engineering. Our results demonstrate how these reactors can possess and realize extraordinary excellent characteristics, no prompt critical, long-term safe and stable operation with negative feedback, closed uranium-plutonium cycle chain within the vessel, normal operation only with depleted-uranium, and depleted-uranium high burnup in reality, to realize with fission nuclear energy sufficiently satisfying humanity long-term energy resource needs, as well as thoroughly solve the challenges of nuclear criticality safety, uranium resource insuffic...

Dong, Bao-Guo; Gu, Ji-Yuan

2015-01-01T23:59:59.000Z

419

PID Control Effectiveness for Surface Reactor Concepts  

SciTech Connect (OSTI)

Control of space and surface fission reactors should be kept as simple as possible, because of the need for high reliability and the difficulty to diagnose and adapt to control system failures. Fortunately, compact, fast-spectrum, externally controlled reactors are very simple in operation. In fact, for some applications it may be possible to design low-power surface reactors without the need for any reactor control after startup; however, a simple proportional, integral, derivative (PID) controller can allow a higher performance concept and add more flexibility to system operation. This paper investigates the effectiveness of a PID control scheme for several anticipated transients that a surface reactor might experience. To perform these analyses, the surface reactor transient code FRINK was modified to simulate control drum movements based on bulk coolant temperature.

Dixon, David D. [North Carolina State University, Raleigh, NC (United States); Los Alamos National Laboratory, Los Alamos, NM (United States); Marsh, Christopher L. [United States Naval Academy, Annapolis, MD (United States); Los Alamos National Laboratory, Los Alamos, NM (United States); Poston, David I. [Los Alamos National Laboratory, Los Alamos, NM (United States)

2007-01-30T23:59:59.000Z

420

Ordered bed modular reactor design proposal  

SciTech Connect (OSTI)

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

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

Application of Covariances to Fast Reactor Core Analysis  

SciTech Connect (OSTI)

In the present paper, the current status of covariance applications to fast reactor analysis and design is summarized with actual examples. The covariance applications are classified into three fields. First, covariances are used to quantify the uncertainty of nuclear core parameters such as criticality, control rod worth, reaction rate ratio, power distribution, sodium void reactivity, etc. The special features of the Japanese case are to include both the burnup-related parameters such as burnup reactivity loss or fuel composition changes, and the temperature-related parameter, that is, the Doppler reactivity. Second, covariances are used to select the important nuclides, reactions and energy ranges which are dominant to the uncertainty of core parameters, and to give nuclear scientists recommendations to improve the accuracy of the data. Finally, covariances are used to improve the accuracy of core design values by adopting the integral data such as the critical experiments and the power reactor operation data. The way toward improvement is classified into the conventional E/C (Experiment/Calculation) bias method and the more comprehensive cross section adjustment based on the Bayesian theorem and the generalized least square method. In Japan, an adjusted group-constant set, ADJ2000R, is now being used in the design work for future fast reactors.

Ishikawa, M. [Japan Atomic Energy Agency (JAEA), O-arai, Ibaraki, 311-1393 (Japan)], E-mail: ishikawa.makoto@jaea.go.jp

2008-12-15T23:59:59.000Z

422

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

423

Small Reactor for Deep Space Exploration  

SciTech Connect (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

2012-11-29T23:59:59.000Z

424

Nuclear reactor shield including magnesium oxide  

DOE Patents [OSTI]

An improvement in nuclear reactor shielding of a type used in reactor applications involving significant amounts of fast neutron flux, the reactor shielding including means providing structural support, neutron moderator material, neutron absorber material and other components as described below, wherein at least a portion of the neutron moderator material is magnesium in the form of magnesium oxide either alone or in combination with other moderator materials such as graphite and iron.

Rouse, Carl A. (Del Mar, CA); Simnad, Massoud T. (La Jolla, CA)

1981-01-01T23:59:59.000Z

425

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

426

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

427

Multi channel thermal hydraulic analysis of gas cooled fast reactor using genetic algorithm  

SciTech Connect (OSTI)

There are three analyzes to be done in the design process of nuclear reactor i.e. neutronic analysis, thermal hydraulic analysis and thermodynamic analysis. The focus in this article is the thermal hydraulic analysis, which has a very important role in terms of system efficiency and the selection of the optimal design. This analysis is performed in a type of Gas Cooled Fast Reactor (GFR) using cooling Helium (He). The heat from nuclear fission reactions in nuclear reactors will be distributed through the process of conduction in fuel elements. Furthermore, the heat is delivered through a process of heat convection in the fluid flow in cooling channel. Temperature changes that occur in the coolant channels cause a decrease in pressure at the top of the reactor core. The governing equations in each channel consist of mass balance, momentum balance, energy balance, mass conservation and ideal gas equation. The problem is reduced to finding flow rates in each channel such that the pressure drops at the top of the reactor core are all equal. The problem is solved numerically with the genetic algorithm method. Flow rates and temperature distribution in each channel are obtained here.

Drajat, R. Z.; Su'ud, Z.; Soewono, E.; Gunawan, A. Y. [Department of Mathematics, Institut Teknologi Bandung, Bandung 40132 (Indonesia); Department of Physics, Institut Teknologi Bandung, Bandung 40132 (Indonesia); Department of Mathematics, Institut Teknologi Bandung, Bandung 40132 (Indonesia)

2012-05-22T23:59:59.000Z

428

Comparison of Reactivity Control Systems for the Submersion Subcritical Safe Space (S and 4) Reactor  

SciTech Connect (OSTI)

This paper compares the effectiveness of two control mechanisms for the S and 4 reactor, namely: (a) rotating BeO drums with 120 deg. thin segments of enriched B{sub 4}C in the radial reflector; and (b) sliding windows in the radial reflector. Investigated are the effects of using these control mechanisms on the differential reactor control worth, power generation profiles, and spatial neutrons flux distributions is the S and 4 reactor. For both control mechanism, the radial reflector has the same dimensions and volume. Results show that the difference in reactor performance with the two control mechanisms is small. The sliding reflector configuration features slightly lower mass and power peaking, and relatively more even fission power profiles in the core. The differential control worth for the sliding reflector segments is almost constant compared to that using rotating control drums, potentially simplifying the reactor control operation. The presence of a strong neutron absorber in the rotating drums slightly decreases the amount of excess reactivity at BOL compared with the sliding reflector configuration. However, the higher rate of reactivity depletion in the S and 4 reactor with the latter may negate this advantage.

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

2008-01-21T23:59:59.000Z

429

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

430

Heat dissipating nuclear reactor with metal liner  

DOE Patents [OSTI]

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 is described in this disclosure. 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, E.L.; Hunsbedt, A.; Lazarus, J.D.

1985-11-21T23:59:59.000Z

431

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

432

TA-2 Water Boiler Reactor Decommissioning Project  

SciTech Connect (OSTI)

This final report addresses the Phase 2 decommissioning of the Water Boiler Reactor, biological shield, other components within the biological shield, and piping pits in the floor of the reactor building. External structures and underground piping associated with the gaseous effluent (stack) line from Technical Area 2 (TA-2) Water Boiler Reactor were removed in 1985--1986 as Phase 1 of reactor decommissioning. The cost of Phase 2 was approximately $623K. The decommissioning operation produced 173 m{sup 3} of low-level solid radioactive waste and 35 m{sup 3} of mixed waste. 15 refs., 25 figs., 3 tabs.

Durbin, M.E. (ed.); Montoya, G.M.

1991-06-01T23:59:59.000Z

433

Thermomechanical analysis of fast-burst reactors  

SciTech Connect (OSTI)

Fast-burst reactors are designed to provide intense, short-duration pulses of neutrons. The fission reaction also produces extreme time-dependent heating of the nuclear fuel. An existing transient-dynamic finite element code was modified specifically to compute the time-dependent stresses and displacements due to thermal shock loads of reactors. Thermomechanical analysis was then applied to determine structural feasibility of various concepts for an EDNA-type reactor and to optimize the mechanical design of the new SPR III-M reactor.

Miller, J.D.

1994-08-01T23:59:59.000Z

434

Heavy Water Test Reactor Dome Removal  

SciTech Connect (OSTI)

A high speed look at the removal of the Heavy Water Test Reactor Dome Removal. A project sponsored by the Recovery Act on the Savannah River Site.

None

2011-01-01T23:59:59.000Z

435

Advanced Reactor Thermal Hydraulic Modeling | Argonne Leadership...  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Advanced Reactor Thermal Hydraulic Modeling PI Name: Paul Fischer PI Email: fischer@mcs.anl.gov Institution: Argonne National Laboratory Allocation Program: INCITE Allocation Hours...

436

Advanced Reactor Thermal Hydraulic Modeling | Argonne Leadership...  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Fischer (ANL), Aleks Obabko (ANL), and Hank Childs (LBNL) Advanced Reactor Thermal Hydraulic Modeling PI Name: Paul Fischer PI Email: fischer@mcs.anl.gov Institution: Argonne...

437

Progress Update: P-Reactor Grout  

SciTech Connect (OSTI)

A progress update, the Recovery Act at work at the Savannah River Site. The new phase of work on the permanent closure of two cold war nuclear reactors.

Cody, Tom

2010-01-01T23:59:59.000Z

438

Progress Update: P-Reactor Grout  

ScienceCinema (OSTI)

A progress update, the Recovery Act at work at the Savannah River Site. The new phase of work on the permanent closure of two cold war nuclear reactors.

Cody, Tom

2012-06-14T23:59:59.000Z

439

Light Water Reactor Sustainability Program Contact Information  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (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...

440

Recovery Act Progress Update: Reactor Closure Feature  

SciTech Connect (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

2010-01-01T23:59:59.000Z

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.


441

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

442

Accelerators for Subcritical Molten-Salt Reactors  

SciTech Connect (OSTI)

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

443

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

444

Italian hybrid and fission reactors scenario analysis  

SciTech Connect (OSTI)

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

445

A compilation of reports of the Advisory Committee on Reactor Safeguards, 1997 annual, U.S. Nuclear Regulatory Commission. Volume 19  

SciTech Connect (OSTI)

This compilation contains 67 ACRS reports submitted to the Commission, or to the Executive Director for Operations, during calendar year 1997. It also includes a report to the Congress on the NRC Safety Research Program. Specific topics include: (1) advanced reactor designs, (2) emergency core cooling systems, (3) fire protection, (4) generic letters and issues, (5) human factors, (6) instrumentation, control and protection systems, (7) materials engineering, (8) probabilistic risk assessment, (9) regulatory guides and procedures, rules, regulations, and (10) safety research, philosophy, technology and criteria.

NONE

1998-04-01T23:59:59.000Z

446

Reactor pressure vessel vented head  

DOE Patents [OSTI]

A head for closing a nuclear reactor pressure vessel shell includes an arcuate dome having an integral head flange which includes a mating surface for sealingly mating with the shell upon assembly therewith. The head flange includes an internal passage extending therethrough with a first port being disposed on the head mating surface. A vent line includes a proximal end disposed in flow communication with the head internal passage, and a distal end disposed in flow communication with the inside of the dome for channeling a fluid therethrough. The vent line is fixedly joined to the dome and is carried therewith when the head is assembled to and disassembled from the shell.

Sawabe, James K. (San Jose, CA)

1994-01-11T23:59:59.000Z

447

Pollution prevention through reactor design  

SciTech Connect (OSTI)

Generation of waste in the chemical processing industries has its beginning in the heart of the process--the reaction system. Pollution prevention will have the greatest impact in minimizing the generation of waste through the design and operation of chemical reactors by reducing generation at the source--source reduction. Pollution prevention by modification of reaction parameters is defined as changing the selectivity of the reaction so that undesirable reactions which produce waste products are minimized while at the same time producing the desirable products.

Hopper, J.R. [Lamar Univ., Beaumont, TX (United States)

1995-09-01T23:59:59.000Z

448

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

449

Advanced Neutron Source reactor control and plant protection systems design  

SciTech Connect (OSTI)

This paper describes the reactor control and plant protection systems' conceptual design of the Advanced Neutron Source (ANS). The Plant Instrumentation, Control, and Data Systems and the Reactor Instrumentation and Control System of the ANS are planned as an integrated digital system with a hierarchical, distributed control structure of qualified redundant subsystems and a hybrid digital/analog protection system to achieve the necessary fast response for critical parameters. Data networks transfer information between systems for control, display, and recording. Protection is accomplished by the rapid insertion of negative reactivity with control rods or other reactivity mechanisms to shut down the fission process and reduce heat generation in the fuel. The shutdown system is designed for high functional reliability by use of conservative design features and a high degree of redundance and independence to guard against single failures. Two independent reactivity control systems of different design principles are provided, and each system has multiple independent rods or subsystems to provide appropriate margin for malfunctions such as stuck rods or other single failures. Each system is capable of maintaining the reactor in a cold shutdown condition independently of the functioning of the other system. A highly reliable, redundant channel control system is used not only to achieve high availability of the reactor, but also to reduce challenges to the protection system by maintaining important plant parameters within appropriate limits. The control system has a number of contingency features to maintain acceptable, off-normal conditions in spite of limited control or plant component failures thereby further reducing protection system challenges.

Anderson, J.L.; Battle, R.E.; March-Leuba, J. (Oak Ridge National Lab., TN (United States)); Khayat, M.I. (Tennessee Univ., Knoxville, TN (United States))

1992-01-01T23:59:59.000Z

450

Systematic speedup of path integrals of a generic N-fold discretized theory  

SciTech Connect (OSTI)

We present and discuss a detailed derivation of an analytical method that systematically improves the convergence of path integrals of a generic N-fold discretized theory. We develop an explicit procedure for calculating a set of effective actions S{sup (p)}, for p=1,2,3,... which have the property that they lead to the same continuum amplitudes as the starting action, but converge to that continuum limit ever faster. Discretized amplitudes calculated using the p-level effective action differ from the continuum limit by a term of order 1/N{sup p}. We obtain explicit expressions for the effective actions for levels p{<=}9. We end by analyzing the speedup of Monte Carlo simulations of two different models: an anharmonic oscillator with quartic coupling and a particle in a modified Poeschl-Teller potential.

Bogojevic, A.; Balaz, A.; Belic, A. [Institute of Physics, P.O. Box 57, 11001 Belgrade (Serbia and Montenegro)

2005-08-01T23:59:59.000Z

451

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

452

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

453

First results from electron-photon damage equivalence studies on a generic ethylene-propylene rubber  

SciTech Connect (OSTI)

As part of a simulator adequacy assessment program, the relative effectiveness of electrons and photons to produce damage in a generic ethylene propylene rubber (EPR) has been investigated. The investigation was limited in extent in that a single EPR material, in three thickness, was exposed to Cobalt-60 photons and three electron beam energies. Basing material damage on changes in the EPR mechanical properties elongation and tensile strength, we observed that EPR damage was a smoothly varying function of absorbed energy and independent of irradiating particle type. EPR damage tracked equally well as a function of both incident particle energy and material front surface dose. Based on these preliminary data, we tentatively concluded that a correlation between particle, particle energy, and material damage (as measured by changes in material elongation and/or tensile strength) has been demonstrated. 14 figs.

Buckalew, W.H.

1986-04-01T23:59:59.000Z

454

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

SciTech Connect (OSTI)

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

455

Feasibility of establishing and operating a generic oil shale test facility  

SciTech Connect (OSTI)

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

456

Use of freeze-casting in advanced burner reactor fuel design  

SciTech Connect (OSTI)

This paper will detail the modeling of a fast reactor with fuel pins created using a freeze-casting process. Freeze-casting is a method of creating an inert scaffold within a fuel pin. The scaffold is created using a directional solidification process and results in open porosity for emplacement of fuel, with pores ranging in size from 300 microns to 500 microns in diameter. These pores allow multiple fuel types and enrichments to be loaded into one fuel pin. Also, each pore could be filled with varying amounts of fuel to allow for the specific volume of fission gases created by that fuel type. Currently fast reactors, including advanced burner reactors (ABR's), are not economically feasible due to the high cost of operating the reactors and of reprocessing the fuel. However, if the fuel could be very precisely placed, such as within a freeze-cast scaffold, this could increase fuel performance and result in a valid design with a much lower cost per megawatt. In addition to competitive costs, freeze-cast fuel would also allow for selective breeding or burning of actinides within specific locations in fast reactors. For example, fast flux peak locations could be utilized on a minute scale to target specific actinides for transmutation. Freeze-cast fuel is extremely flexible and has great potential in a variety of applications. This paper performs initial modeling of freeze-cast fuel, with the generic fast reactor parameters for this model based on EBR-II. The core has an assumed power of 62.5 MWt. The neutronics code used was Monte Carlo N-Particle (MCNP5) transport code. Uniform pore sizes were used in increments of 100 microns. Two different freeze-cast scaffold materials were used: ceramic (MgO-ZrO{sub 2}) and steel (SS316L). Separate models were needed for each material because the freeze-cast ceramic and metal scaffolds have different structural characteristics and overall porosities. Basic criticality results were compiled for the various models. Preliminary results show that criticality is achievable with freeze-cast fuel pins despite the significant amount of inert fuel matrix. Freeze casting is a promising method to achieve very precise fuel placement within fuel pins. (authors)

Lang, A. L.; Yablinsky, C. A.; Allen, T. R. [Dept. of Engineering Physics, Univ. of Wisconsin Madison, 1500 Engineering Drive, Madison, WI 53711 (United States); Burger, J.; Hunger, P. M.; Wegst, U. G. K. [Thayer School of Engineering, Dartmouth College, 8000 Cummings Hall, Hanover, NH 03755 (United States)

2012-07-01T23:59:59.000Z

457

Distribution Category:  

Office of Legacy Management (LM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartment ofDepartment ofof EnergyYou areDowntownRockyDeparttient,of Energy- Distribution

458

Characterization of the TRIGA Mark II reactor full-power steady state  

E-Print Network [OSTI]

In this work, the characterization of the full-power steady state of the TRIGA Mark II nuclear reactor of the University of Pavia is performed by coupling Monte Carlo (MC) simulation for neutronics with "Multiphysics" model for thermal-hydraulics. Neutronic analyses have been performed starting from a MC model of the entire reactor system, based on the MCNP5 code, that was already validated in fresh fuel and zero-power configuration (in which thermal effects are negligible) using the available experimental data as benchmark. In order to describe the full-power reactor configuration, the temperature distribution in the core is necessary. To evaluate it, a thermal-hydraulic model has been developed, using the power distribution results from MC simulation as input. The thermal-hydraulic model is focused on the core active region and takes into account sub-cooled boiling effects present at full reactor power. The obtained temperature distribution is then introduced in the MC model and a benchmark analysis is carr...

Cammi, Antonio; Chiesa, Davide; Clemenza, Massimiliano; Pozzi, Stefano; Previtali, Ezio; Sisti, Monica; Magrotti, Giovanni; Prata, Michele; Salvini, Andrea

2015-01-01T23:59:59.000Z

459

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

SciTech Connect (OSTI)

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

460

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, Victor T. (Idaho Falls, ID)

1993-01-01T23:59:59.000Z

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.


461

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

462

APPLICATION OF DATA ANALYSIS TECHNIQUES TO NUCLEAR REACTOR  

E-Print Network [OSTI]

1 APPLICATION OF DATA ANALYSIS TECHNIQUES TO NUCLEAR REACTOR SYSTEMS CODE ACCURACY ASSESSMENT) has been developed by the authors to provide quantitative comparisons between nuclear reactor systems. 1. INTRODUCTION In recent years, the commercial nuclear reactor industry has focused significant

Kunz, Robert Francis

463

STATEMENT OF CONSIDERATIONS Advance Test Reactor Class Waiver  

Broader source: Energy.gov (indexed) [DOE]

Advance Test Reactor Class Waiver W(C)-2008-004 The Advanced Test Reactor (A TR) is a pressurized water test reactor at the Idaho National Laboratory (INL) that operates at low...

464

22.312 Engineering of Nuclear Reactors, Fall 2002  

E-Print Network [OSTI]

Engineering principles of nuclear reactors, emphasizing power reactors. Power plant thermodynamics, reactor heat generation and removal (single-phase as well as two-phase coolant flow and heat transfer), and structural ...

Todreas, Neil E.

465

22.312 Engineering of Nuclear Reactors, Fall 2004  

E-Print Network [OSTI]

Engineering principles of nuclear reactors, emphasizing power reactors. Power plant thermodynamics, reactor heat generation and removal (single-phase as well as two-phase coolant flow and heat transfer), and structural ...

Buongiorno, Jacopo, 1971-

466

argentine reactor ra-6: Topics by E-print Network  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

materials is much smaller in a fast reactor than in a thermal reactor. Consequently, metals that in a thermal reactor would severely impair neutron economy are acceptable in a...

467

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

468

James P. Mosquera Director, Reactor Plant Components  

E-Print Network [OSTI]

of the application of nuclear reactor power to capital ships of the U.S. Navy, and other assigned projects. Mr for steam generator technology (within the Nuclear Components Division); and power plant systems engineer working for the U.S. Naval Nuclear Propulsion Program (a.k.a. Naval Reactors). This program is a joint

469

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

470

Topaz-II reactor control unit development  

SciTech Connect (OSTI)

The development for a new digital reactor control unit for the Topaz-II reactor is described. The unit is expected to provide the means for automated control during a possible Topaz flight experiment. The breadboard design and development is discussed.

Wyant, F.J.; Jensen, D.; Logothetis, J.

1994-12-31T23:59:59.000Z

471

University of Virginia Reactor Facility Decommissioning Results  

SciTech Connect (OSTI)

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

472

Hydrogasification reactor and method of operating same  

SciTech Connect (OSTI)

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

473

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

474

FISSION REACTORS KEYWORDS: core-barrel vibra-  

E-Print Network [OSTI]

FISSION REACTORS KEYWORDS: core-barrel vibra- tions, in-core neutron noise, shell- mode vibrations CALCULATION OF THE NEUTRON NOISE INDUCED BY SHELL-MODE CORE-BARREL VIBRATIONS IN A 1-D, TWO-GROUP, TWO-REGION SLAB REACTOR MODEL CARL SUNDE,* CHRISTOPHE DEMAZI�RE, and IMRE PÁZSIT Chalmers University of Technology

Demazière, Christophe

475

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

476

Chemical Reactor Models of Digestion Modulation  

E-Print Network [OSTI]

Chemical Reactor Models of Digestion Modulation William Wolesensky & J. David Logan Department give an overview of the application of chemical reactor theory to models of digestion processes and indicate how those models extend to eco-physiological questions of modulation of digestion and feeding

Logan, David

477

Reactor monitoring with Neutrinos Michel Cribier  

E-Print Network [OSTI]

it the automatic and non intrusive monitoring of nuclear reactor by its antineutrino signal could be very valuable, but also book keeping of the fuel element composition before and after their use in the nuclear powerReactor monitoring with Neutrinos Michel Cribier Astroparticule & Cosmologie 10, rue Alice Domon et

Paris-Sud XI, Université de

478

Selective purge for hydrogenation reactor recycle loop  

SciTech Connect (OSTI)

Processes and apparatus for providing improved contaminant removal and hydrogen recovery in hydrogenation reactors, particularly in refineries and petrochemical plants. The improved contaminant removal is achieved by selective purging, by passing gases in the hydrogenation reactor recycle loop or purge stream across membranes selective in favor of the contaminant over hydrogen.

Baker, Richard W. (Palo Alto, CA); Lokhandwala, Kaaeid A. (Union City, CA)

2001-01-01T23:59:59.000Z

479

High Temperature Gas Reactors The Next Generation ?  

E-Print Network [OSTI]

-Proof Advanced Reactor and Gas Turbine #12;Flow through Power Conversion Vessel 8 #12;9 TRISO Fuel Particle1 High Temperature Gas Reactors The Next Generation ? Professor Andrew C Kadak Massachusetts of Brayton vs. Rankine Cycle · High Temperature Helium Gas (900 C) · Direct or Indirect Cycle · Originally

480

High Temperature Gas Reactors Briefing to  

E-Print Network [OSTI]

Meltdown-Proof Advanced Reactor and Gas Turbine #12;TRISO Fuel Particle -- "Microsphere" · 0.9mm diameter · Utilizes gas turbine technology · Lower Power Density · Less Complicated Design (No ECCS) #12;AdvantagesHigh Temperature Gas Reactors Briefing to by Andrew C. Kadak, Ph.D. Professor of the Practice

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.


481

Special-purpose materials for magnetically confined fusion reactors. Third annual progress report  

SciTech Connect (OSTI)

The scope of Special Purpose Materials covers fusion reactor materials problems other than the first-wall and blanket structural materials, which are under the purview of the ADIP, DAFS, and PMI task groups. Components that are considered as special purpose materials include breeding materials, coolants, neutron multipliers, barriers for tritium control, materials for compression and OH coils and waveguides, graphite and SiC, heat-sink materials, ceramics, and materials for high-field (>10-T) superconducting magnets. It is recognized that there will be numerous materials problems that will arise during the design and construction of large magnetic-fusion energy devices such as the Engineering Test Facility (ETF) and Demonstration Reactor (DEMO). Most of these problems will be specific to a particular design or project and are the responsibility of the project, not the Materials and Radiation Effects Branch. Consequently, the Task Group on Special Purpose Materials has limited its concern to crucial and generic materials problems that must be resolved if magnetic-fusion devices are to succeed. Important areas specifically excluded include low-field (8-T) superconductors, fuels for hybrids, and materials for inertial-confinement devices. These areas may be added in the future when funding permits.

Not Available

1981-11-01T23:59:59.000Z

482

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

SciTech Connect (OSTI)

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

483

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

SciTech Connect (OSTI)

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

484

State space modeling of reactor core in a pressurized water reactor  

SciTech Connect (OSTI)

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

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

2014-07-10T23:59:59.000Z

485

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

486

Fluorescence-based characterization of genetically encoded peptides that fold in live cells: progress towards a generic hairpin scaffold  

E-Print Network [OSTI]

bond.7, 8 Another approach is to turn to alternative protein structures that lack disulfide bonds, University of Alberta, Edmonton, Alberta, Canada T6G 2G2 ABSTRACT Binding proteins suitable for expression of validated non-immunoglobulin binding proteins, we have undertaken the development of a generic protein

Campbell, Robert E.

487

GENERIC FEEDBACK: 2014 MATH 34032 EXAM (1) (a) Not very many people managed to find the correct adjoint boundary condition  

E-Print Network [OSTI]

GENERIC FEEDBACK: 2014 MATH 34032 EXAM (1) (a) Not very many people managed to find the correct. (2) (a) Many people tried to introduce both c1 and c2. Since the degenerate kernel in this case has know what the adjoint problem is. Not very many people seemed to realise this. (c) Very few people

Sidorov, Nikita

488

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

489

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

490

Operational control of boiling water reactor stability  

SciTech Connect (OSTI)

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

491

Scanning tunneling microscope assembly, reactor, and system  

DOE Patents [OSTI]

An embodiment of a scanning tunneling microscope (STM) reactor includes a pressure vessel, an STM assembly, and three spring coupling objects. The pressure vessel includes a sealable port, an interior, and an exterior. An embodiment of an STM system includes a vacuum chamber, an STM reactor, and three springs. The three springs couple the STM reactor to the vacuum chamber and are operable to suspend the scanning tunneling microscope reactor within the interior of the vacuum chamber during operation of the STM reactor. An embodiment of an STM assembly includes a coarse displacement arrangement, a piezoelectric fine displacement scanning tube coupled to the coarse displacement arrangement, and a receiver. The piezoelectric fine displacement scanning tube is coupled to the coarse displacement arrangement. The receiver is coupled to the piezoelectric scanning tube and is operable to receive a tip holder, and the tip holder is operable to receive a tip.

Tao, Feng; Salmeron, Miquel; Somorjai, Gabor A

2014-11-18T23:59:59.000Z

492

Fast-acting nuclear reactor control device  

SciTech Connect (OSTI)

This invention consists of a fast-acting nuclear reactor control device for moving and positioning a safety control rod to desired elevations 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 a 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, O.M.; West, P.B.

1992-12-31T23:59:59.000Z

493

TREAT Reactor Control and Protection System  

SciTech Connect (OSTI)

The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS). The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab.

Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.; Lenkszus, F.R.; McDowell, W.P.

1985-01-01T23:59:59.000Z

494

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