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1

Generic small modular reactor plant design.  

SciTech Connect

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

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

2012-12-01T23:59:59.000Z

2

Performance and Safety Analysis of a Generic Small Modular Reactor  

E-Print Network (OSTI)

The high and ever growing demand for electricity coupled with environmental concerns and a worldwide desire to shed petroleum dependence, all point to a shift to utilization of renewable sources of energy. The under developed nature of truly 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 role with a number of Small Modular Reactors (SMRs) currently under development in the U.S. In this work, an SMR model of the Integral Pressurized Water Reactor (IPWR) type is created, analyzed and optimized to meet the publically available performance criteria of the mPower SMR from B&W. The Monte Carlo codes MCNP5/MCNPX are used to model the core. Fuel enrichment, core inventory, core size are all variables optimized to meet the set goals of core lifetime and fuel utilization (burnup). Vital core behavior characteristics such as delayed neutron fraction and reactivity coefficients are calculated and shown to be typical of larger PWR systems, which is necessary to ensure the inherent safety and to achieve rapid deployment of the reactor by leveraging the vast body of operational experience amassed with the larger commercial PWRs. Inherent safety of the model is analyzed with the results of an analytical single channel analysis showing promising behavior in terms of axial and radial fuel element temperature distributions, the critical heat flux, and the departure from nucleate boiling ratio. The new fleet of proposed SMRs is intended to have increased proliferation resistance (PR) compared to the existing fleet of operating commercial PWRs. To quantify this PR gain, a PR analysis is performed using the Proliferation Resistance Analysis and Evaluation Tool for Observed Risk (PRAETOR) code developed by the Nuclear Science and Security Policy Institute at Texas A&M University. The PRAETOR code uses multi-attribute utility analysis to combine 63 factors affecting the PR value of a facility into a single metric which is easily comparable. The analysis compared hypothetical spent fuel storage facilities for the SMR model spent fuel assembly and one for spent fuel from a Westinghouse AP1000. The results showed that from a fuel material standpoint, the SMR and AP1000 had effectively the same PR value. Unable to analyze security systems and methods employed at specific nuclear power plant sites, it is premature to conclude that the SMR plants will not indeed show increased PR as intended.

Kitcher, Evans Damenortey, 1987-

2012-12-01T23:59:59.000Z

3

Comparison of safety parameters and transient behavior of a generic 10 MW reactor with HEU and LEU fuels  

SciTech Connect

Key safety parameters are compared for equilibrium cores of the IAEA generic 10 MW reactor with HEU and LEU fuels. These parameters include kinetics parameters, reactivity feedback coefficients, control rod worths, power peaking factors, and shutdown margins. Reactivity insertion and loss-of-flow transients are compared. Results indicate that HEU and LEU cores will behave in a very similar manner.

Matos, J.E.; Freese, K.E.; Woodruff, W.L.

1983-01-01T23:59:59.000Z

4

Generic Requirements Specification for Upgrading the Safety-Related Reactor Trip and Engineered Safety Features Actuation Systems in Westinghouse PWR Nuclear Power Plants  

Science Conference Proceedings (OSTI)

To address obsolescence concerns, a generic requirements specification for digital upgrades to existing reactor trip systems and engineered safety features actuation systems in a Westinghouse pressurized water reactor (PWR) was developed. System requirements are based on a 4-loop PWR with a solid-state protection system since this typifies the most advanced capability level. However, the specification is applicable to relay-based 2- and 3-loop plants where some or all of the advances in the newest solid-...

2001-10-19T23:59:59.000Z

5

LIMITED POWER BURSTS IN DISTRIBUTED MODELS OF NUCLEAR REACTORS  

E-Print Network (OSTI)

of a nuclear reactor with feedback," in: Applied Problems in the Theory of Oscillations [in RussianLIMITED POWER BURSTS IN DISTRIBUTED MODELS OF NUCLEAR REACTORS M. V. Bazhenov and E. F. Sabaev UDC of Nuclear Reactors [in Russian], l~nergoatomizdat, Moscow (1990). F. R. Gantmakher and V. A. Yakubovich

Bazhenov, Maxim

6

Generic Assessment for Optimized Reactor Coolant System Hydrogen of a Four-loop Westinghouse Pressurized Water Reactor  

Science Conference Proceedings (OSTI)

The Chemistry, Fuel Reliability, and Material Reliability Programs at the Electric Power Research Institute (EPRI) have developed a comprehensive elevated reactor coolant system (RCS) hydrogen program that is focused on qualification of plant operation with dissolved hydrogen concentration in the RCS greater than 50 standard cubic centimeters per kilogram (scc/kg) (1.38 in.3/lbm), up to 60 scc/kg (1.66 in.3/lbm), to mitigate primary water stress corrosion cracking (PWSCC) in nickel-based alloys. Currentl...

2011-12-23T23:59:59.000Z

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

Computational Engineering on the Grid: Crafting a Distributed Virtual Reactor  

Science Conference Proceedings (OSTI)

This paper reports on our research into supporting collaborative distributed applications on the Grid. Our case study application, a Virtual Reactor problem solving environment, was built for simulation of industrially important technology of plasma ...

V. V. Krzhizhanovskaya; V. V. Korkhov; A. Tirado-Ramos; D. J. Groen; I. V. Shoshmina; I. A. Valuev; I. V. Morozov; N. V. Malyshkin; Y. E. Gorbachev; P. M. A. Sloot

2006-12-01T23:59:59.000Z

10

Brookhaven Reactor Experiment Control Facility, a distributed function computer network  

SciTech Connect

A computer network for real-time data acquisition, monitoring and control of a series of experiments at the Brookhaven High Flux Beam Reactor has been developed and has been set into routine operation. This reactor experiment control facility presently services nine neutron spectrometers and one x-ray diffractometer. Several additional experiment connections are in progress. The architecture of the facility is based on a distributed function network concept. A statement of implementation and results is presented. (auth)

Dimmler, D.G.; Greenlaw, N.; Kelley, M.A.; Potter, D.W.; Rankowitz, S.; Stubblefield, F.W.

1975-11-01T23:59:59.000Z

11

Regulatory analysis for the resolution of Generic Safety Issue 105: Interfacing system loss-of-coolant accident in light-water reactors  

SciTech Connect

An interfacing systems loss of coolant accident (ISLOCA) involves failure or improper operation of pressure isolation valves (PIVs) that compose the boundary between the reactor coolant system and low-pressure rated systems. Some ISLOCAs can bypass containment and result in direct release of fission products to the environment. A cost/benefit evaluation, using three PWR analyses, calculated the benefit of two potential modifications to the plants. Alternative 1 is improved plant operations to optimize the operator`s performance and reduce human error probabilities. Alternative 2 adds pressure sensing devices, cabling, and instrumentation between two PIVs to provide operators with continuous monitoring of the first PIV. These two alternatives were evaluated for the base case plants (Case 1) and for each plant, assuming the plants had a particular auxiliary building design in which severe flooding would be a problem if an ISLOCA occurred. The auxiliary building design (Case 2) was selected from a survey that revealed a number of designs with features that provided less than optimal resistance to ECCS equipment loss caused by a ISLOCA-induced environment. The results were judged not to provide sufficient basis for generic requirements. It was concluded that the most viable course of action to resolve Generic Issue 105 is licensee participation in individual plant examinations (IPEs).

Not Available

1993-07-01T23:59:59.000Z

12

Distribution of characteristics of LWR [light water reactor] spent fuel  

SciTech Connect

The purpose of this report is to develop a collective description of the entire spent fuel inventory in terms of various fuel properties relevant to Approved Testing Materials (ATMs) using information available from the Characteristics Data Base (CBD), which is sponsored by the US Department of Energy`s (DOE`s) Office of Civilian Radioactive Waste Management. A number of light-water reactor (LWR) characteristics were analyzed including assembly class representation, fuel burnup, enrichment, fuel fabrication data, defective fuel quantities, and, at PNL`s specific request, linear heat generation rate (LHGR) and the utilization of burnable poisons. A quantitative relationships was developed between burnup and enrichment for BWRs and PWRs. The relationship shows that the existing BWR ATM is near the center of the burnup-enrichment distribution, while the four PWR ATMs bracket the center of the burnup range but are on the low side of the enrichment range. Fuel fabrication data are based on vendor specifications for new fuel. Defective fuel distributions were analyzed in terms of assembly class and vendor design. LHGR values were calculated from utility data on burnup and effective full-power days; these calculations incorporate some unavoidable assumptions which may compromise the value of the results. Only a limited amount of data are available on burnable poisons at this time. Based on this distribution study, suggestions for additional ATMs are made. These are based on the class and design concepts and include BWR/2,3 barrier fuel, and the WE 17 {times} 17 class with integral burnable poison. Both should be at relatively high burnups. 16 refs., 5 figs., 15 tabs.

Reich, W.J.; Notz, K.J. [Oak Ridge National Lab., TN (USA); Moore, R.S. [Automated Sciences Group, Inc., Oak Ridge, TN (USA)

1991-01-01T23:59:59.000Z

13

Effect of coal rank and process conditions on temperature distribution in a liquefaction reactor  

SciTech Connect

The temperature distribution in a liquefaction reactor in the integrated TSL process is studied. The effects of gas and slurry superficial velocities, process solvent characteristics, reactor length, and catalyst sulfiding agent on the exotherm and temperature difference in the reactor are studied. A substantial temperature difference is observed with subbituminous coal as compared with bituminous coal, at comparable reactor conditions. Some of the factors that are believed to have contributed to the large exotherm and temperature difference in the reactor are slow kinetics and high reaction heat for subbituminous coal conversion and pyrrhotite catalysis.

Nalitham, R.V.; Moniz, M.

1986-04-01T23:59:59.000Z

14

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

Science Conference Proceedings (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

15

Final generic environmental statement on the use of recycle plutonium in mixed oxide fuel in light water cooled reactors. Volume 5. Public comments and Nuclear Regulatory Commission responses  

SciTech Connect

Copies of 69 letters are presented commenting on the Draft Generic Environmental Statement (GESMO) WASH-1327 and the NRC's responses to the comments received from Federal, State and local agencies; environmental and public interest groups, members of the academic and industrial communities, and individual citizens. An index to these letters indicating the number assigned to each letter, the author, and organization represented, is provided in the Table of Contents.

1976-08-01T23:59:59.000Z

16

An optimality condition for the assembly distribution in a nuclear reactor Laurent Thevenot  

E-Print Network (OSTI)

We give here an optimality condition for the optimization problem of the assembly distribution in a nuclear reactor, by using the homogenization method. In this paper the reactivity of the reactor core is measured by the critical eigenvalue for both continuous and multigroup neutron transport modelings. In particular, we extend the spectral theory of the critical eigenvalue and prove the dierentiability of this latter with respect to the design parameter, the con guration of the fuels. MS Classi cation: 49K20, 49J50, 35P05, 35F15. Key words: Neutron Transport, Nuclear Reactor, Optimization, Optimality Condition, Homogenization, Critical Eigenvalue, Eigenvalue Derivative.

Departement Of Applied; Laurent Thevenot

2003-01-01T23:59:59.000Z

17

TEMPEST code simulations of hydrogen distribution in reactor containment structures. Final report  

Science Conference Proceedings (OSTI)

The mass transport version of the TEMPEST computer code was used to simulate hydrogen distribution in geometric configurations relevant to reactor containment structures. Predicted results of Battelle-Frankfurt hydrogen distribution tests 1 to 6, and 12 are presented. Agreement between predictions and experimental data is good. Best agreement is obtained using the k-epsilon turbulence model in TEMPEST in flow cases where turbulent diffusion and stable stratification are dominant mechanisms affecting transport. The code's general analysis capabilities are summarized.

Trent, D.S.; Eyler, L.L.

1985-03-01T23:59:59.000Z

18

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

SciTech Connect

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

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

2013-06-01T23:59:59.000Z

19

Steam Generator Management Program: Pressurized Water Reactor Generic Tube Degradation Predictions: Recirculating Steam Generators with Alloy 600TT, Alloy 690TT, and Alloy 800NG Tubing  

Science Conference Proceedings (OSTI)

Mill-annealed Alloy 600 heat transfer tubing in pressurized water reactor (PWR) steam generators (SGs) has experienced numerous modes of degradation. This report describes predictive models for determining expected tube degradation in recirculating steam generators with Alloy 600TT, Alloy 690TT, and Alloy 800NG tubing. Predictions are based on operating experience with similar designs and use improvement factors to characterize benefits resulting from SG design and material ...

2013-12-17T23:59:59.000Z

20

Experimental power density distribution benchmark in the TRIGA Mark II reactor  

Science Conference Proceedings (OSTI)

In order to improve the power calibration process and to benchmark the existing computational model of the TRIGA Mark II reactor at the Josef Stefan Inst. (JSI), a bilateral project was started as part of the agreement between the French Commissariat a l'energie atomique et aux energies alternatives (CEA) and the Ministry of higher education, science and technology of Slovenia. One of the objectives of the project was to analyze and improve the power calibration process of the JSI TRIGA reactor (procedural improvement and uncertainty reduction) by using absolutely calibrated CEA fission chambers (FCs). This is one of the few available power density distribution benchmarks for testing not only the fission rate distribution but also the absolute values of the fission rates. Our preliminary calculations indicate that the total experimental uncertainty of the measured reaction rate is sufficiently low that the experiments could be considered as benchmark experiments. (authors)

Snoj, L.; Stancar, Z.; Radulovic, V.; Podvratnik, M.; Zerovnik, G.; Trkov, A. [Josef Stefan Inst., Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Barbot, L.; Domergue, C.; Destouches, C. [CEA DEN, DER, Instrumentation Sensors and Dosimetry laboratory Cadarache, F-13108 Saint-Paul-Lez-Durance (France)

2012-07-01T23:59:59.000Z

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.


21

Thermionic nuclear reactor with internal heat distribution and multiple duct cooling  

DOE Patents (OSTI)

A Thermionic Nuclear Reactor is described having multiple ribbon-like coolant ducts passing through the core, intertwined among the thermionic fuel elements to provide independent cooling paths. Heat pipes are disposed in the core between and adjacent to the thermionic fuel elements and the ribbon ducting, for the purpose of more uniformly distributing the heat of fission among the thermionic fuel elements and the ducts.

Fisher, C.R.; Perry, L.W. Jr.

1975-11-01T23:59:59.000Z

22

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

Science Conference Proceedings (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

23

A Generic Dynamic Emulator  

E-Print Network (OSTI)

In applied sciences, we often deal with deterministic simulation models that are too slow for simulation-intensive tasks such as calibration or real-time control. In this paper, an emulator for a generic dynamic model, given by a system of ordinary non-linear differential equations, is developed. The non-linear differential equations are linearized and Gaussian white noise is added to account for the non-linearities. The resulting linear stochastic system is conditioned on a set of solutions of the non-linear equations that have been calculated prior to the emulation. A path-integral approach is used to derive the Gaussian distribution of the emulated solution. The solution reveals that most of the computational burden can be shifted to the conditioning phase of the emulator and the complexity of the actual emulation step only scales like $\\mathcal O(Nnm^2)$, where $N$ is the number of time-points at which the solution is to be emulated, $n$ the number of solutions the emulator is conditioned on and $m$ the n...

Albert, Carlo

2011-01-01T23:59:59.000Z

24

Verifying generics and delegates  

Science Conference Proceedings (OSTI)

Recently, object-oriented languages, such as C#, have been extended with language features prevalent in most functional languages: parametric polymorphism and higher-order functions. In the OO world these are called generics and delegates, respectively. ...

Kasper Svendsen; Lars Birkedal; Matthew Parkinson

2010-06-01T23:59:59.000Z

25

Radionuclide characterization at US commercial light-water reactors for decommissioning assessment: Distributions, inventories, and waste disposal considerations  

SciTech Connect

A continuing research program, conducted by Pacific Northwest Laboratory for the US Nuclear Regulatory Commission, characterizing radionuclide concentrations associated with US light-water reactors has been conducted for more than a decade. The research initially focused upon sampling and analytical measurements for the purpose of establishing radionuclide distributions and inventories for decommissioning assessment, since very little empirical data existed. The initial phase of the research program examined radionuclide concentrations and distributions external to the reactor vessel at seven US light water reactors. Later stages of the research program have examined the radionuclide distributions in the highly radioactive reactor internals and fuel assembly. Most recently, the research program is determining radionuclide concentrations in these highly radioactive components and comparing empirical results with those derived from the several nonempirical methodologies employed to estimate radionuclide inventories for disposal classification. The results of the research program to date are summarized, and their implications and significance for the decommissioning process are noted.

Abel, K.H.; Robertson, D.E.; Thomas, C.W.

1992-09-01T23:59:59.000Z

26

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); Few, Douglas A. (Idaho Falls, ID)

2010-09-21T23:59:59.000Z

27

SPATIAL DISTRIBUTION OF FISSION-PRODUCT GAMMA-RAY ENERGY DEPOS WATER REACTOR FUEL ELEMENTS, VOLUME 2  

Science Conference Proceedings (OSTI)

Reports studies undertaken to produce a precise and readily interpretable description of the distribution of absorbed energy in an LWR fuel element. Data are useful in determining the spatial distribution of absorption of fission-product gamma-ray energy around regions of high local power density following the shutdown of a nuclear reactor.

1978-04-01T23:59:59.000Z

28

Generic Deep Geologic Disposal Safety Case | Department of Energy  

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

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

29

REACTOR  

DOE Patents (OSTI)

A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

Roman, W.G.

1961-06-27T23:59:59.000Z

30

Generic selections of subexpressions  

Science Conference Proceedings (OSTI)

Tools for computer languages need position information: compilers for providing better error messages, structure editors for mapping between structural and textual views, and debuggers for navigating through a term, for instance. Manually adding position ... Keywords: datatype-generic programming, haskell, selections

Martijn Van Steenbergen; José Pedro Magalhães; Johan Jeuring

2010-09-01T23:59:59.000Z

31

Development of Generic Background  

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

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

32

Generic safety documentation model  

SciTech Connect

This document is intended to be a resource for preparers of safety documentation for Sandia National Laboratories, New Mexico facilities. It provides standardized discussions of some topics that are generic to most, if not all, Sandia/NM facilities safety documents. The material provides a ``core`` upon which to develop facility-specific safety documentation. The use of the information in this document will reduce the cost of safety document preparation and improve consistency of information.

Mahn, J.A.

1994-04-01T23:59:59.000Z

33

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

34

Type inference for generic Haskell  

E-Print Network (OSTI)

Abstract. The more expressive a type system, the more type information has to be provided in a program. Having to provide a type is sometimes a pain, but lacking expressivity is often even worse. There is a continuous struggle between expressivity and (type-)verbosity. However, even very expressive type systems allow type inference for parts of a program. Generic Haskell is an extension of Haskell that supports defining generic functions. Generic Haskell assumes that the type of a generic function is explicitly specified. This is often no problem, but sometimes it is rather painful to have to specify a type – in particular for generic functions with many dependencies – and sometimes the specified type can be generalized. In this paper, we identify three type inference problems specific to generic functions, and present (partial) solutions to each of them. 1

Alexey Rodriguez; Johan Jeuring; Andres Löh

2005-01-01T23:59:59.000Z

35

FHR Generic Design Criteria  

SciTech Connect

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

36

FHR Generic Design Criteria  

Science Conference Proceedings (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, George F [ORNL; Holcomb, David Eugene [ORNL; Cetiner, Mustafa Sacit [ORNL

2012-06-01T23:59:59.000Z

37

MSS generic model: a beginning  

SciTech Connect

To improve communications and understanding in the field of Mass Storage Systems (MSS), the need exists to formalize the definitions and concepts of MSS. This presentation gives MSS definitions followed by the development of a generic model for MSS.

Collins, W.; Miller, S.

1984-01-01T23:59:59.000Z

38

Generic and maximal Jordan types  

Science Conference Proceedings (OSTI)

Mar 7, 2007 ... •(G,k) we exhibit a “generic Jordan type” of M. In the very special ... specializes to the non-trivial observation that the Jordan type obtained by.

39

Hanford Generic Interim Safety Basis  

SciTech Connect

The purpose of this document is to identify WHC programs and requirements that are an integral part of the authorization basis for nuclear facilities that are generic to all WHC-managed facilities. The purpose of these programs is to implement the DOE Orders, as WHC becomes contractually obligated to implement them. The Hanford Generic ISB focuses on the institutional controls and safety requirements identified in DOE Order 5480.23, Nuclear Safety Analysis Reports.

Lavender, J.C.

1994-09-09T23:59:59.000Z

40

CALCULATION OF EFFECT OF FUEL BURNUP ON FUEL AND POISON DISTRIBUTIONS AND ON FLUX DISTRIBUTION IN THE MARINE REACTOR  

SciTech Connect

The changes in fuel and poison distributions and the changes in flux shape which accompany the consumption of fuel are studied. The technique employed is a perturbation calculation based on a one-velocity group treatment of the neutrons. The geometry is a spherical core surrounded by an infinite reflector. The programming forms for the IBM-650 which performs the computations using the Wolontis interpretative system are included. Two sample calculations were worked out using the code described and the results are plotted. (M.H.R.)

Hinman, G.

1957-01-29T23:59:59.000Z

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


41

Duke Edwardsport Generic Design Specification  

Science Conference Proceedings (OSTI)

This generic design specification (GDS) is sponsored by the CoalFleet for Tomorrow program of the Electric Power Research Institute (EPRI). The CoalFleet Integrated Gasification Combined Cycle (IGCC) Design Guidelines Experts Group produced this report, which is based on available public sources and information. The CoalFleet IGCC Design Guidelines Experts Group consists of more than 10 IGCC experts, including independent consultants and EPRI staff.

2011-01-20T23:59:59.000Z

42

Non-classical transport with angular-dependent path-length distributions. 2: Application to pebble bed reactor cores  

E-Print Network (OSTI)

We describe an analysis of neutron transport in the interior of model pebble bed reactor (PBR) cores, considering both crystal and random pebble arrangements. Monte Carlo codes were developed for (i) generating random realizations of the model PBR core, and (ii) performing neutron transport inside the crystal and random heterogeneous cores; numerical results are presented for two different choices of material parameters. These numerical results are used to investigate the anisotropic behavior of neutrons in each case and to assess the accuracy of estimates for the diffusion coefficients obtained with the diffusion approximations of different models: the atomic mix model, the Behrens correction, the Lieberoth correction, the generalized linear Boltzmann equation (GLBE), and the new GLBE with angular-dependent path-length distributions. This new theory utilizes a non-classical form of the Boltzmann equation in which the locations of the scattering centers in the system are correlated and the distance-to-collision is not exponentially distributed; this leads to an anisotropic diffusion equation. We show that the results predicted using the new GLBE theory are extremely accurate, correctly identifying the anisotropic diffusion in each case and greatly outperforming the other models for the case of random systems.

Richard Vasques; Edward W. Larsen

2013-09-24T23:59:59.000Z

43

A Generic Framework for Web-Based Intelligent Decision Support Systems  

Science Conference Proceedings (OSTI)

Web-based Intelligent Systems Environment (WISE) offers generic, modular, flexible and scalable system solutions for information retrieval, extraction, fusion, knowledge discovery and intelligent decision support, using heterogeneous, distributed data ...

Vladimir Simeunovic; Jelena Jovanovic; Milan Saric; Sanja Vranes

2002-12-01T23:59:59.000Z

44

Generation -IV Reactor Concepts  

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

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

45

Descriptive Model of Generic WAMS  

Science Conference Proceedings (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

46

Numerical simulation of flow distribution for pebble bed high temperature gas cooled reactors  

E-Print Network (OSTI)

The premise of the work presented here is to use a common analytical tool, Computational Fluid dynamics (CFD), along with a difference turbulence models. Eddy viscosity models as well as state-of-the-art Large Eddy Simulation (LES) were used to study the flow past bluff bodies. A suitable CFD code (CFX5.6b) was selected and implemented. Simulation of turbulent transport for the gas through the gaps of the randomly distributed spherical fuel elements (pebbles) was performed. Although there are a number of numerical studies () on flows around spherical bodies, none of them use the necessary turbulence models that are required to simulate flow where strong separation exists. With the development of high performance computers built for applications that require high CPU time and memory; numerical simulation becomes one of the more effective approaches for such investigations and LES type of turbulence models can be used more effectively. Since there are objects that are touching each other in the present study, a special approach was applied at the stage of building computational domain. This is supposed to be a considerable improvement for CFD applications. Zero thickness was achieved between the pebbles in which fission reaction takes place. Since there is a strong pressure gradient as a result of high Reynolds Number on the computational domain, which strongly affects the boundary layer behavior, heat transfer in both laminar and turbulent flows varies noticeably. Therefore, noncircular curved flows as in the pebble-bed situatio n, in detailed local sense, is interesting to be investigated. Since a compromise is needed between accuracy of results and time/cost of effort in acquiring the results numerically, selection of turbulence model should be done carefully. Resolving all the scales of a turbulent flow is too costly, while employing highly empirical turbulence models to complex problems could give inaccurate simulation results. The Large Eddy Simulation (LES) method would achieve the requirements to obtain a reasonable result. In LES, the large scales in the flow are solved and the small scales are modeled. Eddy viscosity and Reynolds stress models were also be used to investigate the applicability of these models for this kind of flow past bluff bodies at high Re numbers.

Yesilyurt, Gokhan

2006-05-01T23:59:59.000Z

47

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

SciTech Connect

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

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

1996-12-01T23:59:59.000Z

48

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

49

Bayesian Generic Priors for Causal Learning  

E-Print Network (OSTI)

informal argumenta- tion: A Bayesian approach to reasoningcausal learning using Bayesian generic priors on generativeHolyoak, K. J. (2007). Bayesian models of judgments of

Lu, Hongjing; Yuille, Alan L.; Liljeholm, Mimi; Cheng, Patricia W.; Holyoak, Keith J.

2008-01-01T23:59:59.000Z

50

Bayesian Generic Priors for Causal Learning  

E-Print Network (OSTI)

informal argumenta- tion: A Bayesian approach to reasoningcausal learning using Bayesian generic priors on generativeHolyoak, K. J. (2007). Bayesian models of judgments of

Hongjing Lu; Alan L. Yuille; Mimi Liljeholm; Patricia W. Cheng; Keith J. Holyoak

2011-01-01T23:59:59.000Z

51

Implementing a generic component-based framework for telecontrol applications  

Science Conference Proceedings (OSTI)

The rapid growth of telecontrol systems is one of the major trends in today's network-oriented community. The implementation of generic frameworks, consisting of reusable components that can form the basis for the development of such systems, is a necessity. ... Keywords: component-based systems, distributed control system (DCS), reusable software components, patterns and frameworks, supervisory control and data acquisition system (SCADA), telecontrol systems

Avraam N. Chimaris; George A. Papadopoulos

2007-08-01T23:59:59.000Z

52

Measurement of the axial distribution of radioactivity in the auxiliary charcoal bed of the Molten Salt Reactor Experiment at ORNL  

SciTech Connect

The Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory commenced operation in 1964 and was shut down in 1969. It was fueled with {sup 233}UF{sub 4} in a carrier salt of LiF-BeF{sub 2}-ZrF{sub 4}, and it operated at 1,200 F. After it was shut down, the fuel was heated annually to 200 C to recombine fluorine (with the fuel) released due to radiation-induced reactions in the fuel salt. However, a competing reaction oxidized uranium to UF{sub 6}, which was released (along with F{sub 2}) from the fuel and trapped in one of four charcoal filters in the auxiliary charcoal bed (ACB). One of the tasks for decommissioning of the MSRE requires that at least 90% of the estimated 3 kg of {sup 233}U, and radioactive decay products, in this filter be removed, and one of the proposed methods is to vacuum the charcoal above a specified axial position in the filter. This requires that the axial distribution of activity in the filter be measured in a 60 rad/h radiation field to determine where this penetration can be made. To accomplish this, the shielded detector with a pinhole collimator, and with a laser positioning capability, was remotely translated to various axial positions to accomplish these measurements. Activities in the steel screen, and various regions of the charcoal bed, are estimated, and uncertainties in these estimates are generally {lt}1%. Results from this analysis are used for continued operational decisions for decommissioning of the MSRE.

Miller, L.F.; Buckner, M.; Buchanan, M.

1999-07-01T23:59:59.000Z

53

Converting java programs to use generic libraries  

Science Conference Proceedings (OSTI)

Java 1.5 will include a type system (called JSR-14) that supports parametric polymorphism, or generic classes. This will bring many benefits to Java programmers, not least because current Java practice makes heavy use of ... Keywords: JSR-14, Java 1.5, Java 5, generic types, instantiation types, parameterized types, parametric polymorphism, raw types, type inference

Alan Donovan; Adam Kiežun; Matthew S. Tschantz; Michael D. Ernst

2004-10-01T23:59:59.000Z

54

Generic OTS, EPRI Generic Operator Training Simulator, Version 2.0  

Science Conference Proceedings (OSTI)

The EPRI Generic Operator Training Simulator, Generic OTS Version 2.0 is a PC-based training simulator that allows hands-on training for dispatchers, using a generic 29 station model. Simulation-based training has long been recognized as an important component of power system operator training. The Generic OTS, which can be run on multiple platforms, allows for realistic simulations of many power system phenomena. A generic Power and Light model can be employed, or the user has the ability to include the...

2006-12-13T23:59:59.000Z

55

Heavy Water and Graphite Reactors - Reactors designed/built by...  

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

experiments, necessary to achieve higher precision for the determination of reactor power distribution patterns, effect of non-uniform void distributions, kinetic behavior,...

56

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

57

Converting Java Programs to Use Generic Libraries  

E-Print Network (OSTI)

Java 1.5 will include a type system (called JSR-14) that supports parametric polymorphism, or generic classes. This will bring many benefits to Java programmers, not least because current Java practice makes heavy use of ...

Donovan, Alan

2004-03-30T23:59:59.000Z

58

Converting Java programs to use generic libraries  

E-Print Network (OSTI)

Java 1.5 will include a type system (called JSR-14) that supports parametric polymorphism, or generic classes. This will bring many benefits to Java programmers, not least because current Java practise makes heavy use of ...

Donovan, Alan A. A., 1976-

2004-01-01T23:59:59.000Z

59

Directions for improved fusion reactors  

SciTech Connect

Conceptual fusion reactor studies over the past 10 to 15 years have projected systems that may be too large, complex, and costly to be of commercial interest. One main direction for improved fusion reactors points towards smaller, higher-power-density approaches. First-order economic issues (i.e., unit direct cost and cost of electricity) are used to support the need for more compact fusion reactors. A generic fusion physics/engineering/costing model is used to provide a quantiative basis for these arguments for specific fusion concepts.

Krakowski, R.A.; Miller, R.L.; Delene, J.G.

1986-01-01T23:59:59.000Z

60

Spatial Distribution of Fission-Product Gamma-Ray Energy Deposition in Light Water Reactor Fuel Elements, Volume 1  

Science Conference Proceedings (OSTI)

EPRI sponsored research into spatial deposition of gamma-ray energy in the core of nuclear reactors following shutdown. This work by the University of Virginia was intended to provide a benchmark tool for use in safety analyses of postulated LOCAs. Present NRC regulations state that certain important safety margins must be demonstrated for the unlikely event of a LOCA. Variables of interest include peak cladding temperature of the fuel and the amount of H2 produced by radiolysis in the coolant. The peak ...

1978-04-01T23:59:59.000Z

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.


61

NEUTRONIC REACTOR SYSTEM  

DOE Patents (OSTI)

A reactor system incorporating a reactor of the heterogeneous boiling water type is described. The reactor is comprised essentially of a core submerged adwater in the lower half of a pressure vessel and two distribution rings connected to a source of water are disposed within the pressure vessel above the reactor core, the lower distribution ring being submerged adjacent to the uppcr end of the reactor core and the other distribution ring being located adjacent to the top of the pressure vessel. A feed-water control valve, responsive to the steam demand of the load, is provided in the feedwater line to the distribution rings and regulates the amount of feed water flowing to each distribution ring, the proportion of water flowing to the submerged distribution ring being proportional to the steam demand of the load. This invention provides an automatic means exterior to the reactor to control the reactivity of the reactor over relatively long periods of time without relying upon movement of control rods or of other moving parts within the reactor structure.

Treshow, M.

1959-02-10T23:59:59.000Z

62

A generic multi-attribute analysis system  

Science Conference Proceedings (OSTI)

This paper describes a generic decision support system based on an additive multiattribute utility model that is intended to allay many of the operational difficulties involved in the multicriteria decision-making process. The system accounts for uncertainty ... Keywords: Additive multiattribute utility function, Decision support system, Incomplete information, Monte Carlo simulation techniques

Antonio Jiménez; Sixto Ríos-Insua; Alfonso Mateos

2006-04-01T23:59:59.000Z

63

CHATR: a generic speech synthesis system  

Science Conference Proceedings (OSTI)

This paper describes a generic speech synthesis system called CHATR which is being developed at ATR. CHATR is designed in a modular way so that module parameters and even which modules are actually used may be set and selected at runtime. Although some ...

Alan W. Black; Paul Taylor

1994-08-01T23:59:59.000Z

64

A generalized theory for non-classical transport with angular-dependent path-length distributions 2: Anisotropic diffusion in model pebble bed reactor cores  

E-Print Network (OSTI)

We describe an analysis of neutron transport in the interior of model pebble bed reactor (PBR) cores, considering both crystal and random pebble arrangements. Monte Carlo codes were developed for (i) generating random realizations of the model PBR core, and (ii) performing neutron transport inside the crystal and random heterogeneous cores; numerical results are presented for two different choices of material parameters. These numerical results are used to investigate the anisotropic behavior of neutrons in each case and to assess the accuracy of estimates for the diffusion coefficients obtained with the diffusion approximations of different models: the atomic mix model, the Behrens correction, the Lieberoth correction, the generalized linear Boltzmann equation (GLBE), and the new GLBE with angular-dependent path-length distributions. This new theory utilizes a non-classical form of the Boltzmann equation in which the locations of the scattering centers in the system are correlated and the distance-to-collisi...

Vasques, Richard

2013-01-01T23:59:59.000Z

65

Distribution:  

Office of Legacy Management (LM)

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

66

SHREC 2009 - Shape Retrieval Contest on a New Generic ...  

Science Conference Proceedings (OSTI)

... SHREC 2009 - Shape Retrieval Contest on a New Generic Shape Benchmark. Call for Participation: SHREC 2009 - Shape ...

67

Evolution of the Generic Lock System at Jefferson Lab  

SciTech Connect

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

68

Nuclear Maintenance Applications Center: Reactor Recirculation Pump Seal Maintenance Guide  

Science Conference Proceedings (OSTI)

In 1993, the Electric Power Research Institute (EPRI) published Main Coolant Pump Seal Maintenance Guide (report TR-100855). That report provides guidance on maintenance and troubleshooting of main coolant pump seals but was generic in naturethe information was relevant to pressurized water reactors (PWRs) and boiling water reactors (BWRs).

2008-12-18T23:59:59.000Z

69

Generic-model based human-body modeling  

Science Conference Proceedings (OSTI)

This paper presents a generic-model based human-body modeling method which take the anatomical structure of the human body into account. The generic model contains anatomical structure of bones and muscles of the human body. For a given target skin mesh, ... Keywords: anatomically-based modeling, generic model, human body modeling

Xiaomao Wu; Lizhuang Ma; Ke-Sen Huang; Yan Gao; Zhihua Chen

2005-09-01T23:59:59.000Z

70

Design of generic coal conversion facilities: Indirect coal liquefaction, Fischer-Tropsch synthesis  

SciTech Connect

A comprehensive review of Fischer-Tropsch (F-T) technology, including fixed, fluidized, and bubble column reactors, was undertaken in order to develop an information base before initiating the design of the Fischer-Tropsch indirect liquefaction PDU as a part of the Generic Coal Conversion Facilities to be built at the Pittsburgh Energy Technology Center (PETC). The pilot plant will include a fixed bed and slurry bubble column reactor for the F-T mode of operation. The review encompasses current status of both these technologies, their key variables, catalyst development, future directions, and potential improvement areas. However, more emphasis has been placed on the slurry bubble column reactor since this route is likely to be the preferred technology for commercialization, offering process advantages and, therefore, better economics than fixed and fluidized bed approaches.

Not Available

1991-10-01T23:59:59.000Z

71

Generic sequential sampling for metamodel approximations  

SciTech Connect

Metamodels approximate complex multivariate data sets from simulations and experiments. These data sets often are not based on an explicitly defined function. The resulting metamodel represents a complex system's behavior for subsequent analysis or optimization. Often an exhaustive data search to obtain the data for the metalnodel is impossible, so an intelligent sampling strategy is necessary. While inultiple approaches have been advocated, the majority of these approaches were developed in support of a particular class of metamodel, known as a Kriging. A more generic, cotninonsense approach to this problem allows sequential sampling techniques to be applied to other types of metamodeis. This research compares recent search techniques for Kriging inetamodels with a generic, inulti-criteria approach combined with a new type of B-spline metamodel. This B-spline metamodel is competitive with prior results obtained with a Kriging metamodel. Furthermore, the results of this research highlight several important features necessary for these techniques to be extended to more complex domains.

Turner, C. J. (Cameron J.); Campbell, M. I. (Matthew I.)

2003-01-01T23:59:59.000Z

72

Orlando Gasification Project Generic Design Specification  

Science Conference Proceedings (OSTI)

Integrated Gasification Combined Cycle (IGCC) technology is now entering commercial service, so there is a need to develop specifications that encourage greater standardization in IGCC design. This Generic Design Specification (GDS), sponsored by EPRI's CoalFleet for Tomorrow program with support from 50 power companies, provides technical information from Front End Engineering Design (FEED) results submitted by Southern Company Services to DOE regarding their Orlando Gasification Project (OGP), a 270 MW...

2008-05-12T23:59:59.000Z

73

Semantic Models for Distributed Object Reflection  

Science Conference Proceedings (OSTI)

A generic formal model of distributed object reflection is proposed, that combines logical reflection with a structuring of distributed objects as nested configurations of metaobject that can control subobjects under them. The model provides mathematical ...

José Meseguer; Carolyn L. Talcott

2002-06-01T23:59:59.000Z

74

PROGRESS ON GENERIC PHASE-FIELD METHOD DEVELOPMENT  

Science Conference Proceedings (OSTI)

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

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

2012-09-26T23:59:59.000Z

75

Space reactor system and subsystem investigations: assessment of technology issues for the reactor and shield subsystem. SP-100 Program  

SciTech Connect

As part of Rockwell's effort on the SP-100 Program, preliminary assessment has been completed of current nuclear technology as it relates to candidate reactor/shield subsystems for the SP-100 Program. The scope of the assessment was confined to the nuclear package (to the reactor and shield subsystems). The nine generic reactor subsystems presented in Rockwell's Subsystem Technology Assessment Report, ESG-DOE-13398, were addressed for the assessment.

Atkins, D.F.; Lillie, A.F.

1983-06-30T23:59:59.000Z

76

Generic turbine design study. Final report  

SciTech Connect

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

1993-06-01T23:59:59.000Z

77

Nuclear Reactors  

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

Reactors Nuclear reactors created not only large amounts of plutonium needed for the weapons programs, but a variety of other interesting and useful radioisotopes. They produced...

78

Business Case Slide 4: DU Use Categories: Generic Aspects - 1  

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

DU Use Categories: Generic Aspects - 1 High-volume uses Capitalize on density and chemical identity with fuel Uses are all in radiologically regulated applications Benefit Avoid...

79

Business Case Slide 21: High-Value: Generic - 2  

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

High-Value: Generic - 2 Issues All still at research stage Have theoretical or initial experimental information that does not contravene viability Most involve use in...

80

Generic Text Summarization Using Probabilistic Latent Semantic Indexing  

E-Print Network (OSTI)

Generic Text Summarization Using Probabilistic Latent Semantic Indexing Harendra Bhandari Graduate School of Information Science Nara Institute of Science and Technology Nara 630-0192, Japan harendra

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

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

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

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

82

SHREC 2010 - Shape Retrieval Contest based on Generic 3D ...  

Science Conference Proceedings (OSTI)

... this track is to evaluate the performance of 3D shape retrieval approaches on a Generic 3D shape benchmark based on the Google 3D Warehouse. ...

83

A generic theorem prover of CSP refinement  

E-Print Network (OSTI)

Abstract. We describe a new tool called Csp-Prover which is an interactive theorem prover dedicated to refinement proofs within the process algebra Csp. It aims specifically at proofs for infinite state systems, which may also involve infinite non-determinism. Semantically, Csp-Prover supports both the theory of complete metric spaces as well as the theory of complete partial orders. Both these theories are implemented for infinite product spaces. Technically, Csp-Prover is based on the theorem prover Isabelle. It provides a deep encoding of Csp. The tool’s architecture follows a generic approach which makes it easy to adapt it for various Csp models besides those studied here: the stable failures model F and the traces model T. 1

Yoshinao Isobe; Markus Roggenbach

2005-01-01T23:59:59.000Z

84

Testing generic predictions of dark energy  

E-Print Network (OSTI)

Constraints on the expansion history of the universe from measurements of cosmological distances make predictions for large-scale structure growth. Since these predictions depend on assumptions about dark energy evolution and spatial curvature, they can be used to test general classes of dark energy models by comparing predictions for those models with direct measurements of the growth history. I present predictions from current distance measurements for the growth history of dark energy models including a cosmological constant and quintessence. Although a time-dependent dark energy equation of state significantly weakens predictions for growth from measured distances, for quintessence there is a generic limit on the growth evolution that could be used to falsify the whole class of quintessence models. Understanding the allowed range of growth for dark energy models in the context of general relativity is a crucial step for efforts to distinguish dark energy from modified gravity.

Mortonson, Michael J

2010-01-01T23:59:59.000Z

85

A generic reaction-based biogeochemical simulator  

SciTech Connect

This paper presents a generic biogeochemical simulator, BIOGEOCHEM. The simulator can read a thermodynamic database based on the EQ3/EQ6 database. It can also read user-specified equilibrium and kinetic reactions (reactions not defined in the format of that in EQ3/EQ6 database) symbolically. BIOGEOCHEM is developed with a general paradigm. It overcomes the requirement in most available reaction-based models that reactions and rate laws be specified in a limited number of canonical forms. The simulator interprets the reactions, and rate laws of virtually any type for input to the MAPLE symbolic mathematical software package. MAPLE then generates Fortran code for the analytical Jacobian matrix used in the Newton-Raphson technique, which are compiled and linked into the BIOGEOCHEM executable. With this feature, the users are exempted from recoding the simulator to accept new equilibrium expressions or kinetic rate laws. Two examples are used to demonstrate the new features of the simulator.

Fang, Yilin; Yabusaki, Steven B.; Yeh, Gour T.; C.T. Miller, M.W. Farthing, W.G. Gray, and G.F. Pinder

2004-06-17T23:59:59.000Z

86

Distribution ICategory: General Reactor Technology  

E-Print Network (OSTI)

1945 when the first atomic bomb exploded. When fear exists, there will, in a free society, be those who between Argonne National laboratory and Harvard University *A short version of this report, and on the other, predictions of catastrophe that cannot, in fact, occur. In this report we address one particular

Shlyakhter, Ilya

87

Minimizing dependencies within generic classes for faster and smaller programs  

Science Conference Proceedings (OSTI)

Generic classes can be used to improve performance by allowing compile-time polymorphism. But the applicability of compile-time polymorphism is narrower than that of runtime polymorphism, and it might bloat the object code. We advocate a programming ... Keywords: generalized hoisting, generics, scary assignments and initializations, templates

Dan Tsafrir; Robert W. Wisniewski; David F. Bacon; Bjarne Stroustrup

2009-10-01T23:59:59.000Z

88

Generic Security Proof of Quantum Key Exchange using Squeezed States  

E-Print Network (OSTI)

Recently, a Quantum Key Exchange protocol that uses squeezed states was presented by Gottesman and Preskill. In this paper we give a generic security proof for this protocol. The method used for this generic security proof is based on recent work by Christiandl, Renner and Ekert.

Karin Poels; Pim Tuyls; Berry Schoenmakers

2005-08-09T23:59:59.000Z

89

Geomaterials Research Project The Evolution of Generic Material Standards for  

E-Print Network (OSTI)

Geomaterials Research Project The Evolution of Generic Material Standards for Block Manhattan College School of Engineering Civil and Environmental Engineering Department Bronx, New York, U.S.A. May 2012 #12;ii Geomaterials Research Project The Evolution of Generic Material Standards for Block

Horvath, John S.

90

Thesis: a generic, collaborative framework for interval constraint solving  

Science Conference Proceedings (OSTI)

The paper abstracts the contents of a PhD dissertation entitled A Generic, Collaborative Framework for Interval Constraint Solving which has been recently defended. This thesis presents a generic framework for defining and solving interval constraints ... Keywords: constraint, cooperation, indexical, lattice, propagation

Antonio J. Fernández

2004-07-01T23:59:59.000Z

91

A Generic, Collaborative Framework for Interval Constraint Solving: Thesis  

Science Conference Proceedings (OSTI)

The paper abstracts the contents of a PhD dissertation entitled A Generic, Collaborative Framework for Interval Constraint Solving which has been recently defended. This thesis presents a generic framework for defining and solving interval constraints ... Keywords: Constraint, cooperation, indexical, lattice, propagation

Antonio J. Fernández

2004-08-01T23:59:59.000Z

92

Generic Disposal System Modeling, Fiscal Year 2011 Progress Report |  

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

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

93

Descriptive Model of a Generic WAMS | Department of Energy  

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

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

94

Effective Higgs vertices in the generic MSSM  

Science Conference Proceedings (OSTI)

In this article we consider chirally enhanced corrections to Higgs vertices in the most general MSSM. We include the contributions stemming from bilinear terms, from the trilinear A terms, and from their nonholomorphic analogues, the A{sup '} terms, which couple squarks to the ''wrong'' Higgs field. We perform a consistent renormalization of the Higgs vertices beyond the decoupling limit (M{sub SUSY{yields}{infinity}}), using a purely diagrammatic approach. The cancellation of the different contributions in and beyond the decoupling limit is discussed and the possible size of decoupling effects which occur if the supersymmetry particles are not much heavier than the electroweak scale are examined. In the decoupling limit we recover the results obtained in the effective-field-theory approach. For the nonholomorphic A{sup '} terms we find the well known tan{beta} enhancement in the down sector similar to the one for terms proportional to {mu}. Because of the a priori generic flavor structure of these trilinear terms large flavor-changing neutral Higgs couplings can be induced. We also discover new tan{beta} enhanced contributions involving the usual holomorphic A terms, which were not discussed before in the literature. These corrections occur only if also flavor-diagonal nonholomorphic corrections to the Higgs couplings are present. This reflects the fact that the A terms, and also the chirality-changing self-energies, are physical quantities and cannot be absorbed into renormalization constants.

Crivellin, Andreas [Albert Einstein Center for Fundamental Physics, Institute for Theoretical Physics, University of Bern, CH-3012 Bern (Switzerland)

2011-03-01T23:59:59.000Z

95

A Generic Framework for Network Forensics  

E-Print Network (OSTI)

Internet is the most powerful medium as on date, facilitating varied services to numerous users. It has also become the environment for cyber warfare where attacks of many types (financial, ideological, revenge) are being launched. The e-commerce transactions being carried out online are of major interest to cybercriminals. The Internet needs to be protected from these attacks and an appropriate response has to be generated to handle them to reduce the impact. Network forensics is the science that deals with capture, recording, and analysis of network traffic for investigative purpose and incident response. There are many tools which assist in capturing data transferred over the networks so that an attack or the malicious intent of the intrusions may be investigated. This paper presents a generic framework for network forensic analysis by specifically identifying the steps connected only to network forensics from the already proposed models for digital investigation. Each of the phases in the framework is elucidated. A comparison of the proposed model is done with the existing models for digital investigation. Research challenges in various phases of the model are approached with specific reference to network forensics.

Emmanuel S. Pilli; R. C. Joshi; Rajdeep Niyogi

2010-01-01T23:59:59.000Z

96

System Design - Lessons Learned, Generic Concepts, Characteristics...  

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

A: Approved for public release; distribution is unlimited. 20 Molten Carbonate Fuel Cell System Hot Box MCFC Stack Fuel Cell Characteristics *Manufacturer: Manufacturer: Fuel...

97

NUCLEAR REACTOR  

DOE Patents (OSTI)

A boiling-water nuclear reactor is described wherein control is effected by varying the moderator-to-fuel ratio in the reactor core. This is accomplished by providing control tubes containing a liquid control moderator in the reactor core and providing means for varying the amount of control moderatcr within the control tubes.

Treshow, M.

1961-09-01T23:59:59.000Z

98

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A reactor in which at least a portion of the moderator is in the form of movable refractory balls is described. In addition to their moderating capacity, these balls may serve as carriers for fissionable material or fertile material, or may serve in a coolant capacity to remove heat from the reactor. A pneumatic system is used to circulate the balls through the reactor.

Daniels, F.

1959-10-27T23:59:59.000Z

99

Materials Challenges in Next Generation Nuclear Reactors  

Science Conference Proceedings (OSTI)

Materials under active consideration for use in different reactor components ... A Theoretical Model of Corrosion Rate Distribution in Liquid LBE Flow Loop at ...

100

CONVECTION REACTOR  

DOE Patents (OSTI)

An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

Hammond, R.P.; King, L.D.P.

1960-03-22T23:59:59.000Z

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


101

Integration of EBS Models with Generic Disposal System Models | Department  

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

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

102

Integration of EBS Models with Generic Disposal System Models | Department  

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

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

103

Generic CTS Data Upload Templates | Department of Energy  

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

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

104

3116 WASTE DETERMINATIONS PUBLIC MEETINGS AND GENERIC TECHNICAL ISSUES  

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

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

105

Evaluation of Generic EBS Design Concepts and Process Models Implications  

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

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

106

Multi-document summarisation using generic relation extraction  

Science Conference Proceedings (OSTI)

Experiments are reported that investigate the effect of various source document representations on the accuracy of the sentence extraction phase of a multi-document summarisation task. A novel representation is introduced based on generic relation extraction ...

Ben Hachey

2009-08-01T23:59:59.000Z

107

Business Case Slide 5: DU Use Categories: Generic Aspects - 2  

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

DU Use Categories: Generic Aspects - 2 Production of DUO2 Virtually all potential DU uses require DUO2 Planned UDS product is DUO2DU3O8 Changes in UDS flowsheet to produce DUO2...

108

Business Case Slide 20: High-Value: Generic - 1  

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

High-Value: Generic - 1 Potential Benefits EM Revenues: high-value applications can afford to pay more than a nominal amount for DU Positive program visibility: "swords to...

109

Developing a Generic Agent-based Intelligent Tutoring System  

E-Print Network (OSTI)

We developed a generic teaching environment that uses agents to support learning. An agent platform developed in our research group is used. The system is generic in the sense that the cognition core (domain model, student model and instruction model) is separated from the exercise modules and user interfaces. The architecture allows different user interfaces. An application of the environment for nurse training has been implemented.

Hospers Kroezen Nijholt; M. Hospers; E. Kroezen; A. Nijholt; R. Op Den Akker; D. Heylen

2003-01-01T23:59:59.000Z

110

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

SciTech Connect

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

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

2012-09-17T23:59:59.000Z

111

J. Symbolic Computation (1999) 11, 1{000 Generic and Cogeneric Monomial Ideals  

E-Print Network (OSTI)

J. Symbolic Computation (1999) 11, 1{000 Generic and Cogeneric Monomial Ideals EZRA MILLER, BERND by simplicial complexes. There are numerous equivalent ways to say that a monomial ideal is generic or cogeneric lexicographic initial ideals of generic lattice ideals are generic. Cohen-Macaulayness for cogeneric ideals

Miller, Ezra N.

112

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

DOE Green Energy (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

113

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)

Fraas, A.P.; Mills, C.B.

1961-11-21T23:59:59.000Z

114

REACTOR DEVELOPMENT PROGRAM PROGRESS REPORT, FEBRUARY 1961  

SciTech Connect

Design, development, and testing efforts were continued on BORAX-V, EBR- I, EBR-II, EBWR, JUGGERNAUT ZPRIII, ZPR-VI, and ZPR-W. An evaluation program is outlined for Pebble Bed Reactor designs. Fast and thermal reactor safety studies were conducted. Experimental and theoretical studies in applied nuclear and reactor physics are dsscribed. Developments made in reactor components, fuels, and materials are discussed. Heat engineering studies were conducted on steam separation, and velocity and void distributions in two-phase systems. Fluidization and fluoride volatility separation, and chemical-metallurgical separation processes were studied. Advanced reactor concepts that were discusssd includsed. Basic Radiation Effects Beactor, Biogeonuclear Reactor, Fast Reactor Test Facility, compact high-power density fast reactors, AHFR hydraulic test loop, Packed Bed Reactor, and direct conversion. (For preceding period see ANL- 6328.) (B.O.G.)

1961-03-15T23:59:59.000Z

115

REACTOR COOLING  

DOE Patents (OSTI)

A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

Quackenbush, C.F.

1959-09-29T23:59:59.000Z

116

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

Wigner, E.P.

1958-04-22T23:59:59.000Z

117

Categorizing Threat Building and Using a Generic Threat Matrix | Department  

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

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

118

A NEUTRONIC REACTOR  

DOE Patents (OSTI)

A nuclear reactor for producing thermoelectric power is described. The reactor core comprises a series of thermoelectric assemblies, each assembly including fissionable fuel as an active element to form a hot junction and a thermocouple. The assemblies are disposed parallel to each other to form spaces and means are included for Introducing an electrically conductive coolant between the assemblies to form cold junctions of the thermocouples. An electromotive force is developed across the entire series of the thermoelectric assemblies due to fission heat generated in the fuel causing a current to flow perpendicular to the flow of coolant and is distributed to a load outside of the reactor by means of bus bars electrically connected to the outermost thermoelectric assembly.

Luebke, E.A.; Vandenberg, L.B.

1959-09-01T23:59:59.000Z

119

MERCHANT MARINE SHIP REACTOR  

DOE Patents (OSTI)

A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.

Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.

1961-05-01T23:59:59.000Z

120

MERCHANT MARINE SHIP REACTOR  

DOE Patents (OSTI)

A nuclear reactor for use in a merchant marine ship is described. The reactor is of pressurized, light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements that are confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass. (AEC)

Sankovich, M.F.; Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Gestson, D.K.

1961-05-01T23:59:59.000Z

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

Generic Error Model of Human-Robot Interaction  

E-Print Network (OSTI)

Wrong human-robot interactions are at the origin of severe damages. Safety requirements ask the analysis of these interactions. At first, erroneous interactions have to be identified. In this paper, we propose to use UML (Unified Modeling Language) to specify human robot interaction. Then, generic error models, associated with the message feature provided by UML, are presented. These error models allow interaction errors to be automatically deduced from the modeling of the human-robot interactions. The use of these generic error models is illustrated on a medical robot for teleechography.

J. Guiochet; et al.

2004-01-01T23:59:59.000Z

122

Advances in reactor physics education: Visualization of reactor parameters  

Science Conference Proceedings (OSTI)

Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for reactor operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and a typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software. (authors)

Snoj, L.; Kromar, M.; Zerovnik, G. [Josef Stefan Inst., Jamova cesta 39, SI-1000 Ljubljana (Slovenia)

2012-07-01T23:59:59.000Z

123

Automated startup of the MIT research reactor  

SciTech Connect

This summary describes the development, implementation, and testing of a generic method for performing automated startups of nuclear reactors described by space-independent kinetics under conditions of closed-loop digital control. The technique entails first obtaining a reliable estimate of the reactor's initial degree of subcriticality and then substituting that estimate into a model-based control law so as to permit a power increase from subcritical on a demanded trajectory. The estimation of subcriticality is accomplished by application of the perturbed reactivity method. The shutdown reactor is perturbed by the insertion of reactivity at a known rate. Observation of the resulting period permits determination of the initial degree of subcriticality. A major advantage to this method is that repeated estimates are obtained of the same quantity. Hence, statistical methods can be applied to improve the quality of the calculation.

Kwok, K.S. (Massachusetts Inst. of Tech., Cambridge (United States))

1992-01-01T23:59:59.000Z

124

NUCLEAR REACTOR  

DOE Patents (OSTI)

A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

Moore, R.V.; Bowen, J.H.; Dent, K.H.

1958-12-01T23:59:59.000Z

125

Generic Terminology Versus Specific Jargon in Power System Switching  

Science Conference Proceedings (OSTI)

This Technical Update report summarizes work performed in 2008 on the "Generic Terminology versus Specific Jargon in Power System Switching" task conducted by the EPRI Power Switching and Reliability project. Previous work was documented in reports 1013915 and 1012374. Research indicates that use of jargon or unfamiliar terms in power switching can result in switching errors, hazardous situations, and unintended power outages.

2008-11-11T23:59:59.000Z

126

Short communication: A generic framework for land-use modelling  

Science Conference Proceedings (OSTI)

In this paper we present the generic modelling system SITE (SImulation of Terrestrial Environments), a software package to develop and apply models simulating regional land-use dynamics. The modelling system includes (i) a framework managing the model ... Keywords: Cellular automata, Genetic algorithm, Integrated framework, Land-use model, Map comparison

Christian Schweitzer; Joerg A. Priess; Subhashree Das

2011-08-01T23:59:59.000Z

127

A generic design environment for the rural industry knowledge acquisition  

Science Conference Proceedings (OSTI)

This paper describes a new knowledge acquisition method using a generic design environment where context-sensitive knowledge is used to build specific DSS for rural business. Although standard knowledge acquisition methods have been applied in rural ... Keywords: DSS process, Design environment, Knowledge acquisition, Rural application

Shah Jahan Miah; Don Kerr; John Gammack; Tom Cowan

2008-12-01T23:59:59.000Z

128

Glift: Generic, efficient, random-access GPU data structures  

Science Conference Proceedings (OSTI)

This article presents Glift, an abstraction and generic template library for defining complex, random-access graphics processor (GPU) data structures. Like modern CPU data structure libraries, Glift enables GPU programmers to separate algorithms from ... Keywords: Adaptive, GPGPU, GPU, adaptive shadow maps, data structures, graphics hardware, multiresolution, octree textures, parallel computation

Aaron E. Lefohn; Shubhabrata Sengupta; Joe Kniss; Robert Strzodka; John D. Owens

2006-01-01T23:59:59.000Z

129

Generic repository design concepts and thermal analysis (FY11).  

SciTech Connect

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

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

2011-08-01T23:59:59.000Z

130

Reactor Materials  

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

The reactor materials crosscut effort will enable the development of innovative and revolutionary materials and provide broad-based, modern materials science that will benefit all four DOE-NE...

131

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor incorporating seed and blanket assemblies is designed. Means are provided for obtaining samples of the coolant from the blanket assemblies and for varying the flow of coolant through the blanket assemblies. (AEC)

Sherman, J.; Sharbaugh, J.E.; Fauth, W.L. Jr.; Palladino, N.J.; DeHuff, P.G.

1962-10-23T23:59:59.000Z

132

NEUTRONIC REACTORS  

DOE Patents (OSTI)

A nuclear reactor is described wherein horizontal rods of thermal- neutron-fissionable material are disposed in a body of heavy water and extend through and are supported by spaced parallel walls of graphite.

Wigner, E.P.

1960-11-22T23:59:59.000Z

133

REACTOR SHIELD  

DOE Patents (OSTI)

Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

1959-02-17T23:59:59.000Z

134

Generic on-line/off-line threshold signatures  

E-Print Network (OSTI)

Tauman [16] and the distributed key generation protocol ofdiscrete logarithm distributed key generation algorithm ofscheme with distributed key generation and distributed

Crutchfield, C; Molnar, D; Turner, D; Wagner, D

2006-01-01T23:59:59.000Z

135

NUCLEAR REACTOR  

DOE Patents (OSTI)

High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

Grebe, J.J.

1959-07-14T23:59:59.000Z

136

Assessment of industrial attitudes toward generic research needs in tribology  

SciTech Connect

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

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

1985-09-01T23:59:59.000Z

137

Storing and managing data in a distributed hash table  

E-Print Network (OSTI)

Distributed hash tables (DHTs) have been proposed as a generic, robust storage infrastructure for simplifying the construction of large-scale, wide-area applications. For example, UsenetDHT is a new design for Usenet News ...

Sit, Emil, 1977-

2008-01-01T23:59:59.000Z

138

Independent Verification of Research Reactor Operation (Analysis of the Georgian IRT-M Reactor by the Isotope Ratio Method)  

SciTech Connect

The U.S. Department of Energy’s Office of Nonproliferation and International Security (NA-24) develops technologies to aid in implementing international nuclear safeguards. The Isotope Ratio Method (IRM) was successfully developed in 2005 – 2007 by Pacific Northwest National Laboratory (PNNL) and the Republic of Georgia’s Andronikashvili Institute of Physics as a generic technology to verify the declared operation of water-moderated research reactors, independent of spent fuel inventory. IRM estimates the energy produced over the operating lifetime of a fission reactor by measuring the ratios of the isotopes of trace impurity elements in non-fuel reactor components.The Isotope Ratio Method is a technique for estimating the energy produced over the operating lifetime of a fission reactor by measuring the ratios of the isotopes of impurity elements in non-fuel reactor components.

Cliff, John B.; Frank, Douglas P.; Gerlach, David C.; Gesh, Christopher J.; Little, Winston W.; Reid, Bruce D.; Tsiklauri, Georgi V.; Abramidze, Sh; Rostomashvili, Z.; Kiknadze, G.; Dzhavakhishvily, O.; Nabakhtiani, G.

2010-08-11T23:59:59.000Z

139

Research reactors - an overview  

SciTech Connect

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

West, C.D.

1997-03-01T23:59:59.000Z

140

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

SciTech Connect

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

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

2001-05-15T23:59:59.000Z

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

Brookhaven Medical Research Reactor  

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

Medical Research Reactor BMRR The last of the Lab's reactors, the Brookhaven Medical Research Reactor (BMRR), was shut down in December 2000. The BMRR was a three megawatt...

142

NEUTRONIC REACTOR  

DOE Patents (OSTI)

This patent relates to neutronic reactors of the heterogeneous water cooled type, and in particular to a fuel element charging and discharging means therefor. In the embodiment illustrated the reactor contains horizontal, parallel coolant tubes in which the fuel elements are disposed. A loading cart containing a magnzine for holding a plurality of fuel elements operates along the face of the reactor at the inlet ends of the coolant tubes. The loading cart is equipped with a ram device for feeding fuel elements from the magazine through the inlot ends of the coolant tubes. Operating along the face adjacent the discharge ends of the tubes there is provided another cart means adapted to receive irradiated fuel elements as they are forced out of the discharge ends of the coolant tubes by the incoming new fuel elements. This cart is equipped with a tank coataining a coolant, such as water, into which the fuel elements fall, and a hydraulically operated plunger to hold the end of the fuel element being discharged. This inveation provides an apparatus whereby the fuel elements may be loaded into the reactor, irradiated therein, and unloaded from the reactor without stopping the fiow of the coolant and without danger to the operating personnel.

Ohlinger, L.A.; Wigner, E.P.; Weinberg, A.M.; Young, G.J.

1958-09-01T23:59:59.000Z

143

J. Symbolic Computation (1999) 11, 1-000 Generic and Cogeneric Monomial Ideals  

E-Print Network (OSTI)

J. Symbolic Computation (1999) 11, 1-000 Generic and Cogeneric Monomial Ideals initial ideals of generic * *lattice ideals are generic. Cohen-Macaulayness for cogeneric ideals is characterized combina* *torially; in the cogeneric case the Cohen-Macaulay type is greater than or equal

Miller, Ezra N.

144

POWER REACTOR  

DOE Patents (OSTI)

A fast nuclear reactor system ls described for producing power and radioactive isotopes. The reactor core is of the heterogeneous, fluid sealed type comprised of vertically arranged elongated tubular fuel elements having vertical coolant passages. The active portion is surrounded by a neutron reflector and a shield. The system includes pumps and heat exchangers for the primary and secondary coolant circuits. The core, primary coolant pump and primary heat exchanger are disposed within an irapenforate tank which is filled with the primary coolant, in this case a liquid metal such as Na or NaK, to completely submerge these elements. The tank is completely surrounded by a thick walled concrete shield. This reactor system utilizes enriched uranium or plutonium as the fissionable material, uranium or thorium as a diluent and thorium or uranium containing less than 0 7% of the U/sup 235/ isotope as a fertile material.

Zinn, W.H.

1958-07-01T23:59:59.000Z

145

REACTOR CONTROL  

DOE Patents (OSTI)

A control system employed with a high pressure gas cooled reactor in which a control rod is positioned for upward and downward movement into the neutron field from a position beneath the reactor is described. The control rod is positioned by a coupled piston cylinder releasably coupled to a power drive means and the pressurized coolant is directed against the lower side of the piston. The coolant pressure is offset by a higher fiuid pressure applied to the upper surface of the piston and means are provided for releasing the higher pressure on the upper side of the piston so that the pressure of the coolant drives the piston upwardly, forcing the coupled control rod into the ncutron field of the reactor. (AEC)

Fortescue, P.; Nicoll, D.

1962-04-24T23:59:59.000Z

146

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor of the homogeneous liquid fuel type is described wherein the fissionable isotope is suspended or dissolved in a liquid moderator such as water. The reactor core is comprised essentially of a spherical vessel for containing the reactive composition surrounded by a reflector, preferably of beryllium oxide. The reactive composition may be an ordinary water solution of a soluble salt of uranium, the quantity of fissionable isotope in solution being sufficient to provide a critical mass in the vessel. The liquid fuel is stored in a tank of non-crtttcal geometry below the reactor vessel and outside of the reflector and is passed from the tank to the vessel through a pipe connecting the two by air pressure means. Neutron absorbing control and safety rods are operated within slots in the reflector adjacent to the vessel.

Christy, R.F.

1958-07-15T23:59:59.000Z

147

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

148

NEUTRONIC REACTORS  

DOE Patents (OSTI)

A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

Wigner, E.P.; Young, G.J.

1958-10-14T23:59:59.000Z

149

NUCLEAR REACTOR  

DOE Patents (OSTI)

This patent covers a power-producing nuclear reactor in which fuel rods of slightly enriched U are moderated by heavy water and cooled by liquid metal. The fuel rods arranged parallel to one another in a circle are contained in a large outer closed-end conduit that extends into a tank containing the heavy water. Liquid metal is introduced into the large conduit by a small inner conduit that extends within the circle of fuel rods to a point near the lower closed end of the outer conduit. (AEC) Production Reactors

Young, G.

1963-01-01T23:59:59.000Z

150

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

Wigner, E.P.; Weinberg, A.W.; Young, G.J.

1958-04-15T23:59:59.000Z

151

Power Burst Facility (PBF) Reactor Reactor Decommissioning  

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

Reactor Decommissioning Click here to view Click here to view Reactor Decommissioning Click on an image to enlarge A crane removes the reactor vessel from the Power Burst Facility...

152

Generic Terminology vs. Specific Jargon in Power System Switching  

Science Conference Proceedings (OSTI)

EPRI research has revealed that terminology used in switching operations varies among utilities, operating regions, and even within utility divisions. But just how different is the terminology, and can these differences compromise efficiency....or even safety? To find out, EPRI engaged Berkshire Harrison to solicit, collect, and analyze switching jargon used by member organizations, identify potential conflicts, and investigate the feasibility of compiling a "dictionary" of generic switching terms and th...

2007-09-19T23:59:59.000Z

153

Developing Generic Dynamic Models for the 2030 Eastern Interconnection Grid  

SciTech Connect

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

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

2013-12-01T23:59:59.000Z

154

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A reactor is described comprising a plurality of horizontal trays containing a solution of a fissionable material, the trays being sleeved on a vertical tube which contains a vertically-reciprocable control rod, a gas-tight chamber enclosing the trays, and means for conducting vaporized moderator from the chamber and for replacing vaporized moderator in the trays. (AEC)

Wigner, E.P.

1962-12-25T23:59:59.000Z

155

Neutronic reactor  

DOE Patents (OSTI)

A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield.

Wende, Charles W. J. (West Chester, PA)

1976-08-17T23:59:59.000Z

156

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor is described that includes spaced vertical fuel elements centrally disposed in a pressure vessel, a mass of graphite particles in the pressure vessel, means for fluidizing the graphite particles, and coolant tubes in the pressure vessel laterally spaced from the fuel elements. (AEC)

Post, R.G.

1963-05-01T23:59:59.000Z

157

NUCLEAR REACTOR  

DOE Patents (OSTI)

This patent relates to a combination useful in a nuclear reactor and is comprised of a casing, a mass of graphite irapregnated with U compounds in the casing, and at least one coolant tube extending through the casing. The coolant tube is spaced from the mass, and He is irtroduced irto the space between the mass and the coolant tube. (AEC)

Starr, C.

1963-01-01T23:59:59.000Z

158

NEUTRONIC REACTOR  

DOE Patents (OSTI)

BS>A reactor cooled by water, biphenyl, helium, or other fluid with provision made for replacing the fuel rods with the highest plutonium and fission product content without disassembling the entire core and for promptly cooling the rods after their replacement in order to prevent build-up of heat from fission product activity is described.

Creutz, E.C.; Ohlinger, L.A.; Weinberg, A.M.; Wigner, E.P.; Young, G.J.

1959-10-27T23:59:59.000Z

159

NEUTRONIC REACTORS  

DOE Patents (OSTI)

The design of control rods for nuclear reactors are described. In this design the control rod consists essentially of an elongated member constructed in part of a neutron absorbing material and having tube means extending therethrough for conducting a liquid to cool the rod when in use.

Anderson, H.L.

1958-10-01T23:59:59.000Z

160

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

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

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

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

NEUTRON DENSITY CONTROL IN A NEUTRONIC REACTOR  

DOE Patents (OSTI)

The method and means for controlling the neutron density in a nuclear reactor is described. It describes the method and means for flattening the neutron density distribution curve across the reactor by spacing the absorbing control members to varying depths in the central region closer to the center than to the periphery of the active portion of the reactor to provide a smaller neutron reproduction ratio in the region wherein the members are inserted, than in the remainder of the reactor thereby increasing the over-all potential power output.

Young, G.J.

1959-06-30T23:59:59.000Z

162

REACTOR UNLOADING  

DOE Patents (OSTI)

This patent is related to gas cooled reactors wherein the fuel elements are disposed in vertical channels extending through the reactor core, the cooling gas passing through the channels from the bottom to the top of the core. The invention is a means for unloading the fuel elements from the core and comprises dump values in the form of flat cars mounted on wheels at the bottom of the core structure which support vertical stacks of fuel elements. When the flat cars are moved, either manually or automatically, for normal unloading purposes, or due to a rapid rise in the reproduction ratio within the core, the fuel elements are permtted to fall by gravity out of the core structure thereby reducing the reproduction ratio or stopping the reaction as desired.

Leverett, M.C.

1958-02-18T23:59:59.000Z

163

NUCLEAR REACTOR  

DOE Patents (OSTI)

A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

Treshow, M.

1958-08-19T23:59:59.000Z

164

Neutronic reactor  

DOE Patents (OSTI)

A graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels.

Lewis, Warren R. (Richland, WA)

1978-05-30T23:59:59.000Z

165

NUCLEAR REACTORS  

DOE Patents (OSTI)

An active portion assembly for a fast neutron reactor is described wherein physical distortions resulting in adverse changes in the volume-to-mass ratio are minimized. A radially expandable locking device is disposed within a cylindrical tube within each fuel subassembly within the active portion assembly, and clamping devices expandable toward the center of the active portion assembly are disposed around the periphery thereof. (AEC)

Koch, L.J.; Rice, R.E. Jr.; Denst, A.A.; Rogers, A.J.; Novick, M.

1961-12-01T23:59:59.000Z

166

REACTOR CONTROL  

DOE Patents (OSTI)

This patent relates to nuclear reactors of the type which utilize elongited rod type fuel elements immersed in a liquid moderator and shows a design whereby control of the chain reaction is obtained by varying the amount of moderator or reflector material. A central tank for containing liquid moderator and fuel elements immersed therein is disposed within a surrounding outer tank providing an annular space between the two tanks. This annular space is filled with liquid moderator which functions as a reflector to reflect neutrons back into the central reactor tank to increase the reproduction ratio. Means are provided for circulating and cooling the moderator material in both tanks and additional means are provided for controlling separately the volume of moderator in each tank, which latter means may be operated automatically by a neutron density monitoring device. The patent also shows an arrangement for controlling the chain reaction by injecting and varying an amount of poisoning material in the moderator used in the reflector portion of the reactor.

Ruano, W.J.

1957-12-10T23:59:59.000Z

167

Light Water Reactors Technology Development - Nuclear Reactors  

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

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

168

FastTrack Generic Data Link 1.0  

Science Conference Proceedings (OSTI)

The FastTrack Generic Data Link version 1.0 software system utilizes EPRI's Data Transfer Tool to allow users to transfer waste processing data from plant or corporate databases into a holding database that can be queried, on demand, by FastTrack users over a corporate intranet. This software system consists of four required software elements: the Data Transfer Tool 3.1, FastTrack 2.0 (provided separately as EPRI product 1009563), the FastTrack Data Link databases, and the FastTrack Data Link web service...

2005-06-02T23:59:59.000Z

169

An Infrastructure for Distributed Computing and Context Aware Computing  

E-Print Network (OSTI)

This paper focuses what we see as main issues in context awareness, small size and low power of today's mobile computing devices. Until now, these devices are heterogenous and mostly used standalone to accomplish certain task. That means there is no a single generic type device and user needs to be aware of different devices to perform different tasks, which is of course practically inconvenient. We propose a generic device and a network infrastructure which address low power consumption, context aware computing and protable device for the ubiquitous computing environment. Where the generic device (as a mobile device) interacts with the different distributed devices which are fixed in the physical infrastructure.

Sujan Pandey; Sujan P; Peter Zipf; Oliver Soffke; Mihail Petrov; Tudor Murgan; Max Mühlhäuser; Manfred Glesner

2003-01-01T23:59:59.000Z

170

Monte Carlo simulations for generic granite repository studies  

SciTech Connect

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

171

Revised Petition To Establish A New Generic Subclass  

E-Print Network (OSTI)

Petition to the Federal Trade Commission (the “Commission”) for the establishment of a new generic subclass within the existing polyester category for fibers made from poly(trimethylene terephthalate) (“PTT”) to restate and supplement the contents of the Petition dated February, 21, 2006 in order to address certain questions raised by the Commission. Petitioners note the Commission’s April 18, 2006 action taken in response to the February 21, 2006 Petition assigning the designation “PTT001 ” for PTT fiber for temporary use until a final determination is made by the Commission as to the merits of this Petition. Petitioners propose in order of preference the following names for a new generic subclass of polyester that may be used with respect to PTT fibers: 1. triexta; 2. resisoft; and 3. durares. 1 Mohawk was founded more than 120 years ago and today is the leading producer and distributor of flooring worldwide. Mohawk products serve all major flooring categories: carpet, rugs, hardwood, laminate, ceramic tile, and vinyl flooring. Mohawk has launched a line of carpets manufactured from PTT and sells such carpets under the trademark SmartStrand. 2 Founded in 1802, DuPont is a science company operating in more than 70 countries. DuPont offers a wide range of innovative products and services for markets including agriculture, nutrition, electronics, communications, safety and protection, home and construction, transportation and apparel. DuPont markets PTT under the trademark Sorona®. 3

unknown authors

2006-01-01T23:59:59.000Z

172

Revised Industry Steam Generator Program Generic License Change Package  

E-Print Network (OSTI)

License Change Package for NRC review and endorsement. Industry events during the subsequent months have delayed the NRC review and presented an opportunity for the industry to improve the submittal. A summary of the major changes from the February submittal is provided in Enclosure 1. The resulting package (Enclosures 2 through 8), which supercedes the earlier version in its entirety, is enclosed for your endorsement. Although the events of the last ten months have resulted in some changes to the industry steam generator program guidance, the fundamental principles remain sound. Equally as important, the program has demonstrated its resiliency. The steam generator program guidance is designed to accommodate new knowledge and experience, and that is precisely what is occurring. The enclosed Generic License Change Package includes changes that reflect recent experience. In addition, the appropriate underlying technical documents are currently being revised or supplemental guidance is being developed as necessary to reflect new information. The industry and the NRC worked diligently to address the technical and licensing issues that ultimately resulted in the February 2000 version of the Steam Generator Program Generic License Change Package. Although these documents have been revised since the previous submittal, the differences do not represent a change in our position on the issues.

David J. Modeen; Mr. Samuel; J. Collins; U. S. Nuclear; Regulatory Commission; Xuo Mr; Samuel J. Collins

2000-01-01T23:59:59.000Z

173

Nuclear reactor  

DOE Patents (OSTI)

A nuclear reactor in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assmblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters and the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters and the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance.

Pennell, William E. (Greensburg, PA); Rowan, William J. (Monroeville, PA)

1977-01-01T23:59:59.000Z

174

ELECTRONUCLEAR REACTOR  

DOE Patents (OSTI)

An electronuclear reactor is described in which a very high-energy particle accelerator is employed with appropriate target structure to produce an artificially produced material in commercial quantities by nuclear transformations. The principal novelty resides in the combination of an accelerator with a target for converting the accelerator beam to copious quantities of low-energy neutrons for absorption in a lattice of fertile material and moderator. The fertile material of the lattice is converted by neutron absorption reactions to an artificially produced material, e.g., plutonium, where depleted uranium is utilized as the fertile material.

Lawrence, E.O.; McMillan, E.M.; Alvarez, L.W.

1960-04-19T23:59:59.000Z

175

Photocatalytic reactor  

DOE Patents (OSTI)

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

Bischoff, Brian L. (Knoxville, TN); Fain, Douglas E. (Oak Ridge, TN); Stockdale, John A. D. (Knoxville, TN)

1999-01-01T23:59:59.000Z

176

Distribution System Cyber Security Architecture  

Science Conference Proceedings (OSTI)

The National Institute of Standards and Technology (NIST) has developed the NISTIR 7628: Guidelines for Smart Grid Cyber Security, while ASAP-SG has developed both the Security Profile for AMI (v2.0) and the Security Profile for Distribution Management. These documents are necessarily high level and generic, covering a broad range of smart grid assets, and focus exclusively on determining the security requirements. However, utilities and their vendors often cannot get a clear picture of what existing tec...

2010-12-31T23:59:59.000Z

177

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

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

Field Evolution of a Generic Clay Repository: Model Validation and Demonstration Rev 2 Shale and clay-rich rock formations have been considered as potential host rocks for...

178

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

Science Conference Proceedings (OSTI)

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

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

1989-01-01T23:59:59.000Z

179

CONTROL MEANS FOR REACTOR  

DOE Patents (OSTI)

An apparatus for controlling a nuclear reactor includes a tank just below the reactor, tubes extending from the tank into the reactor, and a thermally expansible liquid neutron absorbent material in the tank. The liquid in the tank is exposed to a beam of neutrons from the reactor which heats the liquid causing it to expand into the reactor when the neutron flux in the reactor rises above a predetermincd danger point. Boron triamine may be used for this purpose.

Manley, J.H.

1961-06-27T23:59:59.000Z

180

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

Science Conference Proceedings (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

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.


181

Nuclear Reactor Accidents  

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

Reactor Accidents The accidents at the Three Mile Island (TMI) and Chernobyl nuclear reactors have triggered particularly intense concern about radiation hazards. The TMI accident,...

182

Principles of Reactor Physics  

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

Nuclear Reactor Physics M A Smith Argonne National Laboratory Nuclear Engineering Division Phone: 630-252-9747, Email: masmith@anl.gov Abstract: Nuclear reactor physics deals with...

183

Power Flow Control Using Distributed Saturable Reactors  

V S /G V R /0q Z jX 2 cos S R S S Rsin SS V V V V V PQ XX G G 3 Managed by UT-Battelle for the Department of Energy Simple Example 3 x 150 MVA lines supply 250 MW load

184

DistributionCategory: Magnetic Fusion Reactor Materials  

E-Print Network (OSTI)

that its use would not infringe privately owned rights. Reference herein to any specific commercial product. Several analyses of this vapor shielding effect have been performed [l-71.Some previouswork focused on one function q (Z,t) is calculated, in the condensed target material, with detailed models that include

Harilal, S. S.

185

GENERIC MODEL FOR MAGNETIC EXPLOSIONS APPLIED TO SOLAR FLARES  

Science Conference Proceedings (OSTI)

An accepted model for magnetospheric substorms is proposed as the basis for a generic model for magnetic explosions and is applied to solar flares. The model involves widely separated energy-release and particle-acceleration regions, with energy transported Alfvenically between them. On a global scale, these regions are coupled by a large-scale current that is set up during the explosion by redirection of pre-existing current associated with the stored magnetic energy. The explosion-related current is driven by an electromotive force (EMF) due to the changing magnetic flux enclosed by this current. The current path and the EMF are identified for an idealized quadrupolar model for a flare.

Melrose, D. B. [Sydney Institute for Astronomy, School of Physics, University of Sydney, NSW 2006 (Australia)

2012-04-10T23:59:59.000Z

186

Development of Parsing Tools for CASL using Generic Language Technology  

E-Print Network (OSTI)

An environment for the Common Algebraic Specification Language CASL consists of several independent tools. A number of CASL tools have been built using the algebraic speci cation formalism ASF+SDF and the ASF+SDF Meta-Environment. CASL supports user-defined syntax which is non-trivial to process: ASF+SDF offers a powerful parsing technology (Generalized LR). Its interactive development environment facilitates rapid prototyping complemented by early detection and correction of errors. A number of core technologies developed for the ASF+SDF Meta-Environment can be reused in the context of CASL. Furthermore, an instantiation of a generic format developed for the representation of ASF+SDF specifications and terms provides a CASL-specific exchange format.

Mark Van Den Brand; Jeroen Scheerder

2000-01-01T23:59:59.000Z

187

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

SciTech Connect

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

188

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A power plant is described comprising a turbine and employing round cylindrical fuel rods formed of BeO and UO/sub 2/ and stacks of hexagonal moderator blocks of BeO provided with passages that loosely receive the fuel rods so that coolant may flow through the passages over the fuels to remove heat. The coolant may be helium or steam and fiows through at least one more heat exchanger for producing vapor from a body of fluid separate from the coolant, which fluid is to drive the turbine for generating electricity. By this arrangement the turbine and directly associated parts are free of particles and radiations emanating from the reactor. (AEC)

Daniels, F.

1962-12-18T23:59:59.000Z

189

Reactor and method of operation  

DOE Patents (OSTI)

A nuclear reactor having a flattened reactor activity curve across the reactor includes fuel extending over a lesser portion of the fuel channels in the central portion of the reactor than in the remainder of the reactor.

Wheeler, John A. (Princeton, NJ)

1976-08-10T23:59:59.000Z

190

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

191

Economic analysis of nuclear reactors  

SciTech Connect

The report presents several methods for estimating the power costs of nuclear reactors. When based on a consistent set of economic assumptions, total power costs may be useful in comparing reactor alternatives. The principal items contributing to the total power costs of a nuclear power plant are: (1) capital costs, (2) fuel cycle costs, (3) operation and maintenance costs, and (4) income taxes and fixed charges. There is a large variation in capital costs and fuel expenses among different reactor types. For example, the standard once-through LWR has relatively low capital costs; however, the fuel costs may be very high if U/sub 3/O/sub 8/ is expensive. In contrast, the FBR has relatively high capital costs but low fuel expenses. Thus, the distribution of expenses varies significantly between these two reactors. In order to compare power costs, expenses and revenues associated with each reactor may be spread over the lifetime of the plant. A single annual cost, often called a levelized cost, may be obtained by the methods described. Levelized power costs may then be used as a basis for economic comparisons. The paper discusses each of the power cost components. An exact expression for total levelized power costs is derived. Approximate techniques of estimating power costs will be presented.

Owen, P.S.; Parker, M.B.; Omberg, R.P.

1979-05-01T23:59:59.000Z

192

A Hybrid Architectural Style for Distributed Parallel Processing of Generic Data Streams  

Science Conference Proceedings (OSTI)

Immersive, interactive applications grouped under theconcept of Immersipresence require on-line processing andmixing of multimedia data streams and structures. One criticalissue seldom addressed is the integration of differentsolutions to technical challenges, ...

Alexandre R. J. Francois

2004-05-01T23:59:59.000Z

193

DAGuE: A generic distributed DAG engine for High Performance Computing  

Science Conference Proceedings (OSTI)

The frenetic development of the current architectures places a strain on the current state-of-the-art programming environments. Harnessing the full potential of such architectures is a tremendous task for the whole scientific computing community. We ... Keywords: Architecture aware scheduling, HPC, Heterogeneous architectures, Micro-task DAG

George Bosilca; Aurelien Bouteiller; Anthony Danalis; Thomas Herault; Pierre Lemarinier; Jack Dongarra

2012-01-01T23:59:59.000Z

194

Basis for NGNP Reactor Design Down-Selection  

Science Conference Proceedings (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

195

Basis for NGNP Reactor Design Down-Selection  

SciTech Connect

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

196

Modelling structures in generic space, a condition for adaptiveness of monitoring cognitive agent  

Science Conference Proceedings (OSTI)

The adaptiveness of agents is one of the basic conditions for the autonomy. This paper describes an approach of adaptiveness forMonitoring Cognitive Agents based on the notion of "generic spaces". This notion allows the definition of virtual "generic ... Keywords: autonomous agent, cognitive agent, diagnosis, expert systems, monitoring

Marc Goc; Michel Gaeta

2005-01-01T23:59:59.000Z

197

Multi-stage programming with functors and monads: Eliminating abstraction overhead from generic code  

Science Conference Proceedings (OSTI)

We use multi-stage programming, monads and Ocaml's advanced module system to demonstrate how to eliminate all abstraction overhead from generic programs, while avoiding any inspection of the resulting code. We demonstrate this clearly with Gaussian Elimination ... Keywords: Functor, Generative, Genericity, Linear algebra, MetaOCaml, Staging, Symbolic

Jacques Carette; Oleg Kiselyov

2011-05-01T23:59:59.000Z

198

On generic context lemmas for higher-order calculi with sharing  

Science Conference Proceedings (OSTI)

This paper proves several generic variants of context lemmas and thus contributes to improving the tools for observational semantics of deterministic and non-deterministic higher-order calculi that use a small-step reduction semantics. The generic (sharing) ... Keywords: Context lemma, Functional programming languages, Lambda calculus, Observational semantics

Manfred Schmidt-Schau?; David Sabel

2010-03-01T23:59:59.000Z

199

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

SciTech Connect

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

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

1997-08-01T23:59:59.000Z

200

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.


201

NEUTRONIC REACTOR MANIPULATING DEVICE  

DOE Patents (OSTI)

A cable connecting a control rod in a reactor with a motor outside the reactor for moving the rod, and a helical conduit in the reactor wall, through which the cable passes are described. The helical shape of the conduit prevents the escape of certain harmful radiations from the reactor. (AEC)

Ohlinger, L.A.

1962-08-01T23:59:59.000Z

202

An Engineering Test Reactor  

SciTech Connect

A relatively inexpensive reactor for the specific purpose of testing a sub-critical portion of another reactor under conditions that would exist during actual operation is discussed. It is concluded that an engineering tool for reactor development work that bridges the present gap between exponential and criticality experiments and the actual full scale operating reactor is feasible. An example of such a test reactor which would not entail development effort to ut into operation is depicted.

Fahrner, T.; Stoker, R.L.; Thomson, A.S.

1951-03-16T23:59:59.000Z

203

Reactor Pressure Vessel Task of Light Water Reactor Sustainability...  

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

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment Reactor Pressure...

204

Generic Disposal System Modeling--Fiscal Year 2011 Progress Report  

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

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

205

Memorandum on Chemical Reactors and Reactor Hazards  

SciTech Connect

Two important problems in the investigation of reactor hazards are the chemical reactivity of various materials employed in reactor construction and the chracteristics of heat transfer under transient conditions, specifically heat transfer when driven by an exponentially increasing heat source (exp t/T). Although these problems are independent of each other, when studied in relation to reactor hazards they may occur in a closely coupled sequence. For example the onset of a dangerous chemical reactor may be due to structural failure of various reactor components under an exponentially rising heat source originating with a runaway nuclear reactor. For this reason, these two problems should eventually be studied together after an exploratory experimental survey has been made in which they are considered separately.

Mills, M.M.; Pearlman, H.; Ruebsamen, W.; Steele, G., Chrisney, J.

1951-07-05T23:59:59.000Z

206

GENERIC STATUS OF THE AMERICAN PIPISTRELLES (VESPERTILIONIDAE) WITH DESCRIPTION OF  

E-Print Network (OSTI)

relationships in the cosmopolitan bat family Vespertilionidae (vesper bats) based on DNA sequence vari- ation, including all taxa with cosmopolitan distributions. An important trend emerging from the study

Horacek, Ivan

207

Design Methodology for Development of Behavioral Synthesis Generic and Synthetic Benchmarks  

E-Print Network (OSTI)

The importance of benchmarking as an essential methodology in evaluating and comparing digital systems synthesis tools is well established. Equally well are the limitations of the majority of current benchmarks recognized. Our primary goal is to lay out the theoretically and statistically sound foundations for addressing the key issues related to the selection of generic benchmarks and the synthesis and analysis of statistical synthetic benchmarks. The methodology for the synthetic benchmark generation uses as an intermediate step the methodology for the generic benchmark selection. The benefits of our methodology for the generic benchmark selection are: (1) it provides the sound statistical foundations for the generic benchmark selection, (2) it establishes the fact that the development of a generic benchmark is a well-defined optimization problem. The advantages of the statistical synthetic benchmarks over the generic benchmarks include: (1) having an unlimited supply of synthetic design examples, (2) obtaining the real-life like design examples which suit a particular design or analysis need, (3) having a test set which better fits all available real-life design examples. Due to the fact that the synthetic benchmark generation is based on the generic benchmark selection, the entire process of the generic and synthetic benchmark development is in sequence. It consists of the following five-stages: (1) collecting and analyzing real life design examples, (2) extracting and selecting the relevant properties that are used to specify designs, (3) deriving a generic benchmark set from the real life design examples, (4) developing the synthetic benchmark generator, and (5) constructing a synthetic benchmark set for the current applications or the future applications based ...

Hm Ar Ks

1996-01-01T23:59:59.000Z

208

Basic and Applied Science Research Reactors - Reactors designed...  

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

BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th...

209

Efficient hardware generation of random variates with arbitrary distributions  

E-Print Network (OSTI)

This paper presents a technique for efficiently generating random numbers from a given probability distribution. This is achieved by using a generic hardware architecture, which transforms uniform random numbers according to a distribution mapping stored in RAM, and a software approximation generator that creates distribution mappings for any given target distribution. This technique has many features not found in current non-uniform random number generators, such as the ability to adjust the target distribution while the generator is running, per-cycle switching between distributions, and the ability to generate distributions with discontinuities in the Probability Density Function. 1.

David B. Thomas; Wayne Luk

2006-01-01T23:59:59.000Z

210

Hexagonal-geometry fast-reactor nodal modeling  

Science Conference Proceedings (OSTI)

Several nodal methods have been developed for simulating power distributions in thermal reactors, and have been tested for applicability to fast reactor problems. The testing was performed in rectangular geometry. Since fast reactor configurations typically use hexagonal assemblies, the most promising of the techniques tested was extended to hexagonal geometry and applied to a range of test cases. The approach selected was the generalized coarse mesh procedure (GCMDT) using different interpolation parameters in different zones.

Caro, E.; Becker, M.

1982-01-01T23:59:59.000Z

211

Generic Planning Target Margin for Rectal Cancer Treatment Setup Variation  

SciTech Connect

Purpose: To calculate the generic planning target margin (GPTM) for patients receiving radiation therapy (RT) for rectal cancer placed in a prone position with a customized cradle for small-bowel exclusion. Methods and Materials: A total of 25 consecutive rectal cancer patients were treated for 25 or 28 fractions in a prone position using a cradle to maximize small bowel exclusion. Treatment planning computed tomography (CT) scans were used to create orthogonally digitally reconstructed radiographs (DRRs) for portal image registration, which were compared with daily portal images from an electronic portal-imaging device (EPID). Translation values needed to align the DRRs and EPIDs were recorded for the superior to inferior (SI), right to left (RL), and anterior to posterior (AP) directions, and used to calculate the GPTM using the four-parameter model. Age, weight, and body mass index were tested compared with the setup variation using a Pearson correlation and a t test for significance. Gender versus setup variation was compared with a t test. Results: A total of 1,723 EPID images were reviewed. The GPTM was 10 mm superior, 8 mm inferior, 7 mm RL and 10 mm AP. Age and gender were unrelated to setup variation. Weight was significantly associated with systematic AP variation (p < 0.05). BMI was significantly associated with systematic SI (p < 0.05) and AP (p < 0.01) variation and random RL variation (p < 0.05). Conclusions: The GPTM for rectal cancer is asymmetric with a maximum of 10 mm in the superior, anterior and posterior dimensions. Body mass index may effect setup variation. Research using advanced treatment planning should include these margins in the planning target volume definition.

Robertson, John M. [Department of Radiation Oncology, William Beaumont Hospital, Royal Oak, MI (United States)], E-mail: jrobertson@beaumont.edu; Campbell, Jonathon P.; Yan Di [Department of Radiation Oncology, William Beaumont Hospital, Royal Oak, MI (United States)

2009-08-01T23:59:59.000Z

212

Neutron Imaging Reveals Lithium Distribution - ORNL Neutron Sciences  

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

imaging instrument at Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR) have successfully mapped the three-dimensional spatial distribution of lithium...

213

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

214

Guidebook to nuclear reactors  

SciTech Connect

A general introduction to reactor physics and theory is followed by descriptions of commercial nuclear reactor types. Future directions for nuclear power are also discussed. The technical level of the material is suitable for laymen.

Nero, A.V. Jr.

1976-05-01T23:59:59.000Z

215

Reactor Sharing Program  

Science Conference Proceedings (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

216

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

217

NEUTRONIC REACTOR POWER PLANT  

DOE Patents (OSTI)

This patent relates to a nuclear reactor power plant incorporating an air-cooled, beryllium oxide-moderated, pebble bed reactor. According to the invention means are provided for circulating a flow of air through tubes in the reactor to a turbine and for directing a sidestream of the circu1ating air through the pebble bed to remove fission products therefrom as well as assist in cooling the reactor. (AEC)

Metcalf, H.E.

1962-12-25T23:59:59.000Z

218

N-TSVS: a Videoconference Tool for Generic Cooperative Groups  

Science Conference Proceedings (OSTI)

This paper presents the architecture and the design of a distributed videoconferencing application called N-TSVS. The structure of the cooperative groups supported by N-TSVS is defined and modelled by graphs: their vertices describe the group members ...

T. Villemur; P. Owezarski; M. Diaz

1998-10-01T23:59:59.000Z

219

Implementing an Observatory Control System-I. A Generic Approach  

E-Print Network (OSTI)

An architectural framework for implementing a distributed observatory control system is presented here. It has been partially realized and tested in the 2m optical and infrared Observatory at Pune, India.

Sunu Engineer

2004-01-06T23:59:59.000Z

220

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

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

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

222

FAST NEUTRON REACTOR  

DOE Patents (OSTI)

A reactor comprising fissionable material in concentration sufficiently high so that the average neutron enengy within the reactor is at least 25,000 ev is described. A natural uranium blanket surrounds the reactor, and a moderating reflector surrounds the blanket. The blanket is thick enough to substantially eliminate flow of neutrons from the reflector.

Soodak, H.; Wigner, E.P.

1961-07-25T23:59:59.000Z

223

NUCLEAR REACTOR CONTROL SYSTEM  

DOE Patents (OSTI)

A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.

Epler, E.P.; Hanauer, S.H.; Oakes, L.C.

1959-11-01T23:59:59.000Z

224

Materials and Fuels for the Current and Advanced Nuclear Reactors III  

Science Conference Proceedings (OSTI)

Jul 15, 2013 ... Lifetime extension of reactors - nuclear materials aging, degradation and ... A Theoretical Model of Corrosion Rate Distribution in Liquid LBE ...

225

The Argonaut Reactor - Reactors designed/built by Argonne National  

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

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

226

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

SciTech Connect

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

NONE

1995-08-01T23:59:59.000Z

227

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

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

Ron Walli Communications 865.576.0226 CASL milestone validates reactor model using TVA data This CASL visualization shows the thermal distribution of neutrons in Watts Bar Unit 1...

228

Nuclear reactor overflow line  

DOE Patents (OSTI)

The overflow line for the reactor vessel of a liquid-metal-cooled nuclear reactor includes means for establishing and maintaining a continuous bleed flow of coolant amounting to 5 to 10% of the total coolant flow through the overflow line to prevent thermal shock to the overflow line when the reactor is restarted following a trip. Preferably a tube is disposed concentrically just inside the overflow line extending from a point just inside the reactor vessel to an overflow tank and a suction line is provided opening into the body of liquid metal in the reactor vessel and into the annulus between the overflow line and the inner tube.

Severson, Wayne J. (Pittsburgh, PA)

1976-01-01T23:59:59.000Z

229

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

230

Spinning fluids reactor  

SciTech Connect

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

231

Fission reactors and materials  

SciTech Connect

The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions.

Frost, B.R.T.

1981-12-01T23:59:59.000Z

232

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.

Cao, Jun

2011-01-01T23:59:59.000Z

233

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

234

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

2011-01-12T23:59:59.000Z

235

AC distribution system for TFTR pulsed loads  

DOE Green Energy (OSTI)

This paper outlines the AC distribution system associated with the Tokamak Fusion Test Reactor and discusses the significant areas related to design, protection, and equipment selection, particularly where there is a departure from normal utility and industrial applications.

Carroll, R.F.; Ramakrishnan, S.; Lemmon, G.N.; Moo, W.I.

1977-01-01T23:59:59.000Z

236

Generic Repository Concepts and Thermal Analysis for Advanced Fuel Cycles  

SciTech Connect

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

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

2011-01-01T23:59:59.000Z

237

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

238

Cost Estimate for an Away-From-Reactor Generic Interim Storage Facility (GISF) for Spent Nuclear Fuel  

Science Conference Proceedings (OSTI)

As nuclear power plants began to run out of storage capacity in spent nuclear fuel (SNF) storage pools, many nuclear operating companies added higher density pool storage racks to increase pool capacity. Most nuclear power plant storage pools have been re-racked one or more times. As many spent fuel storage pools were re-racked to the maximum extent possible, nuclear operating companies began to employ interim dry storage technologies to store SNF in certified casks and canister-based systems outside of ...

2009-05-20T23:59:59.000Z

239

Experimental techniques for hydrodynamic characterization of multiphase flows in slurry-phase bubble-column reactors  

DOE Green Energy (OSTI)

Slurry-phase bubble-column Fischer-Tropsch (FT) reactors are recognized as one of the more promising technologies for converting synthesis gas from coal into liquid fuel products (indirect liquefaction). However, hydrodynamic effects must be considered when attempting to scale these reactors to sizes of industrial interest. The objective of this program is to facilitate characterization of reactor hydrodynamics by developing and applying noninvasive tomographic diagnostics capable of measuring gas holdup spatial distribution in these reactors.

Torczynski, J.R.; O`Hern, T.J.; Adkins, D.R.; Shollenberger, K.A.; Mondy, L.A.; Jackson, N.B.

1994-09-01T23:59:59.000Z

240

HORIZONTAL BOILING REACTOR SYSTEM  

DOE Patents (OSTI)

Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

Treshow, M.

1958-11-18T23:59:59.000Z

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.


241

NUCLEAR REACTORS AND EARTHQUAKES  

SciTech Connect

A book is presented which supplies pertinent seismological information to engineers in the nuclear reactor field. Data are presented on the occurrence, intensity, and wave shapes. Techniques are described for evaluating the response of structures to such events. Certain reactor types and their modes of operation are described briefly. Various protection systems are considered. Earthquake experience in industrial and reactor plants is described. (D.L.C.)

1961-01-01T23:59:59.000Z

242

THERMAL NEUTRONIC REACTOR  

DOE Patents (OSTI)

A novel thermal reactor was designed in which a first reflector formed from a high atomic weight, nonmoderating material is disposed immediately adjacent to the reactor core. A second reflector composed of a moderating material is disposed outwardly of the first reflector. The advantage of this novel reflector arrangement is that the first reflector provides a high slow neutron flux in the second reflector, where irradiation experiments may be conducted with a small effect on reactor reactivity.

Spinrad, B.I.

1960-01-12T23:59:59.000Z

243

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

244

Advanced Nuclear Research Reactor  

SciTech Connect

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

Lolich, J.V.

2004-10-06T23:59:59.000Z

245

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

246

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

247

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

248

HOMOGENEOUS NUCLEAR POWER REACTOR  

DOE Patents (OSTI)

A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

King, L.D.P.

1959-09-01T23:59:59.000Z

249

Generic medicines and regulatory policies : The effects of differentiated price caps and co-payments on drug prices.  

E-Print Network (OSTI)

??When entering the market, pharmaceutical firms face various regulatory factors that will influence the producer prices for their products. Price competition occurs when generic medicines… (more)

Hofmann, Karen

2008-01-01T23:59:59.000Z

250

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

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

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

251

Incipient Transient Detection in Reactor Systems: Experimental and Theoretical Investigation  

SciTech Connect

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

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

2005-09-27T23:59:59.000Z

252

Foreign Research Reactor/Domestic Research Reactor Receipt Coordinator...  

National Nuclear Security Administration (NNSA)

Research ReactorDomestic Research Reactor Receipt Coordinator, Savannah River Nuclear Solutions | National Nuclear Security Administration Our Mission Managing the Stockpile...

253

ADMINISTRATION OF ORNL RESEARCH REACTORS  

SciTech Connect

Organization of the ORNL Operations division for administration of the Oak Ridge Research Reactor, the Low Intensity Testing Reactor, and the Oak Ridge Graphite Reactor is described. (J.R.D.)

Casto, W.R.

1962-08-20T23:59:59.000Z

254

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

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

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

255

Production reactor characteristics  

SciTech Connect

Reactors for the production of special nuclear materials share many similarities with commercial nuclear power plants. Each relies on nuclear fission, uses uranium fuel, and produces large quantities of thermal power. However, there are some important differences in production reactor characteristics that may best be discussed in terms of mission, role, and technology.

Thiessen, C.W.; Hootman, H.E.

1990-01-01T23:59:59.000Z

256

Advanced converter reactors  

SciTech Connect

Advanced converter reactors (ACRs) of primary US interest are those which can be commercialized within about 20 years, and are: Advanced Light-Water Reactors, Spectral-Shift-Control Reactors, Heavy-Water Reactors (CANDU type), and High-Temperature Gas-Cooled Reactors. These reactors can operate on uranium, thorium, or uranium-thorium fuel cycles, but have the greatest fuel utilization on thorium type cycles. The water reactors tend to operate more economically on uranium cycles, while the HTGR is more economical on thorium cycles. Thus, the HTGR had the greatest practical potential for improving fuel utilization. If the US has 3.4 to 4 million tons U/sub 3/O/sub 8/ at reasonable costs, ACRs can make important contributions to maintaining a high nuclear power level for many decades; further, they work well with fast breeder reactors in the long term under symbiotic fueling conditions. Primary nuclear data needs of ACRs are integral measurements of reactivity coefficients and resonance absorption integrals.

Kasten, P.R.

1979-01-01T23:59:59.000Z

257

NEUTRONIC REACTOR BURIAL ASSEMBLY  

DOE Patents (OSTI)

A burial assembly is shown whereby an entire reactor core may be encased with lead shielding, withdrawn from the reactor site and buried. This is made possible by a five-piece interlocking arrangement that may be easily put together by remote control with no aligning of bolt holes or other such close adjustments being necessary.

Treshow, M.

1961-05-01T23:59:59.000Z

258

The Integral Fast Reactor  

SciTech Connect

Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 5 refs., 3 figs.

Till, C.E.; Chang, Y.I. (Argonne National Lab., IL (USA)); Lineberry, M.J. (Argonne National Lab., Idaho Falls, ID (USA))

1990-01-01T23:59:59.000Z

259

THE MISSING CAVITIES IN THE SEEDS POLARIZED SCATTERED LIGHT IMAGES OF TRANSITIONAL PROTOPLANETARY DISKS: A GENERIC DISK MODEL  

Science Conference Proceedings (OSTI)

Transitional circumstellar disks around young stellar objects have a distinctive infrared deficit around 10 {mu}m in their spectral energy distributions, 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 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 {mu}m-sized grains, with little discontinuity in the surface density of the {mu}m-sized grains at the cavity edge. Here we present a generic disk model that can simultaneously account for the general features in IRS, SMA, and Subaru observations. Particularly, the scattered light images for this model are computed, which agree with the general trend seen in Subaru data. Decoupling between the spatial distributions of the {mu}m-sized dust and mm-sized dust inside the cavity is suggested by the model, which, if confirmed, necessitates a mechanism, such as dust filtration, for differentiating the small and big dust in the cavity clearing process. Our model also suggests an inwardly increasing gas-to-dust ratio in the inner disk, and different spatial distributions for the small dust inside and outside the cavity, echoing the predictions in grain coagulation and growth models.

Dong, R.; Rafikov, R.; Zhu, Z.; Brandt, T.; Janson, M. [Department of Astrophysical Sciences, Princeton University, Princeton, NJ 08544 (United States); Hartmann, L. [Department of Astronomy, University of Michigan, 500 Church Street, Ann Arbor, MI 48105 (United States); Whitney, B. [Astronomy Department, University of Wisconsin-Madison, 475 N. Charter Street, Madison, WI 53706 (United States); Muto, T. [Tokyo Institute of Technology, 2-12-1 Ookayama, Meguro, Tokyo 152-8551 (Japan); Hashimoto, J.; Kuzuhara, M. [National Astronomical Observatory of Japan, 2-21-1 Osawa, Mitaka, Tokyo 181-8588 (Japan); Grady, C. [ExoPlanets and Stellar Astrophysics Laboratory, Code 667, Goddard Space Flight Center, Greenbelt, MD 20771 (United States); Follette, K. [Steward Observatory, University of Arizona, 933 North Cherry Avenue, Tucson, AZ 85721 (United States); Tanii, R.; Itoh, Y. [Graduate School of Science, Kobe University, 1-1 Rokkodai, Nada-ku, Kobe 657-8501 (Japan); Thalmann, C. [Astronomical Institute 'Anton Pannekoek', University of Amsterdam, Science Park 904, 1098 XH Amsterdam (Netherlands); Wisniewski, J. [Department of Astronomy, University of Washington, Box 351580, Seattle, WA 98195 (United States); Mayama, S. [Graduate University for Advanced Studies (SOKENDAI), Shonan International Village, Hayama-cho, Miura-gun, Kanagawa 240-0193 (Japan); Abe, L. [Laboratoire Lagrange, UMR7293, Universite de Nice-Sophia Antipolis, CNRS, Observatoire de la Cote d'Azur, 06300 Nice (France); Brandner, W. [Max Planck Institute for Astronomy, Heidelberg (Germany); Carson, J., E-mail: rdong@astro.princeton.edu [Department of Physics and Astronomy, College of Charleston, 58 Coming Street, Charleston, SC 29424 (United States); and others

2012-05-10T23:59:59.000Z

260

REACTOR DEVELOPMENT PROGRAM PROGRESS REPORT  

SciTech Connect

Progress on reactor programs and in general engineering research and development programs is summarized. Research and development are reported on water-cooled reactors including EBWR and Borax-V, sodium-cooled reactors including ZPR-III, IV, and IX, Juggernaut, and EBR-I and II. Other work included a review of fast reactor technology, and studies on nuclear superheat, thermal and fast reactor safety, and reactor physics. Effort was also devoted to reactor materials and fuels development, heat engineering, separation processes and advanced reactor concepts. (J.R.D.)

1961-04-01T23:59:59.000Z

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.


261

Effect of changes in DOE pricing policies for enrichment and reprocessing on research reactor fuel cycle costs  

SciTech Connect

Fuel cycle costs with HEU and LEU fuels for the IAEA generic 10 MW reactor are updated to reflect the change in DOE pricing policy for enrichment services as of October 1985 and the published charges for LEU reprocessing services as of February 1986. The net effects are essentially no change in HEU fuel cycle costs and a reduction of about 8 to 10% in the fuel cycle costs for LEU silicide fuel.

Matos, J.E.; Freese, K.E.

1986-11-03T23:59:59.000Z

262

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

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

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

263

GenDocSum+MCLR: Generic document summarization based on maximum coverage and less redundancy  

Science Conference Proceedings (OSTI)

With the rapid growth of information on the Internet and electronic government recently, automatic multi-document summarization has become an important task. Multi-document summarization is an optimization problem requiring simultaneous optimization ... Keywords: Differential evolution algorithm, Generic document summarization, Less redundancy, Maximum coverage, Optimization model

Rasim M. Alguliev; Ramiz M. Aliguliyev; Makrufa S. Hajirahimova

2012-11-01T23:59:59.000Z

264

Generic integration of environmental decision support systems - state-of-the-art  

Science Conference Proceedings (OSTI)

Environmental Information Systems (EIS) and Environmental Decision Support Systems (EDSS) are major building blocks in environmental management and science today. They are used at all levels of public bodies (community, state, national and international ... Keywords: EU FP5, Environmental Decision Support Systems (EDSS), Environmental Information Systems (EIS), GIMMI, Generic integration, I-MARQ

Ralf Denzer

2005-10-01T23:59:59.000Z

265

Generic Qualification of the ABB Common Qualified PLC Based Platform for Safety-Related Applications  

Science Conference Proceedings (OSTI)

As its nuclear plants age, the electric power industry is focusing on the development of cost-effective replacement systems for obsolete instrumentation, control, and safety systems. This report describes the generic qualification a platform for safety-related applications that incorporates programmable logic controllers (PLCs), a technology with an excellent track record in non-nuclear applications for critical control and safety functions.

1999-08-18T23:59:59.000Z

266

A generic model for sustainability with process- and product-specific instances  

Science Conference Proceedings (OSTI)

Motivation: Software systems as we know them often have a economic purpose and/or fulfill human or social needs of their users. The economic purpose is analysed by economy itself; the latter goals are analysed in software engineering by user-centric ... Keywords: case study, definition, generic reference model, requirements, sustainability

Birgit Penzenstadler; Henning Femmer

2013-03-01T23:59:59.000Z

267

Accepted to Energy Policy, December 2011. A generic framework for the description and analysis of  

E-Print Network (OSTI)

Accepted to Energy Policy, December 2011. ERG/201104 A generic framework for the description policies. The framework can be employed to capture the evolution of energy security in an energy system associated with the system's constituent processes. Energy security policies are treated as flows

Hughes, Larry

268

A generic ontology-based approach for requirement analysis and its application in network management software  

Science Conference Proceedings (OSTI)

This paper describes a generic ontology-based approach that eases the requirement analysis (RA) work. The approach enables the user to derive method-specific RA tools for different applications. The derivation process is based on a unified framework ... Keywords: Acquisition Ontology, Knowledge Acquisition, Knowledge-based Systems, Methodology Ontology, Requirement Analysis

Chau-Young Ivan Lin; Cheng-Seen Ho

1999-01-01T23:59:59.000Z

269

A generic architecture for redesign of organizations triggered by changing environmental circumstances  

Science Conference Proceedings (OSTI)

Artificial Intelligence has contributed (formal) design models and software support tools to application areas such as architecture, engineering and software design. This paper explores the effectiveness of applying design models to the area of organization ... Keywords: Generic architecture, Multi-agent systems, Organizational change, Organizational redesign

Mark Hoogendoorn; Catholijn M. Jonker; Jan Treur

2011-05-01T23:59:59.000Z

270

The generic genetic algorithm incorporates with rough set theory - An application of the web services composition  

Science Conference Proceedings (OSTI)

Evolutionary computing (EC) techniques have been used traditionally used for solving challenging optimization problems. But the increase in data and information has reduced the performance capacity of the GA, but highlighted the cost of finding a solution ... Keywords: Generic, Genetic algorithm, Rough set, Web services, Web services composition

Wen-Yau Liang; Chun-Che Huang

2009-04-01T23:59:59.000Z

271

Universal Sound Absorption in Amorphous Solids: A Theory of Elastically Coupled Generic Blocks  

E-Print Network (OSTI)

Glasses are known to exhibit quantitative universalities at low temperatures, the most striking of which is the ultrasonic attenuation coefficient 1/Q. In this work we develop a theory of coupled generic blocks with a certain randomness property to show that universality emerges essentially due to the interactions between elastic blocks, regardless of their microscopic nature.

Dervis C. Vural; Anthony J. Leggett

2011-03-29T23:59:59.000Z

272

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

273

Slurry reactor design studies  

SciTech Connect

The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. (Bechtel Group, Inc., San Francisco, CA (USA)); Akgerman, A. (Texas A and M Univ., College Station, TX (USA)); Smith, J.M. (California Univ., Davis, CA (USA))

1990-06-01T23:59:59.000Z

274

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

275

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

276

NUCLEAR REACTOR FUEL SYSTEMS  

DOE Patents (OSTI)

Homogeneous reactor fuel solutions are reported which provide automatic recombination of radiolytic gases and exhibit large thermal expansion characteristics, thereby providing stability at high temperatures and enabling reactor operation without the necessity of apparatus to recombine gases formed by the radiolytic dissociation of water in the fuel and without the necessity of liquid fuel handling outside the reactor vessel except for recovery processes. The fuels consist of phosphoric acid and water solutions of enriched uranium, wherein the uranium is in either the hexavalent or tetravalent state.

Thamer, B.J.; Bidwell, R.M.; Hammond, R.P.

1959-09-15T23:59:59.000Z

277

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

278

CONTROL FOR NEUTRONIC REACTOR  

DOE Patents (OSTI)

S>A control rod operating device in a nuclear reactor of the type in which the control rod is gradually withdrawn from the reactor to a position desired during stable operation is described. The apparatus is comprised essentially of a stop member movable in the direction of withdrawal of the control rod, a follower on the control rod engageable with the stop and means urging the follower against the stop in the direction of withdrawal. A means responsive to disengagement of the follower from the stop is provided for actuating the control rod to return to the reactor shut-down position.

Lichtenberger, H.V.; Cameron, R.A.

1959-03-31T23:59:59.000Z

279

Fast Breeder Reactor studies  

Science Conference Proceedings (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

280

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

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

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 inline 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, J.E.

1991-12-31T23:59:59.000Z

282

Liquid hydrogen flow problems in Kiwi reactors  

DOE Green Energy (OSTI)

The Kiwi series of reactors were the first ones tested in the US Rover Program in the development of nuclear rocket engines for space propulsion. The early experiments with liquid hydrogen showed that parallel flow systems were prone to uneven flow distributions and violent fluctuations in pressure and flow that were capable of destroying a reactor core. Kiwi flow distribution problems were solved by using multiple feed lines into the nozzle cooling system and carefully balancing impedance among them. The violent pressure and flow fluctuations were eliminated after their cause was identified as resonance phenomena driven by the response to flow disturbances of heat transfer through a superheated hydrogen layer. Smooth flow operations were assured by rapidly bringing operating pressures beyond several times the critical pressure of hydrogen. After this initial rough start, solid core nuclear rocket engines successfully passed milestones of achievements during the remainder of the Rover program.

Thurston, R.S.

1992-09-01T23:59:59.000Z

283

Reactor Thermal-Hydraulics  

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

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

284

Reactor hot spot analysis  

SciTech Connect

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

Vilim, R.B.

1985-08-01T23:59:59.000Z

285

Molten metal reactors  

SciTech Connect

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

286

Compact power reactor  

DOE Patents (OSTI)

There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector.

Wetch, Joseph R. (Woodland Hills, CA); Dieckamp, Herman M. (Canoga Park, CA); Wilson, Lewis A. (Canoga Park, CA)

1978-01-01T23:59:59.000Z

287

NEUTRONIC REACTOR CONSTRUCTION AND OPERATION  

DOE Patents (OSTI)

A method is given for operating a nuclear reactor having a negative coefficient of reactivity to compensate for the change in reactor reactivity due to the burn-up of the xenon peak following start-up of the reactor. When it is desired to start up the reactor within less than 72 hours after shutdown, the temperature of the reactor is lowered prior to start-up, and then gradually raised after start-up.

West, J.M.; Weills, J.T.

1960-03-15T23:59:59.000Z

288

Reactor-specific spent fuel discharge projections, 1984 to 2020  

Science Conference Proceedings (OSTI)

The original spent fuel utility data base (SFDB) has been adjusted to produce agreement with the EIA nuclear energy generation forecast. The procedure developed allows the detail of the utility data base to remain intact, while the overall nuclear generation is changed to match any uniform nuclear generation forecast. This procedure adjusts the weight of the reactor discharges as reported on the SFDB and makes a minimal (less than 10%) change in the original discharge exposures in order to preserve discharges of an integral number of fuel assemblies. The procedure used in developing the reactor-specific spent fuel discharge projections, as well as the resulting data bases themselves, are described in detail in this report. Discussions of the procedure cover the following topics: a description of the data base; data base adjustment procedures; addition of generic power reactors; and accuracy of the data base adjustments. Reactor-specific discharge and storage requirements are presented. Annual and cumulative discharge projections are provided. Annual and cumulative requirements for additional storage are shown for the maximum at-reactor (AR) storage assumption, and for the maximum AR with transshipment assumption. These compare directly to the storage requirements from the utility-supplied data, as reported in the Spent Fuel Storage Requirements Report. The results presented in this report include: the disaggregated spent fuel discharge projections; and disaggregated projections of requirements for additional spent fuel storage capacity prior to 1998. Descriptions of the methodology and the results are included in this report. Details supporting the discussions in the main body of the report, including descriptions of the capacity and fuel discharge projections, are included. 3 refs., 6 figs., 12 tabs.

Heeb, C.M.; Libby, R.A.; Holter, G.M.

1985-04-01T23:59:59.000Z

289

Solar Thermal Reactor Materials Characterization  

DOE Green Energy (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

290

RAPHAEL: The European Union's (Very) High Temperature Reactor Technology Project  

SciTech Connect

Since the late 1990, the European Union (EU) was conducting work on High Temperature Reactors (HTR) confirming their high potential in terms of safety (inherent safety features), environmental impact (robust fuel with no significant radioactive release), sustainability (high efficiency, potential suitability for various fuel cycles), and economics (simplifications arising from safety features). In April 2005, the EU Commission has started a new 4-year Integrated Project on Very High Temperature Reactors (RAPHAEL: Reactor for Process Heat And Electricity) as part of its 6{sup th} Framework Programme. The European Commission and the 33 partners from industry, R and D organizations and academia finance the project together. After the successful performance of earlier HTR-related EU projects which included the recovery of some earlier German experience and the re-establishment of strategically important R and D capabilities in Europe, RAPHAEL focuses now on key technologies required for an industrial VHTR deployment, both specific to very high temperature and generic to all types of modular HTR with emphasis on combined process heat and electricity generation. Advanced technologies are explored in order to meet the performance challenges required for a VHTR (900-1000 deg C, up to 200 GWd/tHM). To facilitate the planned sharing of significant parts of RAPHAEL results with the signatories of the Generation IV International Forum (GIF) VHTR projects, RAPHAEL is structured in a similar way as the corresponding GIF VHTR projects. (authors)

Fuetterer, Michael A. [European Commission, Joint Research Centre, Institute for Energy, P.O. Box 2, NL-1755 ZG Petten (Netherlands); Besson, D.; Bogusch, E.; Carluec, B.; Hittner, D.; Verrier, D. [AREVA Framatome-ANP (France); Billot, Ph.; Phelip, M. [Commissariat a l'Energie Atomique (France); Buckthorpe, D. [NNC Ltd, Knutsford (United Kingdom); Casalta, S. [European Commission, DG RTD, Brussels (Belgium); Chauvet, V. [STEP, Paris (France); Van Heek, A. [Nuclear Research and Consultancy Group, Petten (Netherlands); Von Lensa, W. [Forschungszentrum Juelich (Germany); Pirson, J. [Tractebel Engineering, Brussels (Belgium); Scheuermann, W. [Institut fuer Kernenergetik, University of Stuttgart (Germany)

2006-07-01T23:59:59.000Z

291

Preliminary analysis of the postulated changes needed to achieve rail cask handling capabilities at selected light water reactors  

SciTech Connect

Reactor-specific railroad and crane information for all LWRs in the US was extracted from current sources of information. Based on this information, reactors were separated into two basic groups consisting of reactors with existing, usable rail cask capabilities and those without these capabilities. The latter group is the main focus of this study. The group of reactors without present rail cask handling capabilities was further separated into two subgroups consisting of reactors considered essentially incapable of handling a large rail cask of about 100 tons and reactors where postulated facility changes could result in rail cask handling capabilities. Based on a selected population of 127 reactors, the results of this assessment indicate that usable rail cask capabilities exist at 83 (65%) of the reactors. Twelve (27%) of the remaining 44 reactors are deemed incapable of handling a large rail cask without major changes, and 32 reactors are considered likely candidates for potentially achieving rail cask handling capabilities. In the latter group, facility changes were postulated that would conceptually enable these reactors to handle large rail casks. The estimated cost per plant of required facility changes varied widely from a high of about $35 million to a low of <$0.3 million. Only 11 of the 32 plants would require crane upgrades. Spur track and right-of-way costs would apparently vary widely among sites. These results are based on preliminary analyses using available generic cost data. They represent lower bound values that are useful for developing an initial assessment of the viability of the postulated changes on a system-wide basis, but are not intended to be absolute values for specific reactors or sites.

Konzek, G.J.

1986-02-01T23:59:59.000Z

292

Export possibilities for small nuclear reactors  

Science Conference Proceedings (OSTI)

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

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

2007-07-01T23:59:59.000Z

293

REACTOR GROUT THERMAL PROPERTIES  

DOE Green Energy (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

294

Buildings Energy Data Book: 1.5 Generic Fuel Quad and Comparison  

Buildings Energy Data Book (EERE)

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

295

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

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

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

296

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

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

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

297

A 125GeV Higgs Boson and Muon g-2 in More Generic Gauge Mediation  

E-Print Network (OSTI)

Recently, the ATLAS and CMS collaborations reported exciting hints of a Standard Model-like Higgs boson with a mass around 125GeV. A Higgs boson this heavy is difficult to realize in conventional models of gauge mediation. Here we revisit the lightest Higgs boson mass in "more generic gauge mediation," where the Higgs doublets mix with the messenger doublets. We show that a Higgs boson mass around 125GeV can be realized in more generic gauge mediation models, even for a relatively light gluino mass ~1TeV. We also show that the muon anomalous magnetic moment can be within 1sigma of the experimental value for these models, even when the Higgs boson is relatively heavy. We also discuss the LHC constraints and the prospects of discovery.

Evans, Jason L; Shirai, Satoshi; Yanagida, Tsutomu T

2012-01-01T23:59:59.000Z

298

B Reactor | Department of Energy  

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

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

299

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

SciTech Connect

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

300

Advanced Reactor Development and Technology - Nuclear Engineering...  

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

Capabilities Nuclear Systems Modeling and Design Analysis Reactor Physics and Fuel Cycle Analysis Nuclear Data Program Advanced Reactor Development Overview Advanced Fast Reactor...

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


301

Zero Power Reactor simulation | Argonne National Laboratory  

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

Zero Power Reactor simulation Share Description Ever wanted to see a nuclear reactor core in action? Here's a detailed simulation of the Zero Power Reactor experiment, run by...

302

THE MATERIALS OF FAST BREEDER REACTORS  

E-Print Network (OSTI)

jet aircraft engines, and nuclear reactor fuel elements. Ancomponents of a nuclear reactor core are susceptible tothe nuclear physics of the thermal and fast neutron reactors

Olander, Donald R.

2013-01-01T23:59:59.000Z

303

Generic Models and Model Validation for Wind and Solar PV Generation: Technical Update  

Science Conference Proceedings (OSTI)

The influx of variable generation technologies, particularly wind generation, into the bulk transmission grid in the U.S. and internationally has been significant over the past decade. This trend will most likely continue in light of national (in other countries) and state renewable portfolio standards. Thus, there is at present a need for generic, standard and publicly available models for variable generation technologies for the purpose of power system planning studies. EPRI has been a key participant ...

2011-12-20T23:59:59.000Z

304

Generic Job and Task Analysis: Handling Planned Outages and Hot Line Work  

Science Conference Proceedings (OSTI)

Switching safety and reliability are ongoing concerns in the utility industry. This report summarizes the results of a generic job and task analysis for planned switching and tagging involving the deenergization of electric utility lines and equipment and for the switching and tagging associated with hot line work. The results of this job and task analysis provide utilities with an opportunity to compare their own practices with practices employed by other utilities.

2002-08-30T23:59:59.000Z

305

Development of Power System Restoration Tool Based on Generic Restoration Milestones  

Science Conference Proceedings (OSTI)

The importance of power system restoration is well recognized by the power industry. System reliability depends heavily on the efficiency of system restoration. EPRI has been working on developing a prototype decision support tool, the System Restoration Navigator (SRN), for evaluating system restoration strategies, based on the concept of Generic Restoration Milestones (GRMs). This tool is intended to be used for offline planning at the current stage and later on for dispatcher training and online decis...

2010-12-22T23:59:59.000Z

306

Fire Events Database and Generic Ignition Frequency Model for U.S. Nuclear Power Plants  

Science Conference Proceedings (OSTI)

This report contains a revision of the EPRI Fire Events Database for U.S. Nuclear Power Plants last published in EPRI Report 1000894 in October 2000. This report also contains a revision of the generic fire ignition frequency models that were published in NSAC-178L, "Fire Events Database of U.S. Nuclear Power Plants" (January 1993) and EPRI TR-105929, "Fire Ignition Frequency Model at Shutdown for U.S. Nuclear Power Plants" (December 1995).

2001-11-09T23:59:59.000Z

307

Generic Letter 96-06 Waterhammer Issues Resolution: Technical Basis Report - Proprietary  

Science Conference Proceedings (OSTI)

The U.S. Nuclear Regulatory Commission (NRC) Generic Letter (GL) 96-06 identified potential issues for waterhammer effects during postulated events that can cause potential damage to service water systems. The User's Manual (Volume 1) provides methods recommended for evaluating the impact of potential waterhammer loads on plant service water system components. This Technical Basis Report (Volume 2) provides detailed background information and the technical basis for the methods defined in the User's Manual.

2002-04-30T23:59:59.000Z

308

Generic Letter 96-06 Waterhammer Issues Resolution: Technical Basis Report - Non Proprietary  

Science Conference Proceedings (OSTI)

The U.S. Nuclear Regulatory Commission (NRC) Generic Letter (GL) 96-06 identified potential issues for waterhammer effects during postulated events that can cause potential damage to service water systems. The User's Manual (Volume 1) provides methods recommended for evaluating the impact of potential waterhammer loads on plant service water system components. This Technical Basis Report (Volume 2) provides detailed background information and the technical basis for the methods defined in the User's Manual.

2002-04-30T23:59:59.000Z

309

Gas-cooled reactors  

SciTech Connect

Experience to date with operation of high-temperature gas-cooled reactors has been quite favorable. Despite problems in completion of construction and startup, three high-temperature gas-cooled reactor (HTGR) units have operated well. The Windscale Advanced Gas-Cooled Reactor (AGR) in the United Kingdom has had an excellent operating history, and initial operation of commercial AGRs shows them to be satisfactory. The latter reactors provide direct experience in scale-up from the Windscale experiment to fullscale commercial units. The Colorado Fort St. Vrain 330-MWe prototype helium-cooled HTGR is now in the approach-to-power phase while the 300-MWe Pebble Bed THTR prototype in the Federal Republic of Germany is scheduled for completion of construction by late 1978. THTR will be the first nuclear power plant which uses a dry cooling tower. Fuel reprocessing and refabrication have been developed in the laboratory and are now entering a pilot-plant scale development. Several commercial HTGR power station orders were placed in the U.S. prior to 1975 with similar plans for stations in the FRG. However, the combined effects of inflation, reduced electric power demand, regulatory uncertainties, and pricing problems led to cancellation of the 12 reactors which were in various stages of planning, design, and licensing.

Schulten, R.; Trauger, D.B.

1976-01-01T23:59:59.000Z

310

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

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

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

311

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

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

312

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

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

313

Damage monitoring of refractory wall in a generic entrained-bed slagging gasification system  

E-Print Network (OSTI)

fusion could enable such large emissions cuts. There is no reason, however, why one tool should have for nuclear weapons. These new reactors would add sev- eral thousand tons of plutonium to the world's current stock of reactor plutonium (roughly 1000 tons). 8. Fuel-Switching for Electricity Because of the lower

Ray, Asok

314

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

315

REACTOR CONTROL DEVICE  

DOE Patents (OSTI)

A wholly mechanical compact control device is designed for automatically rendering the core of a fission reactor subcritical in response to core temperatures in excess of the design operating temperature limit. The control device comprises an expansible bellows interposed between the base of a channel in a reactor core and the inner end of a fuel cylinder therein which is normally resiliently urged inwardly. The bellows contains a working fluid which undergoes a liquid to vapor phase change at a temperature substantially equal to the design temperature limit. Hence, the bellows abruptiy expands at this limiting temperature to force the fuel cylinder outward and render the core subcritical. The control device is particularly applicable to aircraft propulsion reactor service. (AEC)

Graham, R.H.

1962-09-01T23:59:59.000Z

316

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

317

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

318

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

319

Analytical modeling of core hydraulics and flow management in breeder reactors  

SciTech Connect

An analytical model representing the hydraulic behavior of the primary system of fast breeder nuclear reactors is discussed. A computer code capable of detailing the core flow distribution and characterizing the flow and pressure drop in each reactor component is presented. Application of this method to the reactor core thermal-hydraulic design has allowed optimization of the flow management with consequent upgrading in performance, reduction of unnecessary conservatism and very substantial cost savings. Typical quantitative examples are presented.

Carelli, M.D.; Willis, J.M.

1979-01-01T23:59:59.000Z

320

Prototyping a Generic, Unified Land Surface Classification and Screening Methodology for GPM-Era Microwave Land Precipitation Retrieval Algorithms  

Science Conference Proceedings (OSTI)

A prototype generic, unified land surface classification and screening methodology for Global Precipitation Measurement (GPM)-era microwave land precipitation retrieval algorithms by using ancillary datasets is developed. As an alternative to the ...

Arief Sudradjat; Nai-Yu Wang; Kaushik Gopalan; Ralph R. Ferraro

2011-06-01T23:59:59.000Z

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
to obtain the most current and comprehensive results.


321

Nuclear reactor apparatus  

DOE Patents (OSTI)

A lifting, rotating and sealing apparatus for nuclear reactors utilizing rotating plugs above the nuclear reactor core. This apparatus permits rotation of the plugs to provide under the plug refueling of a nuclear core. It also provides a means by which positive top core holddown can be utilized. Both of these operations are accomplished by means of the apparatus lifting the top core holddown structure off the nuclear core while stationary, and maintaining this structure in its elevated position during plug rotation. During both of these operations, the interface between the rotating member and its supporting member is sealingly maintained.

Wade, Elman E. (Ruffs Dale, PA)

1978-01-01T23:59:59.000Z

322

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

323

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

324

Perspectives on reactor safety  

SciTech Connect

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

325

Reactor Neutrino Experiments  

E-Print Network (OSTI)

Precisely measuring $\\theta_{13}$ is one of the highest priority in neutrino oscillation study. Reactor experiments can cleanly determine $\\theta_{13}$. Past reactor neutrino experiments are reviewed and status of next precision $\\theta_{13}$ experiments are presented. Daya Bay is designed to measure $\\sin^22\\theta_{13}$ to better than 0.01 and Double Chooz and RENO are designed to measure it to 0.02-0.03. All are heading to full operation in 2010. Recent improvements in neutrino moment measurement are also briefed.

Jun Cao

2007-12-06T23:59:59.000Z

326

Diagnostics for hybrid reactors  

SciTech Connect

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

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

2012-06-19T23:59:59.000Z

327

THERMAL NUCLEAR REACTOR  

DOE Patents (OSTI)

Nuclear reactors of the graphite moderated air cooled type in which canned slugs or rods of fissile material are employed are discussed. Such a reactor may be provided with a means for detecting dust particles in the exhausted air. The means employed are lengths of dust absorbent cord suspended in vertical holes in the shielding structure above each vertical coolant flow channel to hang in the path of the cooling air issuing from the channels, and associated spindles and drive motors for hauling the cords past detectors, such as Geiger counters, for inspecting the cords periodically. This design also enables detecting the individual channel in which a fault condition may have occurred.

Fenning, F.W.; Jackson, R.F.

1957-09-24T23:59:59.000Z

328

MEANS FOR SHIELDING REACTORS  

DOE Patents (OSTI)

A reactor of the heterageneous, heavy water moderated type is described. The reactor is comprised of a plurality of vertically disposed fuel element tubes extending through a tank of heavy water moderator and adapted to accommodate a flow of coolant water in contact with the fuel elements. A tank containing outgoing coolant water is disposed above the core to function is a radiation shield. Unsaturated liquid hydrocarbon is floated on top of the water in the shield tank to reduce to a minimum the possibility of the occurrence of explosive gaseous mixtures resulting from the neutron bombardment of the water in the shield tank.

Garrison, W.M.; McClinton, L.T.; Burton, M.

1959-03-10T23:59:59.000Z

329

TY Generic  

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

p Building energy simulation programs are effective tools for the evaluation of building energy saving and optimization of design The fact that large discrepancies exist in...

330

Equitable distribution  

Science Conference Proceedings (OSTI)

The problem of distributing available resources occurs in a great variety of networks, each with peculiarities of its own. Coal from mines has to be distributed to central dumps and to small yards. Ice cream must be distributed only to refrigerated stores ...

John A. Gosden

1963-05-01T23:59:59.000Z

331

Reactor sharing experience at the MIT research reactor  

SciTech Connect

This paper provides a number of examples of how educational institutions in the Boston area and elsewhere that do not possess nuclear reactors for training and research purposes have successfully enriched their programs through utilization of the Massachusetts Institute of Technology Research Reactor (MITR) with assistance from the Reactor Sharing Program of the US Department of Energy (DOE).

Clark, L. Jr.; Fecych, W.; Young, H.H.

1985-01-01T23:59:59.000Z

332

THE HOMOGENEOUS SUSPENSION REACTOR PROJECT  

SciTech Connect

The considerations which led to the study of a homogeneous suspension reactor are reviewed briefly. The characteristics of the KEMA Suspension Test Reactor (KSTR) are then summarized. (J.S.R.)

Went, J.J.

1963-02-01T23:59:59.000Z

333

Advanced Nuclear Reactors | Department of Energy  

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

in Sodium Reactor Subassemblies Sodium-cooled Fast Reactor (SFR) Coolant Flow At the heart of a nuclear power plant is the reactor. The fuel assembly is placed inside a reactor...

334

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

335

Reactor operation safety information document  

Science Conference Proceedings (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

336

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

DOE Green Energy (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

337

Distribution Screening for Distributed Generation  

Science Conference Proceedings (OSTI)

As the deployment of renewable distributed generation increases, the need for traditional energy providers to interact with these resources increases. Detailed modeling and simulation of the distribution and distributed resources is a critical element to better analyze, understand and predict these interactions. EPRI has developed a tool for such analysis called OpenDSS. In addition, as part of the renewable integration program an applet was created for screening distributed generation (DG). This report ...

2009-12-23T23:59:59.000Z

338

NEUTRONIC REACTOR FUEL ELEMENT  

DOE Patents (OSTI)

A reactor fuel element of the capillary tube type is described. The element consists of a thin walled tube, sealed at both ends, and having an interior coatlng of a fissionable material, such as uranium enriched in U-235. The tube wall is gas tight and is constructed of titanium, zirconium, or molybdenum.

Kesselring, K.A.; Seybolt, A.U.

1958-12-01T23:59:59.000Z

339

NEUTRONIC REACTOR FUEL ELEMENT  

DOE Patents (OSTI)

A fuel slug for a reactor which acts as a safety device is described. The fuel slug is an aluminum tube with a foil lining the inside surface of the tube, the foil being fabricated of uranium in a lead matrix.

Horning, W.A.; Lanning, D.D.; Donahue, D.J.

1959-10-01T23:59:59.000Z

340

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

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

Neutronic reactor thermal shield  

DOE Patents (OSTI)

1. The method of operating a water-cooled neutronic reactor having a graphite moderator which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40-60 volume percent of the mixture, in contact with the graphite moderator.

Wende, Charles W. J. (West Chester, PA)

1976-06-15T23:59:59.000Z

342

NEUTRONIC REACTOR FUEL PUMP  

DOE Patents (OSTI)

A reactor fuel pump is described which offers long life, low susceptibility to radiation damage, and gaseous fission product removal. An inert-gas lubricated bearing supports a journal on one end of the drive shsft. The other end has an impeller and expansion chamber which effect pumping and gas- liquid separation. (T.R.H.)

Cobb, W.G.

1959-06-01T23:59:59.000Z

343

The First Reactor  

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

Chicago Pile-1 (CP-1) was the world's first nuclear reactor. CP-1 was built on a rackets court, under the abandoned west stands of the original Alonzo Stagg Field stadium, at the University of...

344

Thermal Reactor Safety  

Science Conference Proceedings (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

345

WATER BOILER REACTOR  

DOE Patents (OSTI)

As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

King, L.D.P.

1960-11-22T23:59:59.000Z

346

Cermet fuel reactors  

Science Conference Proceedings (OSTI)

Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs.

Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

1987-09-01T23:59:59.000Z

347

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

348

Nuclear Reactors and Technology  

SciTech Connect

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

349

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

350

Reactor component automatic grapple  

DOE Patents (OSTI)

A grapple for handling nuclear reactor components in a medium such as liquid sodium which, upon proper seating and alignment of the grapple with the component as sensed by a mechanical logic integral to the grapple, automatically seizes the component. The mechanical logic system also precludes seizure in the absence of proper seating and alignment.

Greenaway, Paul R. (Bethel Park, PA)

1982-01-01T23:59:59.000Z

351

NUCLEAR REACTOR FUEL ELEMENT  

DOE Patents (OSTI)

A fuel plate is designed for incorporation into control rods of the type utilized in high-flux test reactors. The fuel plate is designed so that the portion nearest the poison section of the control rod contains about one-half as much fissionable material as in the rest of the plate, thereby eliminating dangerous flux peaking in that portion. (AEC)

Currier, E.L. Jr.; Nicklas, J.H.

1963-06-11T23:59:59.000Z

352

NEUTRONIC REACTOR STRUCTURE  

DOE Patents (OSTI)

A neutronic reactor is described. It has a core consisting of natural uranium and heavy water and having a K-factor greater than unity which is surrounded by a reflector consisting of natural uranium and ordinary water having a Kfactor less than unity.

Weinberg, A.M.; Vernon, H.C.

1961-05-30T23:59:59.000Z

353

NEUTRONIC REACTOR STRUCTURE  

DOE Patents (OSTI)

The neutronic reactor is comprised of a core consisting of natural uranium and heavy water with a K-factor greater than unity. The core is surrounded by a reflector consisting of natural uranium and ordinary water with a Kfactor less than unity. (AEC)

Vernon, H.C.; Weinberg, A.M.

1961-05-30T23:59:59.000Z

354

NEUTRONIC REACTOR SHIELD  

DOE Patents (OSTI)

The reactor radiation shield material is comprised of alternate layers of iron-containing material and compressed cellulosic material, such as masonite. The shielding material may be prefabricated in the form of blocks, which can be stacked together in ary desired fashion to form an effective shield.

Fermi, E.; Zinn, W.H.

1957-09-24T23:59:59.000Z

355

NEUTRONIC REACTOR CONTROL ELEMENT  

DOE Patents (OSTI)

A boron-10 containing reactor control element wherein the boron-10 is dispersed in a matrix material is describeri. The concentration of boron-10 in the matrix varies transversely across the element from a minimum at the surface to a maximum at the center of the element, prior to exposure to neutrons. (AEC)

Beaver, R.J.; Leitten, C.F. Jr.

1962-04-17T23:59:59.000Z

356

Qualification of In-Service Examination of the Yankee Rowe Reactor Pressure Vessel  

Science Conference Proceedings (OSTI)

An effective in-service examination of the reactor pressure vessel was an essential part of the restart program for the Yankee Atomic Power Company plant in Rowe, Massachusetts. This report describes development of an effective examination strategy, demonstration of performance of the examination procedures, and development of data on the distribution of flaws in reactor pressure vessels.

1993-01-01T23:59:59.000Z

357

REACTOR FUEL ELEMENTS TESTING CONTAINER  

DOE Patents (OSTI)

This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

Whitham, G.K.; Smith, R.R.

1963-01-15T23:59:59.000Z

358

EMERGENCY SHUTDOWN FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

An emergency shutdown or scram apparatus for use in a nuclear reactor that includes a neutron absorber suspended from a temperature responsive substance that is selected to fail at a preselected temperature in excess of the normal reactor operating temperature, whereby the neutron absorber is released and allowed to fall under gravity to a preselected position within the reactor core is presented. (AEC)

Paget, J.A.; Koutz, S.L.; Stone, R.S.; Stewart, H.B.

1963-12-24T23:59:59.000Z

359

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

Science Conference Proceedings (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

360

CHEMICAL PROBLEMS OF NON-AQUEOUS FLUID-FUEL REACTORS  

SciTech Connect

The three main chemical problems of non-aqueous fluidfuel reactors are selection of a fuel system which meets nuclear and thermal requirements, control of corrosion of structural materials, and development of an efficient and economical separation process. The problems of the fastneutron reactor with a core of fused NaCl, PbCl/sub 2/, UCl/sub 4/ mixture and a blanket of fused UCl/ sub 4/ are discussed, except for the separation process. Brief treatment is given the Bi- U - Pu fuel system for thermal reactors, including tentative flowsheets for the separation process. A detailed discussion of the experiments of Bareis at Brookhaven and related experiments on the distribution of various metals between liquid Bi and fused-salt solutions is given, the experimental results correlated, and application made to reactor problems. A general discussion is given of nonaqueous high-temperature separation processes (L.M.T.)

Scatchard, G.; Clark, H.M.; Golden, S.; Boltax, A.; Schuhmann, R. Jr.

1952-10-15T23:59:59.000Z

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

Modeling Chemical Reactors I: Quiescent Reactors  

E-Print Network (OSTI)

We introduce a fully generalized quiescent chemical reactor system in arbitrary space $\\vdim =1,2$ or 3, with $n\\in\\mathbb{N}$ chemical constituents $\\alpha_{i}$, where the character of the numerical solution is strongly determined by the relative scaling between the local reactivity of species $\\alpha_{i}$ and the local functional diffusivity $\\mathscr{D}_{ij}(\\alpha)$ of the reaction mixture. We develop an operator time-splitting predictor multi-corrector RK--LDG scheme, and utilize $hp$-adaptivity relying only on the entropy $\\mathscr{S}_{\\mathfrak{R}}$ of the reactive system $\\mathfrak{R}$. This condition preserves these bounded nonlinear entropy functionals as a necessarily enforced stability condition on the coupled system. We apply this scheme to a number of application problems in chemical kinetics; including a difficult classical problem arising in nonequilibrium thermodynamics known as the Belousov-Zhabotinskii reaction where we utilize a concentration-dependent diffusivity tensor $\\mathscr{D}_{ij}(...

Michoski, C E; Schmitz, P G

2010-01-01T23:59:59.000Z

362

Neutronic analysis of a fusion hybrid reactor  

SciTech Connect

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

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

2012-07-01T23:59:59.000Z

363

Survey of Optimization of Reactor Coolant Cleanup Systems: For Boiling Water Reactors and Pressurized Water Reactors  

Science Conference Proceedings (OSTI)

Optimization of the reactor coolant cleanup systems in the boiling water reactor (BWR) and pressurized water reactor (PWR) environment is important for controlling the transport of corrosion products (metals and activated metals), fission products, and coolant impurities (soluble and insoluble) throughout the reactor coolant loop, and this optimization contributes to reducing primary system radiation fields. The removal of radionuclides and corrosion products is just one of many functions (both ...

2013-09-27T23:59:59.000Z

364

On the Global Content PMI: Improved Copy-Protected Internet Content Distribution  

E-Print Network (OSTI)

packages online. If a cracker can figure out a generic attack against the copy-protection scheme usedOn the Global Content PMI: Improved Copy-Protected Internet Content Distribution Tadayoshi Kohno addresses a problem with copy-protecting a large collection of electronic content. The notion and severity

Kohno, Tadayoshi

365

Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report  

SciTech Connect

This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed.

Hauptman, H.M.; Petro, J.N.; Jacobi, O. [and others

1995-04-01T23:59:59.000Z

366

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

367

Prismatic modular reactor analysis with melcor  

E-Print Network (OSTI)

Hydrogen, a more sustainable source of energy, is a potential substitute for hydrocarbon fuel for power generation. The Very High Temperature gas-cooled Reactor (VHTR) concept can produce hydrogen with high efficiency and in large quantities. The US Department of Energy plans to build a VHTR as a next-generation hydrogen/electricity production plant. This reactor concept is very different from that of commercial reactors in the US. In order to acquire licensing eligibility for VHTRs, analysis tools need to be validated and applied to design and evaluate VHTRs under operation conditions and accident scenarios. In this thesis, MELCOR, a severe accident code, was used to analyze one of the VHTR designs – a prismatic core Next Generation Nuclear Plant (NGNP). The NGNP is based on General Atomics‘ (GA) Gas Turbine – Modular Helium Reactor (GT-MHR) 600 MW design. According to the current literature survey, more data is available for the GT-MHR than for the NGNP. Therefore, for the purposes of extending MELCOR capabilities and code validation, a model of the GT-MHR reactor pressure vessel (RPV) was developed. Based on the currently available data, a model of the NGNP RPV was then developed through modifying the GT-MHR RPV model. For both RPV models, coolant outlet temperature under normal operating conditions corresponds well to the data from literature. The reactor cavity cooling systems (RCCS), which passively removes heat from the RPV wall to the outside atmosphere, was then added to this GT-MHR RPV model. With this model addition, the heat removal rate of the RCCS under normal operating conditions was calculated to correspond well to the data from references. Pressurized conduction cooldown (PCC), one of the important postulated accident scenarios for a prismatic core reactor, was simulated with the complete model. MELCOR has been demonstrated to have the ability of modeling a prismatic core VHTR. The calculated outlet temperature and mass flow rate under normal operation correspond well to references. However, the calculation for the heat distribution in the graphite and fuel is unsatisfactory which requires MELCOR modification for the PCC simulation. For future work, a complete model of the NGNP under normal operation conditions will be developed when additional data becomes available.

Zhen, Ni

2008-12-01T23:59:59.000Z

368

Distribution Workshop  

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

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

369

Trace anomaly and Hawking effect in generic 2D dilaton gravity theories  

E-Print Network (OSTI)

Black hole solutions in the context of a generic matter-coupled two-dimensional dilaton gravity theory are discussed both at the classical and semiclassical level. Starting from general assumptions, a criterion for the existence of black holes is given. The relationship between conformal anomaly and Hawking radiation is extended to a broad class of two-dimensional dilaton gravity models. A general and simple formula relating the magnitude of the Hawking effect to the dilaton potential evaluated on the horizon is derived.

Mariano Cadoni

1995-10-09T23:59:59.000Z

370

Reviews of ASME Section 11 pump and valve relief requests: Post Generic Letter 89-04  

SciTech Connect

This paper presents a discussion of ASME Section 11 Pump and Valve Inservice Testing relief request reviews by the NRC and their contractors. Topics that will be discussed include the scope of USNRC reviews in Technical Evaluation Reports (TERs) (and Safety Evaluation, SEs); including the basis for granting relief requests, the status of relief requests in IST Program updates, and the Generic Letter 89-04 approval process; and the level of technical detail required in submitted programs. This presentation is based on the experiences of Brookhaven National Laboratory in reviewing IST Programs for the Mechanical Engineering Branch of the US Nuclear Regulatory Commission.

DiBiasio, A.

1992-07-01T23:59:59.000Z

371

Reviews of ASME Section 11 pump and valve relief requests: Post Generic Letter 89-04  

SciTech Connect

This paper presents a discussion of ASME Section 11 Pump and Valve Inservice Testing relief request reviews by the NRC and their contractors. Topics that will be discussed include the scope of USNRC reviews in Technical Evaluation Reports (TERs) (and Safety Evaluation, SEs); including the basis for granting relief requests, the status of relief requests in IST Program updates, and the Generic Letter 89-04 approval process; and the level of technical detail required in submitted programs. This presentation is based on the experiences of Brookhaven National Laboratory in reviewing IST Programs for the Mechanical Engineering Branch of the US Nuclear Regulatory Commission.

DiBiasio, A.

1992-01-01T23:59:59.000Z

372

Neutronic safety parameters and transient analyses for Poland's MARIA research reactor.  

SciTech Connect

Reactor kinetic parameters, reactivity feedback coefficients, and control rod reactivity worths have been calculated for the MARIA Research Reactor (Swierk, Poland) for M6-type fuel assemblies with {sup 235}U enrichments of 80% and 19.7%. Kinetic parameters were evaluated for family-dependent effective delayed neutron fractions, decay constants, and prompt neutron lifetimes and neutron generation times. Reactivity feedback coefficients were determined for fuel Doppler coefficients, coolant (H{sub 2}O) void and temperature coefficients, and for in-core and ex-core beryllium temperature coefficients. Total and differential control rod worths and safety rod worths were calculated for each fuel type. These parameters were used to calculate generic transients for fast and slow reactivity insertions with both HEU and LEU fuels. The analyses show that the HEU and LEU cores have very similar responses to these transients.

Bretscher, M. M.; Hanan, N. A.; Matos, J. E.; Andrzejewski, K.; Kulikowska, T.

1999-09-27T23:59:59.000Z

373

Test storage of spent reactor fuel in the Climax granite at the Nevada Test Site  

SciTech Connect

A test of retrievable dry geologic storage of spent fuel assemblies from an operating commercial nuclear reactor is underway at the Nevada Test Site. This generic test is located 420 m below the surface in the Climax granitic stock. Eleven canisters of spent fuel approximately 2.3 years out of reactor core (about 2 kW/canister thermal output) will be emplaced in a storage drift along with 6 electrical simulator canisters and their effects will be compared. Two adjacent drifts will contain electrical heaters, which will be operated to simulate within the test array the thermal field of a large repository. The test objectives, technical concepts and rationale, and details of the test are stated and discussed.

Ramspott, L.D.; Ballou, L.B.

1980-02-13T23:59:59.000Z

374

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

375

OXIDATIVE COUPLING OF METHANE USING INORGANIC MEMBRANE REACTORS  

Science Conference Proceedings (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

376

Effect of radial power distribution on MITR-II fuel element and control blade worth  

Science Conference Proceedings (OSTI)

This summary has been prepared to document and discuss several effects on fuel element and control blade reactivity worths that occur due to changes in the radial power distribution of the 5-MW(t) Massachusetts Institute of Technology Research Reactor (MITR-II). These results were obtained from reactor physics measurements and core power distribution calculations made during the past decade.

Bernard, J.A.; Kwok, K.S.; Lanning, D.D.; Clark, L.L. Jr.

1985-01-01T23:59:59.000Z

377

CONTROL ROD FOR A NUCLEAR REACTOR AND METHOD OF PREPARATION  

DOE Patents (OSTI)

BS>An improved control rod is presented for a nuclear reactor. This control rod is comprised of a rare earth metal oxide or rare earth metal carbide such as gadolinium oxide or gadolinium carbide, uniformly distributed in a metal matrix having a low cross sectional area of absorption for thermal neutrons, such as aluminum, beryllium, and zirconium.

Hausner, H.H.

1958-12-30T23:59:59.000Z

378

HOMOGENEOUS NUCLEAR REACTOR  

SciTech Connect

This homogeneous reactor comprises a core occupied by a solution of a fissile material in a moderator liquid and a breeder region enclosing the core and having a suspension of fertile material in the same moderator liquid. There is communication between the core and breeder to allow mass transfer and pressure equalization between the regions. The zones each have a separate circuit for removing heat by a mixer chamber situated inside the reactor vessel. The effluents coming from the two regions are mixed and led to a common device for separation into a clear solution and suspension, which are each led back to its corresponding circuit. To control the relative concentration of the two regions, an evaporator is provided separating a part of the moderator liquid from the solution occupying the core, the condensed separated moderator liquid being led into the breeder region. (NPO)

1960-07-11T23:59:59.000Z

379

AIR COOLED NEUTRONIC REACTOR  

DOE Patents (OSTI)

A nuclear reactor of the air-cooled, graphite moderated type is described. The active core consists of a cubicle mass of graphite, approximately 25 feet in each dimension, having horizontal channels of square cross section extending between two of the opposite faces, a plurality of cylindrical uranium slugs disposed in end to end abutting relationship within said channels providing a space in the channels through which air may be circulated, and a cadmium control rod extending within a channel provided in the moderator. Suitable shielding is provlded around the core, as are also provided a fuel element loading and discharge means, and a means to circulate air through the coolant channels through the fuel charels to cool the reactor.

Fermi, E.; Szilard, L.

1958-05-27T23:59:59.000Z

380

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

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

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

382

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

383

HOMOGENEOUS NUCLEAR REACTOR  

DOE Patents (OSTI)

Nuclear reactors of the homogeneous liquid fuel type are discussed. The reactor is comprised of an elongated closed vessel, vertically oriented, having a critical region at the bottom, a lower chimney structure extending from the critical region vertically upwardly and surrounded by heat exchanger coils, to a baffle region above which is located an upper chimney structure containing a catalyst functioning to recombine radiolyticallydissociated moderator gages. In operation the liquid fuel circulates solely by convection from the critical region upwardly through the lower chimney and then downwardly through the heat exchanger to return to the critical region. The gases formed by radiolytic- dissociation of the moderator are carried upwardly with the circulating liquid fuel and past the baffle into the region of the upper chimney where they are recombined by the catalyst and condensed, thence returning through the heat exchanger to the critical region.

Hammond, R.P.; Busey, H.M.

1959-02-17T23:59:59.000Z

384

Water reactor fuel cladding  

Science Conference Proceedings (OSTI)

This patent describes a nuclear reactor fuel element cladding tube. It comprises: an outer cylindrical layer of a first zirconium alloy selected from the group consisting of Zircaloy-2 and Zircaloy-4; an inner cylindrical layer of a second zirconium alloy consisting essentially of about 0.19 to 0.6 wt.% tin, about 0.19 to less than 0.5 wt.% iron, about 100 to 700 ppm oxygen, less than 2000 ppm total impurities, and the remainder essentially zirconium; the inner layer characterized by aqueous corrosion resistance substantially the same as the first zirconium alloy; the inner layer characterized by improved resistance to PCI crack propagation under reactor operating conditions compared to the first zirconium alloy and substantially the same PCI crack propagation resistance compared to unalloyed zirconium; and the inner cylindrical layer is metallurgically bonded to the outer layer.

Foster, J.P.; McDonald, S.G.

1990-06-12T23:59:59.000Z

385

Power Reactor Decommissioning Experience  

Science Conference Proceedings (OSTI)

During the past two decades the NRC regulated nuclear industry has encountered and dealt with a diverse range of political, financial and technological challenges while decommissioning its nuclear facilities. During that time, the decommissioning of nuclear facilities has evolved into a mature industry in the United States with a number of large power reactors successfully decommissioned and their NRC licenses terminated. One of the challenges discussed in this report is site release standards, required ...

2011-07-08T23:59:59.000Z

386

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

387

Exhaust gas reactor  

SciTech Connect

A reactor for the oxidation of unburned and partially burned components in the exhaust gas from an internal combustion engine comprising a chamber which is substantially circular in cross sections perpendicular to its axis, one or more inlet pipes which pass a mixture of exhaust gas and air substantially tangentially into the chamber near to one end thereof, and an outlet pipe near to the other end of the chamber and which is so arranged that exhaust gas leaves the chamber substantially tangentially. The tangential inlet and tangential outlet of gases minimizes energy losses in the gas passing through the reactor. The ratio of the cross-sectional areas of the inlet pipe(s) to reactor chamber is preferably from 1:9 to 25:36, and similar ranges of crosssectional area ratios are preferred for the outlet pipe and chamber. The ratio of the length of the reaction chamber to diameter is preferably from 1:1 to 4:1. The chamber may be cylindrical or divergent from inlet end to outlet end and may be thermally insulated.

Camarsa, M.; Cocchiara, F.; Garcea, G.P.

1981-11-24T23:59:59.000Z

388

BOILER-SUPERHEATED REACTOR  

DOE Patents (OSTI)

A nuclear power reactor of the type in which a liquid moderator-coolant is transformed by nuclear heating into a vapor that may be used to drive a turbo- generator is described. The core of this reactor comprises a plurality of freely suspended tubular fuel elements, called fuel element trains, within which nonboiling pressurized liquid moderator-coolant is preheated and sprayed through orifices in the walls of the trains against the outer walls thereof to be converted into vapor. Passage of the vapor ovcr other unwetted portions of the outside of the fuel elements causes the steam to be superheated. The moderatorcoolant within the fuel elements remains in the liqUid state, and that between the fuel elements remains substantiaily in the vapor state. A unique liquid neutron-absorber control system is used. Advantages expected from the reactor design include reduced fuel element failure, increased stability of operation, direct response to power demand, and circulation of a minimum amount of liquid moderatorcoolant. (A.G.W.)

Heckman, T.P.

1961-05-01T23:59:59.000Z

389

HIGH STRENGTH CONTROL RODS FOR NEUTRONIC REACTORS  

DOE Patents (OSTI)

Nuclear reactor control rods comprised of highly compressed and sintered finely divided metal alloy panticles and fine metal oxide panticles substantially uniformly distributed theretbrough are described. The metal alloy consists essentially of silver, indium, cadmium, tin, and aluminum, the amount of each being present in centain percentages by weight. The oxide particles are metal oxides of the metal alloy composition, the amount of oxygen being present in certain percentages by weight and all the oxygen present being substantially in the form of metal oxide. This control rod is characterized by its high strength and resistance to creep at elevated temperatures.

Lustman, B.; Losco, E.F.; Cohen, I.

1961-07-11T23:59:59.000Z

390

Modeling Chemical Reactors I: Quiescent Reactors  

E-Print Network (OSTI)

We introduce a fully generalized quiescent chemical reactor system in arbitrary space $\\vdim =1,2$ or 3, with $n\\in\\mathbb{N}$ chemical constituents $\\alpha_{i}$, where the character of the numerical solution is strongly determined by the relative scaling between the local reactivity of species $\\alpha_{i}$ and the local functional diffusivity $\\mathscr{D}_{ij}(\\alpha)$ of the reaction mixture. We develop an operator time-splitting predictor multi-corrector RK--LDG scheme, and utilize $hp$-adaptivity relying only on the entropy $\\mathscr{S}_{\\mathfrak{R}}$ of the reactive system $\\mathfrak{R}$. This condition preserves these bounded nonlinear entropy functionals as a necessarily enforced stability condition on the coupled system. We apply this scheme to a number of application problems in chemical kinetics; including a difficult classical problem arising in nonequilibrium thermodynamics known as the Belousov-Zhabotinskii reaction where we utilize a concentration-dependent diffusivity tensor $\\mathscr{D}_{ij}(\\alpha)$, in addition to solving a simple equilibrium problem in order to evaluate the numerical error behavior.

C. E. Michoski; J. A. Evans; P. G. Schmitz

2010-12-28T23:59:59.000Z

391

Methane coupling by membrane reactor. Quarterly technical progress report, June 25, 1995--September 24, 1995  

DOE Green Energy (OSTI)

The research focus, for the development of a radial flow catalytic membrane reactor for studying methane coupling, concentrated on understanding the effects of pore size, catalyst loading and catalyst distribution inside the membrane, on the hydrocarbon selectivity. The experimental results from the catalytic studies of oxidative coupling of methane in the radial flow membrane reactors are summarized in this report and the reactor performances of three radial flow reactors with pore sizes of 0.02 {mu}m are interpreted by the energy dispersive X-ray digital mapping (EDX) technique. A conventional fixed bed catalytic reactor was set up and run under the same conditions as were used for the radial flow reactor studies. Two sources of samarium oxide catalysts were used in these experiments and their performances were compared in terms of C{sub 2} selectivity at same methane conversion. Compared to the monoclinic form of samarium oxide, the catalyst with cubic form was more active and selective for methane oxidative coupling. A general reactor model for oxidation reactions with distributed feed of oxygen and product removal strategies was proposed. Six types of feeding modes were analyzed and their feed distributions were optimized such that the desired product yields at the reactor outlet were maximized.

NONE

1995-12-12T23:59:59.000Z

392

Design guide for category V reactors transient reactors  

SciTech Connect

The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirements of Category V reactor structures, components, and systems.

Brynda, W J; Karol, R C; Lobner, P R; Powell, R W; Straker, E A

1979-03-01T23:59:59.000Z

393

Solid oxide electrochemical reactor science.  

DOE Green Energy (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

394

Reactor technology: power conversion systems and reactor operation and maintenance  

SciTech Connect

The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He/sup 3/ reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored.

Powell, J.R.

1977-01-01T23:59:59.000Z

395

Nuclear Reactor Severe Accident Experiments  

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

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

396

Designing Reactors to Facilitate Decommissioning  

SciTech Connect

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

397

Unclassified Distribution  

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

63 1 Unclassified Distribution UNIVERSITY :OF CALIFORNU Radiation Lab oratory Contract No, W-7405-eng-48 THE DETECTION OF U T I F I C I B L L Y PRODUCED WOTOMESONS WITH COUNTERS *...

398

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

SciTech Connect

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

399

Acceptability of reactors in space  

SciTech Connect

Reactors are the key to our future expansion into space. However, there has been some confusion in the public as to whether they are a safe and acceptable technology for use in space. The answer to these questions is explored. The US position is that when reactors are the preferred technical choice, that they can be used safely. In fact, it does not appear that reactors add measurably to the risk associated with the Space Transportation System.

Buden, D.

1981-01-01T23:59:59.000Z

400

Acceptability of reactors in space  

SciTech Connect

Reactors are the key to our future expansion into space. However, there has been some confusion in the public as to whether they are a safe and acceptable technology for use in space. The answer to these questions is explored. The US position is that when reactors are the preferred technical choice, that they can be used safely. In fact, it dies not appear that reactors add measurably to the risk associated with the Space Transportation System.

Buden, D.

1981-04-01T23:59:59.000Z

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

Special Distribution  

Office of Legacy Management (LM)

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

402

Grid integrated distributed PV (GridPV).  

SciTech Connect

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

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

2013-08-01T23:59:59.000Z

403

Reactor operations informal report, October 1994  

Science Conference Proceedings (OSTI)

This monthly progress report is divided into two parts. Part one covers the Brookhaven Medical Research Reactor and part two covers the Brookhaven High Flux Beam Reactor. Information is given for each reactor covering the following areas: reactor operation; instrumentation; mechanical maintenance; occurrence reports; and reactor safety.

Hauptman, H.M.; Petro, J.N.; Jacobi, O.; Lettieri, V.; Holden, N.; Ports, D.; Petricek, R.

1994-10-01T23:59:59.000Z

404

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 EIA'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.

Information Center

2011-05-10T23:59:59.000Z

405

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

406

NEUTRONIC REACTOR CHARGING AND DISCHARGING  

DOE Patents (OSTI)

A method and arrangement is presented for removing a fuel element from a neutronic reactor tube through which a liquid coolant is being circulaled. The fuel element is moved into a section of the tube beyond the reactor proper, and then the coolant in the tube between the fuel element and the reactor proper is frozen, so that the fuel element may be removed from the tube without loss of the coolant therein. The method is particularly useful in the case of a liquid metal- cooled reactor.

Zinn, W.H.

1959-07-14T23:59:59.000Z

407

Reactor & Nuclear Systems Publications | ORNL  

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

Publications and Reports NSED Monthly Reports Reactor and Nuclear Systems Publications 2013 Publications 2012 Publications 2011 Publications 2010 and Older Publications Nuclear...

408

Mirror Advanced Reactor Study (MARS)  

DOE Green Energy (OSTI)

Progress in a two year study of a 1200 MWe commercial tandem mirror reactor (MARS - Mirror Advanced Reactor Study) has reached the point where major reactor system technologies are identified. New design features of the magnets, blankets, plug heating systems and direct converter are described. With the innovation of radial drift pumping to maintain low plug density, reactor recirculating power fraction is reduced to 20%. Dominance of radial ion and impurity losses into the halo permits gridless, circular direct converters to be dramatically reduced in size. Comparisons of MARS with the Starfire tokamak design are made.

Logan, B.G.

1983-03-28T23:59:59.000Z

409

Transport Reactor Facility  

SciTech Connect

The Morgantown Energy Technology Center (METC) is currently evaluating hot gas desulfurization (HGD)in its on-site transport reactor facility (TRF). This facility was originally constructed in the early 1980s to explore advanced gasification processes with an entrained reactor, and has recently been modified to incorporate a transport riser reactor. The TRF supports Integrated Gasification Combined Cycle (IGCC) power systems, one of METC`s advanced power generation systems. The HGD subsystem is a key developmental item in reducing the cost and increasing the efficiency of the IGCC concept. The TRF is a unique facility with high-temperature, high-pressure, and multiple reactant gas composition capability. The TRF can be configured for reacting a single flow pass of gas and solids using a variety of gases. The gas input system allows six different gas inputs to be mixed and heated before entering the reaction zones. Current configurations allow the use of air, carbon dioxide, carbon monoxide, hydrogen, hydrogen sulfide, methane, nitrogen, oxygen, steam, or any mixture of these gases. Construction plans include the addition of a coal gas input line. This line will bring hot coal gas from the existing Fluidized-Bed Gasifier (FBG) via the Modular Gas Cleanup Rig (MGCR) after filtering out particulates with ceramic candle filters. Solids can be fed either by a rotary pocket feeder or a screw feeder. Particle sizes may range from 70 to 150 micrometers. Both feeders have a hopper that can hold enough solid for fairly lengthy tests at the higher feed rates, thus eliminating the need for lockhopper transfers during operation.

Berry, D.A.; Shoemaker, S.A.

1996-12-31T23:59:59.000Z

410

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

411

FOOD IRRADIATION REACTOR  

DOE Patents (OSTI)

An irradiation apparatus is described. It comprises a pressure vessel, a neutronic reactor active portion having a substantially greater height than diameter in the pressure vessel, an annular tank surrounding and spaced from the pressure vessel containing an aqueous indium/sup 1//sup 1//sup 5/ sulfate solution of approximately 600 grams per liter concentration, means for circulating separate coolants through the active portion and the space between the annular tank and the pressure vessel, radiator means adapted to receive the materials to be irradiated, and means for flowing the indium/sup 1//sup 1//sup 5/ sulfate solution through the radiator means.

Leyse, C.F.; Putnam, G.E.

1961-05-01T23:59:59.000Z

412

TRIGA reactor operating experience  

SciTech Connect

The Oregon State TRIGA Reactor (OSTR) has been in operation 3 years. Last August it was upgraded from 250 kW to 1000 kW. This was accomplished with little difficulty. During the 3 years of operation no major problems have been experienced. Most of the problems have been minor in nature and easily corrected. They came from lazy susan (dry bearing), Westronics Recorder (dead spots in the range), The Reg Rod Magnet Lead-in Circuit (a new type lead-in wire that does not require the lead-in cord to coil during rod withdrawal hss been delivered, much better than the original) and other small corrections.

Anderson, T.V. [Oregon State University, Corvallis, OR (United States)

1970-07-01T23:59:59.000Z

413

NEUTRONIC REACTOR CONSTRUCTION  

DOE Patents (OSTI)

A cover device is described for the fuel element receiving tube of a neutronic reactor of the heterogeneous, water cooled type wherein said tubes are arranged in a moderator with their longitudinal axes vertical. The cover is provided with means to support a rod-type fuel element from the bottom thereof and means to lock the cover in place, the latter being adapted for remote operation. This cover device is easily removable and seals the opening in the upper end of the fuel tube against leakage of coolant.

Vernon, H.C.; Goett, J.J.

1958-09-01T23:59:59.000Z

414

High flux reactor  

DOE Patents (OSTI)

A high flux reactor is comprised of a core which is divided into two symetric segments housed in a pressure vessel. The core segments include at least one radial fuel plate. The spacing between the plates functions as a coolant flow channel. The core segments are spaced axially apart such that a coolant mixing plenum is formed between them. A channel is provided such that a portion of the coolant bypasses the first core section and goes directly into the mixing plenum. The outlet coolant from the first core segment is mixed with the bypass coolant resulting in a lower inlet temperature to the lower core segment.

Lake, James A. (Idaho Falls, ID); Heath, Russell L. (Idaho Falls, ID); Liebenthal, John L. (Idaho Falls, ID); DeBoisblanc, Deslonde R. (Summit, NJ); Leyse, Carl F. (Idaho Falls, ID); Parsons, Kent (Idaho Falls, ID); Ryskamp, John M. (Idaho Falls, ID); Wadkins, Robert P. (Idaho Falls, ID); Harker, Yale D. (Idaho Falls, ID); Fillmore, Gary N. (Idaho Falls, ID); Oh, Chang H. (Idaho Falls, ID)

1988-01-01T23:59:59.000Z

415

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

416

Reactor refueling containment system  

DOE Patents (OSTI)

This report describes 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, J.E.; Meuschke, R.E.

1992-12-31T23:59:59.000Z

417

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

418

FAST NEUTRONIC REACTOR  

DOE Patents (OSTI)

This patent relates to a reactor and process for carrying out a controlled fast neutron chain reaction. A cubical reactive mass, weighing at least 920 metric tons, of uranium metal containing predominantly U/sup 238/ and having a U/sup 235/ content of at least 7.63% is assembled and the maximum neutron reproduction ratio is limited to not substantially over 1.01 by insertion and removal of a varying amount of boron, the reactive mass being substantially freed of moderator.

Snell, A.H.

1957-12-01T23:59:59.000Z

419

FUEL ASSAY REACTOR  

DOE Patents (OSTI)

A reactor having maximum sensitivity to perturbations is described comprising a core consisting of a horizontally disposed, rectangular, annular fuel zone containing enriched uranium dioxide dispersed in graphite, the concentration of uranium dioxide increasing from the outside to the inside of the fuel zone, an internal reflector of graphite containing an axial test opening disposed within the fuel zone, an external graphite reflector, means for changing the neutron spectrum in the test opening, and means for measuring perturbations in the neutron flux caused by the introduction of different fuel elements into the test opening. (AEC)

Spinrad, B.I.; Sandmeier, H.A.; Martens, F.H.

1962-12-25T23:59:59.000Z

420

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. 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.

Scott, Charles D. (Oak Ridge, TN)

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.


421

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

Scott, Charles D. (Oak Ridge, TN); Marasco, Joseph A. (Kingston, TN)

1996-01-01T23:59:59.000Z

422

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.

Scott, Charles D. (Oak Ridge, TN); Marasco, Joseph A. (Kingston, TN)

1995-01-01T23:59:59.000Z

423

MHD equilibrium properties of tokamak fusion reactor designs  

SciTech Connect

The equilibrium properties of several Tokamak Reactor Designs are analyzed and compared for varying pressure and current profiles using the Princeton Equilibrium Code. It is found that the UWMAK configuration has a broader range of equilibria than the Princeton Reference Design configuration, but that the safety factor on axis is less than unity for peaked current distributions. The Argonne Experimental Power Reactor has a satisfactory range of equilibria, but a means of limiting or diverting the plasma has not yet been proposed, and this may substantially change the results obtained. (auth)

Todd, A. M.M.; Gralnick, S. L.; Dalhed, H. E.

1976-01-01T23:59:59.000Z

424

Austin, TexasPARAMETRIC STUDY OF COMPONENT SELECTION AND OPERATION ON GENERIC DRAIN-BACK SOLAR WATER HEATER CERTIFICATION  

E-Print Network (OSTI)

I. Early experimental results of a research program to investigate the practicality of replacing the Solar Rating and Certification Corporation four day rating procedure with a more cost effective rating method for generic solar domestic hot water systems are reported. The generic systems experimental procedure and facility are described. Results from four experiments give an initial indication that the recirculation flow rate has little effect on system rating. Further experiments are required to assess the feasibility of replacing the current rating method with a simulated rating. 1.

W. T. Carlson; J. H. Davidson; W. S. Duff; D. M. Leone

1990-01-01T23:59:59.000Z

425

Reactor closure design for a pool-type fast reactor  

SciTech Connect

The reactor closure is the topmost structural part of a reactor module. For a pool-type fast reactor it is an especially important structure because it provides the interface between the primary coolant system and the main access area above the closure. The reactor closure comprises a stationary deck, a rotatable plug, the boundary elements of primary system and containment penetrations for equipment and auxiliary systems. This paper evaluates two different reactor closure design concepts, referred to as ''warm'' deck and ''hot'' deck, for a pool-type fast reactor with respect to their design features, technical merits, and economic benefits. The evaluation also includes functional, structural, and thermal analyses of the two deck design concepts. Issues related to their fabrication and shipping to the plant site are also addressed. The warm deck is a thick solid steel plate with under-the-deck insulation consisting of many layers of steel plates. The hot deck is a box-type structure consisting of a bottom plate reinforced with vertical ribs and cylinders. For insulation and radiation shielding, the region of the hot deck above the bottom plate is filled with steel balls. Conventional insulation is added on the top to further reduce heat loss into area above the deck. The design choice of the closure deck is strongly dependent on design features of the reactor; especially on the reactor module support. While the warm deck is preferable with the top support, the hot deck is better suited for the bottom support design of the module.

Chung, H.; Seidensticker, R.W.; Kann, W.J.; Bump, T.R.; Schatmeier, C.

1986-01-01T23:59:59.000Z

426

Licensed reactor nuclear safety criteria applicable to DOE reactors  

SciTech Connect

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

427

Fast Reactor Fuel Type and Reactor Safety Performance  

Science Conference Proceedings (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

428

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

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

429

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

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

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

430

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

Science Conference Proceedings (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

431

Solar energy as an alternate energy source to mixed oxide fuels in light-water cooled reactors  

DOE Green Energy (OSTI)

Supplemental information pertaining to the generic environmental impact statement on the Pu recycling process for mixed oxide light-water cooled reactors (GESMO) was requested from several sources. In particular, the role of alternate sources of energy was to be explored and the implications of these alternate sources to the question of Pu recycle in LWRs were to be investigated. In this vein, solar energy as an alternate source is the main subject of this report, along with other information related to solar energy. The general conclusion is that solar energy should have little effect on the decisions concerning GESMO.

Bertini, H.W.

1977-06-30T23:59:59.000Z

432

Reactor Operations informal monthly report December 1994  

Science Conference Proceedings (OSTI)

Reactor operations at the MRR and HFBR reactors at Brookhaven National Laboratory are presented for December 1994. Reactor run-time and power levels, instrumentation, mechanical maintenance, occurrence reports, and safety information are included.

NONE

1994-12-01T23:59:59.000Z

433

Measurement of theta13 with reactor neutrinos  

E-Print Network (OSTI)

? detectors ~1 km nuclear reactor Figure 2. Concept of a 2-Measurement of ? 13 with Reactor Neutrinos K.M. Heeger a ,power plant, a future reactor neutrino experiment has the

Heeger, Karsten M.; Freedman, Stuart J.; Kadel, Richard W.; Luk, Kam-Biu

2004-01-01T23:59:59.000Z

434

Granular Dynamics in Pebble Bed Reactor Cores  

E-Print Network (OSTI)

in a pebble-bed nuclear reactor,” Phys. Rev. E, vol. 74, no.cycles of the pebble bed reactor,” Nuclear Engineering andoptimization of pebble-bed reactors,” Annals of Nuclear

Laufer, Michael Robert

2013-01-01T23:59:59.000Z

435

Granular Dynamics in Pebble Bed Reactor Cores  

E-Print Network (OSTI)

flow in a pebble-bed nuclear reactor,” Phys. Rev. E, vol.from the current fleet of nuclear reactors far outweigh thethrough the core of a nuclear reactor. This regime includes

Laufer, Michael Robert

2013-01-01T23:59:59.000Z

436

Measurement of theta13 with reactor neutrinos  

E-Print Network (OSTI)

? detectors ~1 km nuclear reactor Figure 2. Concept of a 2-in the core of a nuclear reactor. Re- actor ? e are usuallybetween 0.1- 3 km from a nuclear reactor to measure the rate

Heeger, Karsten M.; Freedman, Stuart J.; Kadel, Richard W.; Luk, Kam-Biu

2004-01-01T23:59:59.000Z

437

Research Reactors Division | ORNL Neutron Sciences  

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

Reactors Division (RRD) is responsible for operation of the High Flux Isotope Reactor (HFIR). Operating at 85 MW, HFIR is the highest flux reactor-based source of neutrons for...

438

REACTOR DEVELOPMENT PROGRAM PROGRESS REPORT, MAY 1962  

SciTech Connect

Research progress is reported on water-cooled reactors, liquid-metal- cooled reactors, general reactor technology, plutonium recycle, advanced systems research and development, and nuclear safety. (M.C.G.)

1962-06-15T23:59:59.000Z

439

Conceptual innovations in hybrid reactors  

SciTech Connect

A number of innovations in the conception of fusion-fission hybrid reactors, including the blanket, the fusion driver, the coupling of the fusion and the fission components as well as the application of hybrid reactors are described, and their feasibility assessed.

Greenspan, E.; Miley, G.H.

1980-01-01T23:59:59.000Z

440

SNAP II REACTOR CORE MATERIALS  

SciTech Connect

A survey was made to select the construction materials for the SDR-1 reactor core vessel and grid plates. Hastelloy C was selected for the reactor vessel, top grid plate, and bottom grid plate. Inconel X was selected for the core hold-down springs. (C.J.G.)

Facha, J.V.

1960-04-11T23:59:59.000Z

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.


441

Neutrino Oscillations with Reactor Neutrinos  

E-Print Network (OSTI)

Prospect measurements of neutrino oscillations with reactor neutrinos are reviewed in this document. The following items are described: neutrinos oscillations status, reactor neutrino experimental strategy, impact of uncertainties on the neutrino oscillation sensitivity and, finally, the experiments in the field. This is the synthesis of the talk delivered during the NOW2006 conference at Otranto (Italy) during September 2006.

Anatael Cabrera

2007-01-11T23:59:59.000Z

442

Solvent refined coal reactor quench system  

SciTech Connect

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

443

RADIATION FACILITY FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

A radiation facility is designed for irradiating samples in close proximity to the core of a nuclear reactor. The facility comprises essentially a tubular member extending through the biological shield of the reactor and containing a manipulatable rod having the sample carrier at its inner end, the carrier being longitudinally movable from a position in close proximity to the reactor core to a position between the inner and outer faces of the shield. Shield plugs are provided within the tubular member to prevent direct radiation from the core emanating therethrough. In this device, samples may be inserted or removed during normal operation of the reactor without exposing personnel to direct radiation from the reactor core. A storage chamber is also provided within the radiation facility to contain an irradiated sample during the period of time required to reduce the radioactivity enough to permit removal of the sample for external handling. (AEC)

Currier, E.L. Jr.; Nicklas, J.H.

1961-12-12T23:59:59.000Z

444

Fuel Reformation: Microchannel Reactor Design  

DOE Green Energy (OSTI)

Fuel processing is used to extract hydrogen from conventional vehicle fuel and allow fuel cell powered vehicles to use the existing petroleum fuel infrastructure. Kilowatt scale micro-channel steam reforming, water-gas shift and preferential oxida-tion reactors have been developed capable of achieving DOE required system performance metrics. Use of a microchannel design effectively supplies heat to the highly endothermic steam reforming reactor to maintain high conversions, controls the temperature profile for the exothermic water gas shift reactor, which optimizes the overall reaction conversion, and removes heat to prevent the unwanted hydrogen oxidation in the prefer-ential oxidation reactor. The reactors combined with micro-channel heat exchangers, when scaled to a full sized 50 kWe automotive system, will be less than 21 L in volume and 52 kg in weight.

Brooks, Kriston P.; Davis, James M.; Fischer, Christopher M.; King, David L.; Pederson, Larry R.; Rawlings, Gregg C.; Stenkamp, Victoria S.; TeGrotenhuis, Ward E.; Wegeng, Robert S.; Whyatt, Greg A.

2005-09-01T23:59:59.000Z

445

COMSOL-based Nuclear Reactor Kinetics Studies at the HFIR  

Science Conference Proceedings (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; Freels, James D [ORNL; Maldonado, G Ivan [ORNL; Primm, Trent [ORNL

2011-01-01T23:59:59.000Z

446

Biosciences Division: Endurance Bioenergy Reactor(tm)  

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

Endurance Bioenergy Reactor(tm) DOE Logo Search BIO ... Search Argonne Home > BIO home > Endurance Bioenergy Reactor(tm) BIO Home Page About BIO News Releases Research Publications...

447

Advanced Reactor Technologies | Department of Energy  

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

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

448

Nuclear Reactor Technologies | Department of Energy  

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

Reactor Technologies Nuclear Reactor Technologies TVA Watts Bar Nuclear Power Plant | Photo courtesy of Tennessee Valley Authority TVA Watts Bar Nuclear Power Plant | Photo...

449

Idaho National Laboratory Advanced Test Reactor Probabilistic...  

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

Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment September 19, 2012...

450

Advanced Reactor Research and Development Funding Opportunity...  

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

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

451

Energy Department Announces Small Modular Reactor Technology...  

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

Small Modular Reactor Technology Partnerships at Savannah River Site Energy Department Announces Small Modular Reactor Technology Partnerships at Savannah River Site March 2, 2012...

452

Light Water Reactor Sustainability Program: Integrated Program...  

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

Light Water Reactor Sustainability Program: Integrated Program Plan Light Water Reactor Sustainability Program: Integrated Program Plan Nuclear power has safely, reliably, and...

453

Energy Department Announces Small Modular Reactor Technology...  

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

Energy Department Announces Small Modular Reactor Technology Partnerships at Savannah River Site Energy Department Announces Small Modular Reactor Technology Partnerships at...

454

Microstructural Characterization of Test Reactor Irradiated RPV ...  

Science Conference Proceedings (OSTI)

Presentation Title, Microstructural Characterization of Test Reactor Irradiated RPV ... Evolution in High Purity Reference V-4Cr-4Ti Alloy for Fusion Reactor.

455

Reactor Tree of Argonne National Laboratory  

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

Argonne Reactors > The Argonne Reactor Tree About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne...

456

Spectral Emissivity Measurements of High Temperature Reactor ...  

Science Conference Proceedings (OSTI)

CASL: The Consortium for Advanced Simulation of Light Water Reactors: A U.S. ... Strategies for Studying High Dose Irradiation Effects in Reactor Components.

457

Materials Degradation Issues in Pressurized Water Reactors  

Science Conference Proceedings (OSTI)

CASL: The Consortium for Advanced Simulation of Light Water Reactors: A U.S. ... Strategies for Studying High Dose Irradiation Effects in Reactor Components.

458

Steady and Periodic Pressure Measurements on a Generic Helicopter Fuselage Model in the Presence of a Rotor  

Science Conference Proceedings (OSTI)

A wind tunnel test of a generic helicopter fuselage model with an independently mounted rotor has been conducted to obtain steady and periodic pressure data on the helicopter body. The model was tested at four advance ratios and three thrust coefficients. ...

Mineck Raymond E.; Gorton Susan Althoff

2000-06-01T23:59:59.000Z

459

Reliable speculative processing of out-of-order event streams in generic publish/subscribe middlewares  

Science Conference Proceedings (OSTI)

In surveillance, sports, finances, etc., distributed event-based systems are used to detect meaningful events with low latency in high data rate event streams. Both known approaches to deal with the predominant out-of-order event arrival at the distributed ... Keywords: distributed event processing, message-oriented middleware, out-of-order event processing, publish/subscribe

Christopher Mutschler, Michael Philippsen

2013-06-01T23:59:59.000Z

460

Evaluation of Relap5 Reactor Core Modeling Capability  

E-Print Network (OSTI)

of Thesis Presented to the Graduate School of the University of Florida in Partial Fulfillment of the Requirements for the Degree of Master of Science EVALUATION OF RELAP5 REACTOR CORE MODELING CAPABILITY By Vincent J.-P. Roux August 2001 Chairman: Professor Samim Anghaie Major Department: Nuclear and Radiological Engineering RELAP5 is a one-dimensional reactor-system simulation code with additional cross-flow calculation capability to include two- and three-dimensional effects in light water nuclear reactor cores. The code is used to model the core, steam generator, and the balance of the Surry reactor, which is a three-loop Westinghouse Pressurized Water Reactor (PWR) system. A detailed RELAP5 model including full nodalization of the system is developed and implemented for this study. The RELAP5 Surry core model uses one or several parallel channels to compare and assess the performance of the cross-flow model. Several inlet flow rates and core power distributions are considered and modeled. Results of the analysis showed the significant contribution of cross-flow in overall temperature and flow distributions in the core. Results of the study also showed that the RELAP5 predictions of cross-flow, at least for single-phase cases, are not consistent with the theory. xi To evaluate the accuracy of RELAP5 cross-flow model, an industry standard Computational Fluid Dynamics code, FLUENT, is used to perform two- and threedimensional calculations. Initial and boundary conditions for the RELAP5 model are used to develop a high-resolution FLUENT model for a pair of parallel reactor core channels. Two models were developed for FLUENT calculation of cross-flow. The first model is a simple tube with axisymmetric non-uniform inlet flow velocities. The second model included differen...

Vincent J. -p. Roux

2001-01-01T23:59:59.000Z

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

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

462

GAS COOLED NUCLEAR REACTORS  

DOE Patents (OSTI)

A gas-cooled nuclear reactor consisting of a graphite reacting core and reflector structure supported in a containing vessel is described. A gas sealing means is included for sealing between the walls of the graphite structure and containing vessel to prevent the gas coolant by-passing the reacting core. The reacting core is a multi-sided right prismatic structure having a pair of parallel slots around its periphery. The containing vessel is cylindrical and has a rib on its internal surface which supports two continuous ring shaped flexible web members with their radially innermost ends in sealing engagement within the radially outermost portion of the slots. The core structure is supported on ball bearings. This design permits thermal expansion of the core stracture and vessel while maintainirg a peripheral seal between the tvo elements.

Long, E.; Rodwell, W.

1958-06-10T23:59:59.000Z

463

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

464

FLUID MODERATED REACTOR  

DOE Patents (OSTI)

A reactor which utilizes fissionable fuel elements in rod form immersed in a moderator or heavy water and a means of circulating the heavy water so that it may also function as a coolant to remove the heat generated by the fission of the fuel are described. In this design, the clad fuel elements are held in vertical tubes immersed in heavy water in a tank. The water is circulated in a closed system by entering near the tops of the tubes, passing downward through the tubes over the fuel elements and out into the tank, where it is drawn off at the bottom, passed through heat exchangers to give up its heat and then returned to the tops of the tubes for recirculation.

Wigner, E.P.; Ohlinger, L.A.; Young, G.J.; Weinberg, A.M.

1957-10-22T23:59:59.000Z

465

Licensed reactor nuclear safety criteria applicable to DOE reactors  

SciTech Connect

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

466

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

SciTech Connect

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

467

IMPROVEMENTS RELATING TO NUCLEAR REACTORS  

SciTech Connect

In order to reduce the pumping power for the coolant in a steam-cooled reactor, in which the steam being passed through successive sections of the reactor core and being superheated there, the sections are connected in series with one another, while a plurality of de-superheaters is provided such that steam flowing from one section to the next passes through a de-superheater. The condensed steam returning to the reactor from the means utilizing the steam heat content is divided into a number of separate streams. The first stream going to the first section in the reactor core is raised at least to saturated steam outside the reactor, while the remaining streams of condensed steam are conveyed to the de-superheaters to be mixed with steam passing therethrough between successive sections of the reactor, cooling in this manner said steam and being themselves converted into steam. Increasing amounts of condensate are added in successive de-superheaters until the steam returning to the reactor from the final desuperheater is equivalent to the full mass flow of steam circulating to the heat utilizing means. (NPO)

1960-08-01T23:59:59.000Z

468

DISTRIBUTION CATEGORY  

Office of Scientific and Technical Information (OSTI)

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

469

Seismic Safety Margins Research Programs. Assessment of potential increases in risk due to degradation of steam generator and reactor coolant pump supports. [PWR  

Science Conference Proceedings (OSTI)

During the NRC licensing review for the North Anna Units 1 and 2 pressurized-water reactors (PWRs), questions were raised regarding the potential for low-fracture toughness of steam-generator and reactor-coolant-pump supports. Because other PWRs may face similar problems, this issue was incorporated into the NRC Program for Resolution of Generic Issues. The work described in this report was performed to provide the NRC with a quantitative evaluation of the value/impact implications of the various options of resolving the fracture-toughness question. This report presents an assessment of the probabilistic risk associated with nil-ductility failures of steam-generator and reactor-coolant-pump structural-support systems during seismic events, performed using the Seismic Safety Margins Research Program codes and data bases.

Bohn, M. P.; Wells, J. E.; Shieh, L. C.; Cover, L. E.; Streit, R. L.

1983-08-01T23:59:59.000Z

470

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

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

471

On reactor type comparisons for the next generation of reactors  

SciTech Connect

In this paper, we present a broad comparison of studies for a selected set of parameters for different nuclear reactor types including the next generation. This serves as an overview of key parameters which provide a semi-quantitative decision basis for selecting nuclear strategies. Out of a number of advanced reactor designs of the LWR type, gas cooled type, and FBR type, currently on the drawing board, the Advanced Light Water Reactors (ALWR) seem to have some edge over other types of the next generation of reactors for the near-term application. This is based on a number of attributes related to the benefit of the vast operating experience with LWRs coupled with an estimated low risk profile, economics of scale, degree of utilization of passive systems, simplification in the plant design and layout, modular fabrication and manufacturing. 32 refs., 1 fig., 3 tabs.

Alesso, H.P.; Majumdar, K.C.

1991-08-22T23:59:59.000Z

472

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

473

United States Domestic Research Reactor Infrastrucutre TRIGA Reactor Fuel Support  

SciTech Connect

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

474

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

475

DENSITY CONTROL IN A REACTOR  

DOE Patents (OSTI)

A reactor is described in which natural-uranium bodies are located in parallel channels which extend through the graphite mass in a regular lattice. The graphite mass has additional channels that are out of the lattice and contain no uranium. These additional channels decrease in number per unit volume of graphite from the center of the reactor to the exterior and have the effect of reducing the density of the graphite more at the center than at the exterior, thereby spreading neutron activity throughout the reactor. (AEC)

Marshall, J. Jr.

1961-10-24T23:59:59.000Z

476

Fusion reactors: a remote possibility  

SciTech Connect

The next generation of controlled thermonuclear reactor experiments will be faced with the handling problems of tritium and neutron activation that will dominate the safety and maintenance problems of future fusion reactors. The nuclear industry has been working with highly radioactive systems for many years and has developed the tools and methods to do safely productive work in the presence of high radiation fields. These methods can be applied to CTR work by extending them to the unique problems associated with fusion reactors. (auth)

Doggett, J.N.

1975-11-14T23:59:59.000Z

477

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

SciTech Connect

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

2011-01-01T23:59:59.000Z

478

Numerical length estimation for tubular flow reactors  

Science Conference Proceedings (OSTI)

Keywords: free boundary value problems, noniterative and iterative numerical methods, tubular flow reactors

Riccardo Fazio

1992-08-01T23:59:59.000Z

479

2012 Annual Report Research Reactor Infrastructure Program  

SciTech Connect

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

Douglas Morrell

2012-11-01T23:59:59.000Z

480

Use of Thorium in Light Water Reactors  

Science Conference Proceedings (OSTI)

Technical Paper / Advances in Nuclear Fuel Management - Use of Alternate Fuels in Light Water Reactors

Michael Todosow; A. Galperin; S. Herring; M. Kazimi; T. Downar; A. Morozov

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

Nuclear reactor I  

DOE Patents (OSTI)

A nuclear reactor, particularly a liquid-metal breeder reactor whose upper internals include provision for channeling the liquid metal flowing from the core-component assemblies to the outlet plenum in vertical paths in direction generally along the direction of the respective assemblies. The metal is channeled by chimneys, each secured to, and extending from, a grid through whose openings the metal emitted by a plurality of core-component assemblies encompassed by the grid flows. To reduce the stresses resulting from structural interaction, or the transmissive of thermal strains due to large temperature differences in the liquid metal emitted from neighboring core-component assemblies, throughout the chimneys and the other components of the upper internals, the grids and the chimneys are supported from the heat plate and the core barrel by support columns (double portal support) which are secured to the head plate at the top and to a member, which supports the grids and is keyed to the core barrel, at the bottom. In addition to being restrained from lateral flow by the chimneys, the liquid metal is also restrained from flowing laterally by a peripheral seal around the top of the core. This seal limits the flow rate of liquid metal, which may be sharply cooled during a scram, to the outlet nozzles. The chimneys and the grids are formed of a highly-refractory, high corrosion-resistant nickel-chromium-iron alloy which can withstand the stresses produced by temperature differences in the liquid metal. The chimneys are supported by pairs of plates, each pair held together by hollow stubs coaxial with, and encircling, the chimneys. The plates and stubs are a welded structure but, in the interest of economy, are composed of stainless steel which is not weld compatible with the refractory metal. The chimneys and stubs are secured together by shells of another nickel-chromium-iron alloy which is weld compatible with, and is welded to, the stubs and has about the same coefficient of expansion as the highly-refractory, high corrosion-resistant alloy.

Ference, Edward W. (Central City, PA); Houtman, John L. (Acme, PA); Waldby, Robert N. (New Stanton, PA)

1977-01-01T23:59:59.000Z

482

STEAM GENERATOR FOR NUCLEAR REACTOR  

DOE Patents (OSTI)

The steam generator described for use in reactor powergenerating systems employs a series of concentric tubes providing annular passage of steam and water and includes a unique arrangement for separating the steam from the water. (AEC)

Kinyon, B.W.; Whitman, G.D.

1963-07-16T23:59:59.000Z

483

METHOD OF OPERATING NUCLEAR REACTORS  

DOE Patents (OSTI)

A method is presented for obtaining enhanced utilization of natural uranium in heavy water moderated nuclear reactors by charging the reactor with an equal number of fuel elements formed of natural uranium and of fuel elements formed of uranium depleted in U/sup 235/ to the extent that the combination will just support a chain reaction. The reactor is operated until the rate of burnup of plutonium equals its rate of production, the fuel elements are processed to recover plutonium, the depleted uranium is discarded, and the remaining uranium is formed into fuel elements. These fuel elements are charged into a reactor along with an equal number of fuel elements formed of uranium depleted in U/sup 235/ to the extent that the combination will just support a chain reaction, and reuse of the uranium is continued as aforesaid until it wlll no longer support a chain reaction when combined with an equal quantity of natural uranium.

Untermyer, S.

1958-10-14T23:59:59.000Z

484

Nuclear Reactors and Technology; (USA)  

SciTech Connect

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

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

1991-01-01T23:59:59.000Z

485

Reactor vessel seal service fixture  

DOE Patents (OSTI)

An apparatus for the preparation of exposed sealing surfaces along the open rim of a nuclear reactor vessel comprised of a motorized mechanism for traveling along the rim and simultaneously brushing the exposed surfaces is described.