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1

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...

2

Applied Reactor Physics TA RG E T AU D I E N C E  

E-Print Network (OSTI)

courses. Most production codes in reactor physics are accompanied with rather complete theory guides devoted to the study of interactions between neutrons and matter in a nuclear reactor. Such an interactionApplied Reactor Physics TA RG E T AU D I E N C E Applied Reactor Physics is designed for an audi

Meunier, Michel

3

Nuclear safety as applied to space power reactor systems  

SciTech Connect

To develop a strategy for incorporating and demonstrating safety, it is necessary to enumerate the unique aspects of space power reactor systems from a safety standpoint. These features must be differentiated from terrestrial nuclear power plants so that our experience can be applied properly. Some ideas can then be developed on how safe designs can be achieved so that they are safe and perceived to be safe by the public. These ideas include operating only after achieving a stable orbit, developing an inherently safe design, ''designing'' in safety from the start and managing the system development (design) so that it is perceived safe. These and other ideas are explored further in this paper.

Cummings, G.E.

1987-01-01T23:59:59.000Z

4

Nuclear safety as applied to space power reactor systems  

SciTech Connect

Current space nuclear power reactor safety issues are discussed with respect to the unique characteristics of these reactors. An approach to achieving adequate safety and a perception of safety is outlined. This approach calls for a carefully conceived safety program which makes uses of lessons learned from previous terrestrial power reactor development programs. This approach includes use of risk analyses, passive safety design features, and analyses/experiments to understand and control off-design conditions. The point is made that some recent accidents concerning terrestrial power reactors do not imply that space power reactors cannot be operated safety.

Cummings, G.E.

1987-01-01T23:59:59.000Z

5

4-D Particle Tracking Velocimetry Applied to Gas-Liquid Reactors  

E-Print Network (OSTI)

4-D Particle Tracking Velocimetry Applied to Gas-Liquid Reactors Michael St¨ohr12 , Christoph Garbe bubbles through space and time (4-D) in a liquid container, e.g. in a gas-liquid reactor. The method cameras looking into the gas-liquid reactor. The air bubbles are visualized by light refracted

Garbe, Christoph S.

6

Liquid metal reactor deactivation as applied to the experimental breeder reactor - II.  

DOE Green Energy (OSTI)

The Experimental Breeder Reactor-II (EBR-II) at Argonne National Laboratory-West (ANL-W) was shutdown in September, 1994. This sodium cooled reactor had been in service since 1964, and by the US Department of Energy (DOE) mandate, was to be placed in an industrially and radiologically safe condition for ultimate decommissioning. The deactivation of a liquid metal reactor presents unique concerns. The first major task associated with the project was the removal of all fueled assemblies. In addition, sodium must be drained from systems and processed for ultimate disposal. Residual quantities of sodium remaining in systems must be deactivated or inerted to preclude future hazards associated with pyrophoricity and generation of potentially explosive hydrogen gas. A Sodium Process Facility (SPF) was designed and constructed to react the elemental sodium from the EBR-II primary and secondary systems to sodium hydroxide for disposal. This facility has a design capacity to allow the reaction of the complete inventory of sodium at ANL-W in less than two years. Additional quantities of sodium from the Fermi-1 reactor are also being treated at the SPF.

Earle, O. K.; Michelbacher, J. A.; Pfannenstiel, D. F.; Wells, P. B.

1999-05-28T23:59:59.000Z

7

Applying Bayesian Neural Networks to Separate Neutrino Events from Backgrounds in Reactor Neutrino Experiments  

E-Print Network (OSTI)

A toy detector has been designed to simulate central detectors in reactor neutrino experiments in the paper. The samples of neutrino events and three major backgrounds from the Monte-Carlo simulation of the toy detector are generated in the signal region. The Bayesian Neural Networks(BNN) are applied to separate neutrino events from backgrounds in reactor neutrino experiments. As a result, the most neutrino events and uncorrelated background events in the signal region can be identified with BNN, and the part events each of the fast neutron and $^{8}$He/$^{9}$Li backgrounds in the signal region can be identified with BNN. Then, the signal to noise ratio in the signal region is enhanced with BNN. The neutrino discrimination increases with the increase of the neutrino rate in the training sample. However, the background discriminations decrease with the decrease of the background rate in the training sample.

Ye Xu; Yixiong Meng; Weiwei Xu

2008-08-02T23:59:59.000Z

8

Applying risk informed methodologies to improve the economics of sodium-cooled fast reactors  

E-Print Network (OSTI)

In order to support the increasing demand for clean sustainable electricity production and for nuclear waste management, the Sodium-Cooled Fast Reactor (SFR) is being developed. The main drawback has been its high capital ...

Nitta, Christopher C

2010-01-01T23:59:59.000Z

9

New Tool for Proliferation Resistance Evaluation Applied to Uranium and Thorium Fueled Fast Reactor Fuel Cycles  

E-Print Network (OSTI)

The comparison of nuclear facilities based on their barriers to nuclear material proliferation has remained a difficult endeavor, often requiring expert elicitation for each system under consideration. However, objectively comparing systems using a set of computable metrics to derive a single number representing a system is not, in essence, a nuclear nonproliferation specific problem and significant research has been performed for business models. For instance, Multi-Attribute Utility Analysis (MAUA) methods have been used previously to provide an objective insight of the barriers to proliferation. In this paper, the Proliferation Resistance Analysis and Evaluation Tool for Observed Risk (PRAETOR), a multi-tiered analysis tool based on the multiplicative MAUA method, is presented. It folds sixty three mostly independent metrics over three levels of detail to give an ultimate metric for nonproliferation performance comparison. In order to reduce analysts' bias, the weighting between the various metrics was obtained by surveying a total of thirty three nonproliferation specialists and nonspecialists from fields such as particle physics, international policy, and industrial engineering. The PRAETOR was used to evaluate the Fast Breeder Reactor Fuel Cycle (FBRFC). The results obtained using these weights are compared against a uniform weight approach. Results are presented for five nuclear material diversion scenarios: four examples include a diversion attempt on various components of a PUREX fast reactor cycle and one scenario involves theft from a PUREX facility in a LWR cycle. The FBRFC was evaluated with uranium-plutonium fuel and a second time using thorium-uranium fuel. These diversion scenarios were tested with both uniform and expert weights, with and without safeguards in place. The numerical results corroborate nonproliferation truths and provide insight regarding fast reactor facilities' proliferation resistance in relation to known standards.

Metcalf, Richard R.

2009-05-01T23:59:59.000Z

10

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

11

The BURNUP package of applied programs used for computing the isotopic composition of materials of an operating nuclear reactor  

SciTech Connect

This paper described the procedure of implementation and the possibilities of the BURNUP program. The purpose of the program is to predict the change in the nuclear composition of the materials of which a reactor is made in the course of its run and compute the radiation characteristics of the materials after their irradiation.

Yudkevich, M. S., E-mail: umark@adis.vver.kiae.ru [National Research Centre Kurchatov Institute (Russian Federation)

2012-12-15T23:59:59.000Z

12

Initial Evaluation of the Heat-Affected Zone, Local Embrittlement Phenomenon as it Applies to Nuclear Reactor Vessels  

Science Conference Proceedings (OSTI)

The objective of this project was to determine if the local brittle zone (LBZ) problem, encountered in the testing of the heat-affected zone (HAZ) part of welds in offshore platform construction, can also be found in reactor pressure vessel (RPV) welds. Both structures have multipass welds and grain coarsening along the fusion line. Literature was obtained that described the metallurgical evidence and the type of research work performed on offshore structure welds.

McCabe, D.E.

1999-09-01T23:59:59.000Z

13

Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part II: Prismatic Reactor Cross Section Generation  

SciTech Connect

The deep-burn prismatic high temperature reactor is made up of an annular core loaded with transuranic isotopes and surrounded in the center and in the periphery by reflector blocks in graphite. This disposition creates challenges for the neutronics compared to usual light water reactor calculation schemes. The longer mean free path of neutrons in graphite affects the neutron spectrum deep inside the blocks located next to the reflector. The neutron thermalisation in the graphite leads to two characteristic fission peaks at the inner and outer interfaces as a result of the increased thermal flux seen in those assemblies. Spectral changes are seen at least on half of the fuel blocks adjacent to the reflector. This spectral effect of the reflector may prevent us from successfully using the two step scheme -lattice then core calculation- typically used for light water reactors. We have been studying the core without control mechanisms to provide input for the development of a complete calculation scheme. To correct the spectrum at the lattice level, we have tried to generate cross-sections from supercell calculations at the lattice level, thus taking into account part of the graphite surrounding the blocks of interest for generating the homogenised cross-sections for the full-core calculation. This one has been done with 2 to 295 groups to assess if increasing the number of groups leads to more accurate results. A comparison with a classical single block model has been done. Both paths were compared to a reference calculation done with MCNP. It is concluded that the agreement with MCNP is better with supercells, but that the single block model remains quite close if enough groups are kept for the core calculation. 26 groups seems to be a good compromise between time and accu- racy. However, some trials with depletion have shown huge variations of the isotopic composition across a block next to the reflector. It may imply that at least an in- core depletion for the number density calculation may be necessary in the complete calculation scheme.

Vincent Descotes

2011-03-01T23:59:59.000Z

14

Radiation physics and shielding codes and analyses applied to design-assist and safety analyses of CANDU{sup R} and ACR{sup TM} reactors  

SciTech Connect

This paper discusses the radiation physics and shielding codes and analyses applied in the design of CANDU and ACR reactors. The focus is on the types of analyses undertaken rather than the inputs supplied to the engineering disciplines. Nevertheless, the discussion does show how these analyses contribute to the engineering design. Analyses in radiation physics and shielding can be categorized as either design-assist or safety and licensing (accident) analyses. Many of the analyses undertaken are designated 'design-assist' where the analyses are used to generate recommendations that directly influence plant design. These recommendations are directed at mitigating or reducing the radiation hazard of the nuclear power plant with engineered systems and components. Thus the analyses serve a primary safety function by ensuring the plant can be operated with acceptable radiation hazards to the workers and public. In addition to this role of design assist, radiation physics and shielding codes are also deployed in safety and licensing assessments of the consequences of radioactive releases of gaseous and liquid effluents during normal operation and gaseous effluents following accidents. In the latter category, the final consequences of accident sequences, expressed in terms of radiation dose to members of the public, and inputs to accident analysis, e.g., decay heat in fuel following a loss-of-coolant accident, are also calculated. Another role of the analyses is to demonstrate that the design of the plant satisfies the principle of ALARA (as low as reasonably achievable) radiation doses. This principle is applied throughout the design process to minimize worker and public doses. The principle of ALARA is an inherent part of all design-assist recommendations and safety and licensing assessments. The main focus of an ALARA exercise at the design stage is to minimize the radiation hazards at the source. This exploits material selection and impurity specifications and relies heavily on experience and engineering judgement, consistent with the ALARA philosophy. Special care is taken to ensure that the best estimate dose rates are used to the extent possible when applying ALARA. Provisions for safeguards equipment are made throughout the fuel-handling route in CANDU and ACR reactors. For example, the fuel bundle counters rely on the decay gammas from the fission products in spent-fuel bundles to record the number of fuel movements. The International Atomic Energy Agency (IAEA) Safeguards system for CANDU and ACR reactors is based on item (fuel bundle) accounting. It involves a combination of IAEA inspection with containment and surveillance, and continuous unattended monitoring. The spent fuel bundle counter monitors spent fuel bundles as they are transferred from the fuelling machine to the spent fuel bay. The shielding and dose-rate analysis need to be carried out so that the bundle counter functions properly. This paper includes two codes used in criticality safety analyses. Criticality safety is a unique phenomenon and codes that address criticality issues will demand specific validations. However, it is recognised that some of the codes used in radiation physics will also be used in criticality safety assessments. (authors)

Aydogdu, K.; Boss, C. R. [Atomic Energy of Canada Limited, Sheridan Science and Technology Park, Mississauga, Ont. L5K 1B2 (Canada)

2006-07-01T23:59:59.000Z

15

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

16

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

17

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

18

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...

19

Industrial tomography applied to reactor safety  

SciTech Connect

Work has begun which explores the use of Computed Axial Tomography (CAT), boundary detection, and internal surface reconstruction techniques in industrial nondestructive testing applications. This initial work is intended to inform the reader of the existence and interrelated nature of these techniques through the use of a realistic simulation of an industrial inspection problem.

Kruger, R.P.

1977-01-01T23:59:59.000Z

20

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

Note: This page contains sample records for the topic "reactor cxs applied" 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

CX-008746: Categorical Exclusion Determination | Department of...  

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

Exclusion Determination Reactor Infrastructure Improvement at Kansas State University Nuclear Reactor - Kansas State University CX(s) Applied: B2.2 Date: 05212012...

22

Microfluidic reactors for the synthesis of nanocrystals  

E-Print Network (OSTI)

Several microfluidic reactors were designed and applied to the synthesis of colloidal semiconductor nanocrystals (NCs). Initially, a simple single-phase capillary reactor was used for the synthesis of CdSe NCs. Precursors ...

Yen, Brian K. H

2007-01-01T23:59:59.000Z

23

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

24

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

25

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

26

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

27

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

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

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

28

Achievements: Nuclear Reactors designed/built by Argonne National  

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

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

29

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

30

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

31

REACTOR DEVELOPMENT PROGRAM PROGRESS REPORT (FOR) JULY 1961  

SciTech Connect

A summary is presented of activities in reactor and general engineering research programs. Discussions are included for developments in EBWR, BORAX-V, ZPR-III. ZPR-VI, ZPR-IX, EBR-I, and EBR-II. Reactor safety studies were performed for fast and thermal reactors. Nuclear technology developments are discussed for applied nuclear and reactor physics, reactor fuels and materials development, heat engineering studies, separations processes, and advanced reactor concepts. (B.O.G.)

1961-08-15T23:59:59.000Z

32

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

33

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

34

Applied Science  

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

Applied Science Applied Science Correlation of predicted and measured iron oxidation states in mixed iron oxides H. D. Rosenfeld and W. L. Holstein Development of a quantitative measurement of a diesel spray core using synchrotron x-rays C.F. Powell, Y. Yue, S. Gupta, A. McPherson, R. Poola, and J. Wang Localized phase transformations by x-ray-induced heating R.A. Rosenberg, Q. Ma, W. Farrell, E.D. Crozier, G.J. Soerensen, R.A. Gordon, and D.-T. Jiang Resonant x-ray scattering at the Se edge in ferroelectric liquid crystal materials L. Matkin, H. Gleeson, R. Pindak, P. Mach, C. Huang, G. Srajer, and J. Pollmann Synchrotron-radiation-induced anisotropic wet etching of GaAs Q. Ma, D.C. Mancini, and R.A. Rosenberg Synchrotron-radiation-induced, selective-area deposition of gold on

35

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

36

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

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

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

37

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

38

Early Exploration - Reactors designed/built by Argonne National Laboratory  

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

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

39

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

40

Request for Naval Reactors Comment on Proposed Prometheus Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to JPL  

SciTech Connect

The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.

D. Kokkinos

2005-04-28T23:59:59.000Z

Note: This page contains sample records for the topic "reactor cxs applied" 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

Applied antineutrino physics workshop.  

Science Conference Proceedings (OSTI)

This workshop is the fourth one of a series that includes the Neutrino Geophysics Conference at Honolulu, Hawaii, which I attended in 2005. This workshop was organized by the Astro-Particle and Cosmology laboratory in the recently opened Condoret building of the University of Paris. More information, including copies of the presentations, on the workshop is available on the website: www.apc.univ-paris7.fr/AAP2007/. The workshop aims at opening neutrino physics to various fields such that it can be applied in geosciences, nuclear industry (reactor and spent fuel monitoring) and non-proliferation. The workshop was attended by over 60 people from Europe, USA, Asia and Brazil. The meeting was also attended by representatives of the Comprehensive nuclear-Test Ban Treaty (CTBT) and the International Atomic Energy Agency (IAEA). The workshop also included a workshop dinner on board of a river boat sailing the Seine river.

Lund, James C.

2008-01-01T23:59:59.000Z

42

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...

43

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

44

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

45

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

46

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

47

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

48

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

49

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

50

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...

51

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

52

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

53

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

54

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

55

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

56

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

57

CX-008723: Categorical Exclusion Determination | Department of...  

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

Exclusion Determination CX-008723: Categorical Exclusion Determination Zero Power Physics Reactor Surveillance Glove Box CX(s) Applied: B2.5 Date: 07092012 Location(s):...

58

CX-007703: Categorical Exclusion Determination | Department of...  

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

Determination Sheetak Inc. - Thermoelectric Reactors for Efficient Automotive Thermal Storage CX(s) Applied: B3.6 Date: 11102011 Location(s): New York, Pennsylvania,...

59

Page not found | Department of Energy  

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

CX-009297: Categorical Exclusion Determination Optimization of Pressurized Oxy-Combustion with Flameless Reactor - Phase I CX(s) Applied: A9 Date: 08312012 Location(s):...

60

CX-009298: Categorical Exclusion Determination | Department of...  

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

8: Categorical Exclusion Determination CX-009298: Categorical Exclusion Determination Optimization of Pressurized Oxy-Combustion with Flameless Reactor - Phase I CX(s) Applied:...

Note: This page contains sample records for the topic "reactor cxs applied" 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

CX-009297: Categorical Exclusion Determination | Department of...  

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

7: Categorical Exclusion Determination CX-009297: Categorical Exclusion Determination Optimization of Pressurized Oxy-Combustion with Flameless Reactor - Phase I CX(s) Applied: A9...

62

CX-009299: Categorical Exclusion Determination | Department of...  

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

9: Categorical Exclusion Determination CX-009299: Categorical Exclusion Determination Optimization of Pressurized Oxy-Combustion with Flameless Reactor - Phase I CX(s) Applied:...

63

CX-001627: Categorical Exclusion Determination | Department of...  

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

7: Categorical Exclusion Determination CX-001627: Categorical Exclusion Determination Test Reactor Cask Implementation CX(s) Applied: B2.5 Date: 04122010 Location(s): Idaho...

64

CX-009036: Categorical Exclusion Determination | Department of...  

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

Exclusion Determination Nanoscale Stable Precipitation-Strengthened Steels for Nuclear Reactor Applications - Los Alamos National Laboratory CX(s) Applied: B3.6, B3.10...

65

CX-000498: Categorical Exclusion Determination | Department of...  

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

Categorical Exclusion Determination Groundwater Monitoring Well Installation at Heavy Water Component Test Reactor CX(s) Applied: B3.1 Date: 06252009 Location(s): Aiken,...

66

CX-000880: Categorical Exclusion Determination | Department of...  

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

Exclusion Determination CX-000880: Categorical Exclusion Determination Deactivation of Heavy Water Components Test Reactor, 770-U CX(s) Applied: B1.28 Date: 02232010...

67

CX-000509: Categorical Exclusion Determination | Department of...  

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

9: Categorical Exclusion Determination CX-000509: Categorical Exclusion Determination Heavy Water Component Test Reactor (HWCTR) Well Installation CX(s) Applied: B3.1 Date: 0805...

68

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...

69

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

70

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

71

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

72

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

73

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

74

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

75

Chicago Pile reactors create enduring research legacy - Argonne's  

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

Chicago Pile reactors create enduring research Chicago Pile reactors create enduring research legacy 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

76

Early Argonne reactor lit the way for worldwide nuclear industry -  

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

Early Argonne reactor lit the way for worldwide Early Argonne reactor lit the way for worldwide nuclear industry 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

77

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

78

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

79

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

80

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

Note: This page contains sample records for the topic "reactor cxs applied" 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

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

82

Polynomial regression with derivative information in nuclear reactor uncertainty quantification*  

E-Print Network (OSTI)

1 Polynomial regression with derivative information in nuclear reactor uncertainty quantification in the outputs. The usual difficulties in modeling the work of the nuclear reactor models include the large size, applying the existing AD tools to nuclear reactor models still takes considerable development effort

Anitescu, Mihai

83

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

84

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

85

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

86

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

87

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

88

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

89

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,...

90

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...

91

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

92

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

93

Conceptual design study of spheromak reactors  

SciTech Connect

Preliminary design studies are carried out for a spheromak fusion reactor. Simplified circuit theory is applied to obtain characteristic relations among various parameters of the spheromak configuration for an aspect ratio A greater than or equal to 1.6. These relations are used to calculate the parameters for the conceptual designs of three types of fusion reactor: (1) DT two-component, (2) DT ignited, and, (3) catalyzed DD ignited reactors. With a total wall loading of approx. 4 MWm/sup -2/, it is found that edge magnetic fields of only approx. 4T (DT) and approx. 9T (cat. DD) are required for ignited reactors of one-meter plasma (minor) radius with output powers in the gigawatt range. Assessment of various methods of generating reactor-grade spheromak plasmas is discussed briefly.

Katsurai, M.; Yamada, M.

1980-07-01T23:59:59.000Z

94

Why Nuclear Energy? - Reactors designed/built by Argonne National  

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

Nuclear Energy: Nuclear Energy: Why Nuclear Energy? 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

95

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

96

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

97

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

98

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

99

Applied Quantum Information Science  

Science Conference Proceedings (OSTI)

Applied Quantum Information Science. Summary: Theory is being developed and used to devise methods for preserving ...

2012-05-30T23:59:59.000Z

100

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...

Note: This page contains sample records for the topic "reactor cxs applied" 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

CX-008819: Categorical Exclusion Determination | Department of...  

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

Plant Project CX(s) Applied: B1.15 Date: 06202012 Location(s): Idaho Offices(s): Naval Nuclear Propulsion Program, Naval Reactors The scope of the Naval Reactors Facility (NRF)...

102

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

103

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

104

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

105

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

106

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

107

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

108

Uncertainty quantification approaches for advanced reactor analyses.  

SciTech Connect

The original approach to nuclear reactor design or safety analyses was to make very conservative modeling assumptions so as to ensure meeting the required safety margins. Traditional regulation, as established by the U. S. Nuclear Regulatory Commission required conservatisms which have subsequently been shown to be excessive. The commission has therefore moved away from excessively conservative evaluations and has determined best-estimate calculations to be an acceptable alternative to conservative models, provided the best-estimate results are accompanied by an uncertainty evaluation which can demonstrate that, when a set of analysis cases which statistically account for uncertainties of all types are generated, there is a 95% probability that at least 95% of the cases meet the safety margins. To date, nearly all published work addressing uncertainty evaluations of nuclear power plant calculations has focused on light water reactors and on large-break loss-of-coolant accident (LBLOCA) analyses. However, there is nothing in the uncertainty evaluation methodologies that is limited to a specific type of reactor or to specific types of plant scenarios. These same methodologies can be equally well applied to analyses for high-temperature gas-cooled reactors and to liquid metal reactors, and they can be applied to steady-state calculations, operational transients, or severe accident scenarios. This report reviews and compares both statistical and deterministic uncertainty evaluation approaches. Recommendations are given for selection of an uncertainty methodology and for considerations to be factored into the process of evaluating uncertainties for advanced reactor best-estimate analyses.

Briggs, L. L.; Nuclear Engineering Division

2009-03-24T23:59:59.000Z

109

Design guide for Category III reactors: pool type reactors. [US DOE  

SciTech Connect

The Department of Energy (DOE) in the ERDA Manual requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. 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 requirement of Category III reactor structures, components, and systems.

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

1978-11-01T23:59:59.000Z

110

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

111

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

112

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

113

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

114

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

115

Design guide for category VI reactors: air-cooled graphite 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 air-cooled graphite 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).

Brynda, W.J.; Karol, R.; Powell, R.W.

1979-02-01T23:59:59.000Z

116

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

117

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

118

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

119

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

120

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

Note: This page contains sample records for the topic "reactor cxs applied" 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

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

122

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

123

SOUTHVIEWDR Center for Applied  

E-Print Network (OSTI)

/Geology Chemistry Biological Sciences Geology Lab Bookstore Reed Milledge Payne Memorial Hall SANFORD DR Center CAES Activity Center Visitors Center (Four Towers) Greenhouses Center for Applied Isotope Study

Hall, Daniel

124

Assessment of tritium breeding requirements for fusion power reactors  

Science Conference Proceedings (OSTI)

This report presents an assessment of tritium-breeding requirements for fusion power reactors. The analysis is based on an evaluation of time-dependent tritium inventories in the reactor system. The method presented can be applied to any fusion systems in operation on a steady-state mode as well as on a pulsed mode. As an example, the UWMAK-I design was analyzed and it has been found that the startup inventory requirement calculated by the present method significantly differs from those previously calculated. The effect of reactor-parameter changes on the required tritium breeding ratio is also analyzed for a variety of reactor operation scenarios.

Jung, J.

1983-12-01T23:59:59.000Z

125

Physics of reactor safety. Quarterly report, October-December 1980. Volume IV  

SciTech Connect

The work in the Applied Physics Division includes reports on reactor safety modeling and assessment by members of the Reactor Safety Appraisals Section. Work on reactor core thermal-hydraulics is performed in ANL's Components Technology Division, emphasizing 3-dimensional code development for LMFBR accidents under natural convection conditions.

Not Available

1981-02-01T23:59:59.000Z

126

Physics of reactor safety. Volume II. Quarterly report, April-June 1980  

Science Conference Proceedings (OSTI)

The work in the Applied Physics Division includes reports on reactor safety modeling and assessment by members of the Reactor Safety Appraisals Section. Work on reactor core thermal-hydraulics is performed in ANL's Components Technology Division, emphasizing 3-dimensional code development for LMFBR accidents under natural convection conditions.

Not Available

1980-08-01T23:59:59.000Z

127

Physics of reactor safety. Quarterly report, January-March 1980. Volume I  

SciTech Connect

The work in the Applied Physics Division includes reports on reactor safety modeling and assessment by members of the Reactor Safety Appraisals Section. Work on reactor core thermal-hydraulics is performed in ANL's Components Technology Division, emphasizing 3-dimensional code development for LMFBR accidents under natural convection conditions.

Not Available

1980-05-01T23:59:59.000Z

128

Physical protection of power reactors  

SciTech Connect

Sandia Laboratories has applied a systematic approach to designing physical protection systems for nuclear facilities to commercial light-water reactor power plants. A number of candidate physical protection systems were developed and evaluated. Focus is placed on the design of access control subsystems at each of three plant layers: the protected area perimeter, building surfaces, and vital areas. Access control refers to barriers, detectors, and entry control devices and procedures used to keep unauthorized personnel and contraband out of the plant, and to control authorized entry into vital areas within the plant.

Darby, J.L.

1979-01-01T23:59:59.000Z

129

Technology, Safety and Costs of Decommissioning Nuclear Reactors At Multiple-Reactor Stations  

SciTech Connect

Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWRs) and large (1155-MWe) boiling water reactors {BWRs) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite nuclear waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services. Five scenarios for decommissioning reactors at a multiple-reactor station are investigated. The number of reactors on a site is assumed to be either four or ten; nuclear waste disposal is varied between immediate offsite disposal, interim onsite storage, and immediate onsite disposal. It is assumed that the decommissioned reactors are not replaced in one scenario but are replaced in the other scenarios. Centralized service facilities are provided in two scenarios but are not provided in the other three. Decommissioning of a PWR or a BWR at a multiple-reactor station probably will be less costly and result in lower radiation doses than decommissioning an identical reactor at a single-reactor station. Regardless of whether the light water reactor being decommissioned is at a single- or multiple-reactor station: • the estimated occupational radiation dose for decommissioning an LWR is lowest for SAFSTOR and highest for DECON • the estimated cost of decommissioning a PWR is lowest for ENTOMB and highest for SAFSTOR • the estimated cost of decommissioning a BWR is lowest for OECON and highest for SAFSTOR. In all cases, SAFSTOR has the lowest occupational radiation dose and the highest cost.

Wittenbrock, N. G.

1982-01-01T23:59:59.000Z

130

Applied Energy Programs  

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

Applied Energy Programs Applied Energy Programs Applied Energy Programs Los Alamos is using its world-class scientific capabilities to enhance national energy security by developing energy sources with limited environmental impact and by improving the efficiency and reliability of the energy infrastructure. CONTACT US Acting Program Director Melissa Fox (505) 663-5538 Email Applied Energy Program Office serves as the hub connecting the Laboratory's scientific and technical resources to DOE sponsors, DoD programs, and to industry. The Applied Energy Program Office manages Los Alamos National Laboratory programs funded by the Department of Energy's (DOE's) Offices of Energy Efficiency/Renewable Energy, Electricity Delivery and Energy Reliability, and Fossil Energy. With energy use increasing across the nation and the

131

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),

132

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

133

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

134

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

135

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

136

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

137

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

138

Pressurized fluidized bed reactor  

DOE Patents (OSTI)

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

Isaksson, J.

1996-03-19T23:59:59.000Z

139

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

140

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

Note: This page contains sample records for the topic "reactor cxs applied" 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

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

142

Essays in applied microeconomics  

E-Print Network (OSTI)

This dissertation consists of three chapters on topics in applied microeconomics. In the first chapter. I investigate whether voters are more likely to support additional spending on local public services when they perceive ...

Aron-Dine, Aviva

2012-01-01T23:59:59.000Z

143

Advanced reactor design study. Assessing nonbackfittable concepts for improving uranium utilization in light water reactors  

Science Conference Proceedings (OSTI)

The objective of the Advanced Reactor Design Study (ARDS) is to identify and evaluate nonbackfittable concepts for improving uranium utilization in light water reactors (LWRs). The results of this study provide a basis for selecting and demonstrating specific nonbackfittable concepts that have good potential for implementation. Lead responsibility for managing the study was assigned to the Pacific Northwest Laboratory (PNL). Nonbackfittable concepts for improving uranium utilization in LWRs on the once-through fuel cycle were selected separately for PWRs and BWRs due to basic differences in the way specific concepts apply to those plants. Nonbackfittable concepts are those that are too costly to incorporate in existing plants, and thus, could only be economically incorporated in new reactor designs or plants in very early stages of construction. Essential results of the Advanced Reactor Design Study are summarized.

Fleischman, R.M.; Goldsmith, S.; Newman, D.F.; Trapp, T.J.; Spinrad, B.I.

1981-09-01T23:59:59.000Z

144

ORISE: Applied health physics projects  

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

for government agencies, including the U.S. Nuclear Regulatory Commission (NRC) Office of Nuclear Reactor Regulation, by supporting the development of technical evaluation reports...

145

Measuring of fissile isotopes partial antineutrino spectra in direct experiment at nuclear reactor  

E-Print Network (OSTI)

The direct measuring method is considered to get nuclear reactor antineutrino spectrum. We suppose to isolate partial spectra of the fissile isotopes by using the method of antineutrino spectrum extraction from the inverse beta decay positron spectrum applied at Rovno experiment. This admits to increase the accuracy of partial antineutrino spectra forming the total nuclear reactor spectrum. It is important for the analysis of the reactor core fuel composition and could be applied for non-proliferation purposes.

V. V. Sinev

2009-02-22T23:59:59.000Z

146

Measuring of fissile isotopes partial antineutrino spectra in direct experiment at nuclear reactor  

E-Print Network (OSTI)

The direct measuring method is considered to get nuclear reactor antineutrino spectrum. We suppose to isolate partial spectra of the fissile isotopes by using the method of antineutrino spectrum extraction from the inverse beta decay positron spectrum applied at Rovno experiment. This admits to increase the accuracy of partial antineutrino spectra forming the total nuclear reactor spectrum. It is important for the analysis of the reactor core fuel composition and could be applied for non-proliferation purposes.

Sinev, V V

2009-01-01T23:59:59.000Z

147

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...

148

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

149

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

150

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

151

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

152

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

153

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

154

Applied Mathematics | Argonne National Laboratory  

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

Applied Mathematics Applied Mathematics Our work in applied mathematics ranges from algorithm design, to development of software tools and technology, to advanced simulations in...

155

Applied Science/Techniques  

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

Applied Science/Techniques Applied Science/Techniques Applied Science/Techniques Print The ALS is an excellent incubator of new scientific techniques and instrumentation. Many of the technical advances that make the ALS a world-class soft x-ray facility are developed at the ALS itself. The optical components in use at the ALS-mirrors and lenses optimized for x-ray wavelengths-require incredibly high-precision surfaces and patterns (often formed through extreme ultraviolet lithography at the ALS) and must undergo rigorous calibration and testing provided by beamlines and equipment from the ALS's Optical Metrology Lab and Berkeley Lab's Center for X-Ray Optics. New and/or continuously improved experimental techniques are also a crucial element of a thriving scientific facility. At the ALS, examples of such "technique" highlights include developments in lensless imaging, soft x-ray tomography, high-throughput protein analysis, and high-power coherent terahertz radiation.

156

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

157

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

158

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

159

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

160

Nuclear reactor reflector  

DOE Patents (OSTI)

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

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

1994-01-01T23:59:59.000Z

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


161

Nuclear reactor reflector  

DOE Patents (OSTI)

A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled. 12 figs.

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

1994-06-07T23:59:59.000Z

162

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

163

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

164

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

165

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

166

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

167

Applied Science/Techniques  

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

Applied Science/Techniques Print Applied Science/Techniques Print The ALS is an excellent incubator of new scientific techniques and instrumentation. Many of the technical advances that make the ALS a world-class soft x-ray facility are developed at the ALS itself. The optical components in use at the ALS-mirrors and lenses optimized for x-ray wavelengths-require incredibly high-precision surfaces and patterns (often formed through extreme ultraviolet lithography at the ALS) and must undergo rigorous calibration and testing provided by beamlines and equipment from the ALS's Optical Metrology Lab and Berkeley Lab's Center for X-Ray Optics. New and/or continuously improved experimental techniques are also a crucial element of a thriving scientific facility. At the ALS, examples of such "technique" highlights include developments in lensless imaging, soft x-ray tomography, high-throughput protein analysis, and high-power coherent terahertz radiation.

168

Electric Power Produced from Nuclear Reactor | National Nuclear Security  

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

Electric Power Produced from Nuclear Reactor | National Nuclear Security Electric Power Produced from Nuclear Reactor | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > About Us > Our History > NNSA Timeline > Electric Power Produced from Nuclear Reactor Electric Power Produced from Nuclear Reactor December 20, 1951 Arco, ID Electric Power Produced from Nuclear Reactor

169

Maria Research Reactor loaded with LEU - Otwock, Poland | National Nuclear  

National Nuclear Security Administration (NNSA)

Maria Research Reactor loaded with LEU - Otwock, Poland | National Nuclear Maria Research Reactor loaded with LEU - Otwock, Poland | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Video Gallery > Maria Research Reactor loaded with LEU - ... Maria Research Reactor loaded with LEU - Otwock, Poland Maria Research Reactor loaded with LEU - Otwock, Poland

170

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

171

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

172

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

173

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

174

Physics of reactor safety. Quarterly report, July-September 1980. Volume III  

SciTech Connect

This Quarterly progress report summarizes work done during the months of July-September 1980 in Argonne National Laboratory's Applied Physics and Components Technology Divisions for the Division of Reactor Safety Research of the US Nuclear Regulatory Commission. The work in the Applied Physics Division includes reports on reactor safety modeling and assessment by members of the Reactor Safety Appraisals Section. Work on reactor core thermal-hydraulics is performed in ANL's Components Technology Division, emphasizing 3-dimensional code development for LMFBR accidents under natural convection conditions. An executive summary is provided including a statement of the findings and recommendations of the report.

Not Available

1980-11-01T23:59:59.000Z

175

Safety and licensing for small and medium power reactors  

SciTech Connect

Proposed new concepts for small and medium power reactors differ substantially from traditional Light Water Reactors (LWRs). Although designers have a large base of experience in safety and licensing, much of it is not relevant to new concepts. It can be a disadvantage if regulators apply LWR rules directly. A fresh start is appropriate. The extensive interactions between industry, regulators, and the public complicates but may enhance safety. It is basic to recognize the features that distinguish nuclear energy safety from that for other industries. These features include: nuclear reactivity, fission product radiation, and radioactive decay heat. Small and medium power reactors offer potential advantages over LWRs, particularly for reactivity and decay heat.

Trauger, D.B.

1987-01-01T23:59:59.000Z

176

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

177

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

178

CX-008734: Categorical Exclusion Determination  

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

Infrastructure for Advanced Reactor Materials Center CX(s) Applied: B2.2 Date: 05/28/2012 Location(s): Idaho Offices(s): Idaho Operations Office

179

CX-005489: Categorical Exclusion Determination | Department of...  

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

89: Categorical Exclusion Determination CX-005489: Categorical Exclusion Determination Well Installations at P-Area Reactor Groundwater Operable Unit CX(s) Applied: B3.1 Date: 03...

180

Page not found | Department of Energy  

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

Download CX-009221: Categorical Exclusion Determination Rerouting of Power Lines at 100-D Reactor in Support of Site Remediation CX(s) Applied: B4.13 Date: 09242012...

Note: This page contains sample records for the topic "reactor cxs applied" 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

CX-003685: Categorical Exclusion Determination | Department of...  

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

Determination CX-003685: Categorical Exclusion Determination Photo Reactor for Growing Algae from Municipal Waste Water for Carbon Dioxide Capture CX(s) Applied: A1, B3.6 Date:...

182

CX-009034: Categorical Exclusion Determination | Department of...  

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

Accelerated Development of Zr-containing New Generation FM Steels for Advanced Nuclear Reactors - Oak Ridge National Laboratory CX(s) Applied: B3.6, B3.10 Date: 0809...

183

CX-010487: Categorical Exclusion Determination  

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

Defense Waste Processing Facility Precipitate Reactor Feed Tank (PRFT) Sample Analysis CX(s) Applied: B3.6 Date: 05/22/2013 Location(s): South Carolina Offices(s): Savannah River Operations Office

184

CX-010434: Categorical Exclusion Determination  

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

LaPine Substation Shunt Reactor Addition CX(s) Applied: B4.6 Date: 06/03/2013 Location(s): Oregon, Oregon Offices(s): Bonneville Power Administration

185

CX-010399: Categorical Exclusion Determination  

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

High Density Fuel Material for Light Water Reactors CX(s) Applied: B1.31 Date: 04/25/2013 Location(s): Idaho Offices(s): Idaho Operations Office

186

CX-010059: Categorical Exclusion Determination  

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

High Density Fuel Material for Light Water Reactors (LWRs) CX(s) Applied: B1.31 Date: 01/14/2013 Location(s): Idaho Offices(s): Nuclear Energy

187

CX-004476: Categorical Exclusion Determination | Department of...  

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

Exclusion Determination Testing of Indian Coal in a Transport Reactor Integrated Gasification (TRIG) System CX(s) Applied: B3.6 Date: 11182010 Location(s): Grand Forks, North...

188

CX-004662: Categorical Exclusion Determination | Department of...  

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

Exclusion Determination Testing of Chinese Coal in a Transport Reactor Integrated Gasification (TRIG) System CX(s) Applied: B3.6 Date: 12092010 Location(s): Grand Forks, North...

189

CX-009633: Categorical Exclusion Determination | Department of...  

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

CX-009633: Categorical Exclusion Determination Upgrade to the Concrete Masonry Unit (CMU) Wall at Test Reactor Area (TRA)-670 CX(s) Applied: B2.5 Date: 11292012...

190

CX-008739: Categorical Exclusion Determination | Department of...  

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

of Radiation Safety and Radiological Monitoring Systems for the Ohio State University Nuclear Reactor CX(s) Applied: B2.2 Date: 05212012 Location(s): Idaho Offices(s): Idaho...

191

Categorical Exclusion (CX) Determinations By Date | Department...  

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

New Reactor Technology for Hydro processing Bio-oils to Produce Gasoline, Diesel and Jet Fuel CX(s) Applied: A9, B3.6 Date: 05022011 Location(s): Maryland Office(s): Energy...

192

CX-009221: Categorical Exclusion Determination | Department of...  

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

Exclusion Determination CX-009221: Categorical Exclusion Determination Rerouting of Power Lines at 100-D Reactor in Support of Site Remediation CX(s) Applied: B4.13 Date: 09...

193

CX-005748: Categorical Exclusion Determination | Department of...  

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

Determination CX-005748: Categorical Exclusion Determination New Reactor Technology for Hydro processing Bio-oils to Produce Gasoline, Diesel and Jet Fuel CX(s) Applied: A9, B3.6...

194

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

195

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...

196

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

197

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...

198

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

199

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

200

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

Note: This page contains sample records for the topic "reactor cxs applied" 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

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

202

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

203

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

204

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

205

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

206

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

207

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

208

Applied and Computational Mathematics Division  

Science Conference Proceedings (OSTI)

Applied and Computational Mathematics Division. Topic Areas. Mathematics; Scientific Computing; Visualization; Quantum Computing. ...

2013-05-09T23:59:59.000Z

209

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

210

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

211

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

212

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

213

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

214

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

215

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

216

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

217

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

218

A preliminary investigation of the Topaz II reactor as a lunar surface power supply  

DOE Green Energy (OSTI)

Reactor power supplies offer many attractive characteristics for lunar surface applications. The Topaz II reactor resulted from an extensive development program in the former Soviet Union. Flight quality reactor units remain from this program and are currently under evaluation in the United States. This paper examines the potential for applying the Topaz II, originally developed to provide spacecraft power, as a lunar surface power supply.

Polansky, G.F. [Sandia National Labs., Albuquerque, NM (United States); Houts, M.G. [Los Alamos National Lab., NM (United States)

1995-12-31T23:59:59.000Z

219

Plateout Phenomena in Direct-Cycle High Temperature Gas-Cooled Reactors  

Science Conference Proceedings (OSTI)

The plateout of condensable radionuclides in the primary coolant circuits of high-temperature gas-cooled reactors (HTGRs) -- particularly direct-cycle HTGRs -- has significant design, operations and maintenance (O&M), and safety implications. This report reviews and evaluates the available international information on plateout phenomena, specifically as it applies to the gas turbine-modular helium reactor (GT-MHR) and the pebble bed modular reactor (PBMR).

2002-06-26T23:59:59.000Z

220

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

Note: This page contains sample records for the topic "reactor cxs applied" 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

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...

222

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

223

Constructal method to optimize solar thermochemical reactor design  

SciTech Connect

The objective of this study is the geometrical optimization of a thermochemical reactor, which works simultaneously as solar collector and reactor. The heat (concentrated solar radiation) is supplied on a small peripheral surface and has to be dispersed in the entire reactive volume in order to activate the reaction all over the material. A similarity between this study and the point to volume problem analyzed by the constructal approach (Bejan, 2000) is evident. This approach was successfully applied to several domains, for example for the coupled mass and conductive heat transfer (Azoumah et al., 2004). Focusing on solar reactors, this work aims to apply constructal analysis to coupled conductive and radiative heat transfer. As a first step, the chemical reaction is represented by a uniform heat sink inside the material. The objective is to optimize the reactor geometry in order to maximize its efficiency. By using some hypothesis, a simplified solution is found. A parametric study provides the influence of different technical and operating parameters on the maximal efficiency and on the optimal shape. Different reactor designs (filled cylinder, cavity and honeycomb reactors) are compared, in order to determine the most efficient structure according to the operating conditions. Finally, these results are compared with a CFD model in order to validate the assumptions. (author)

Tescari, S.; Mazet, N. [PROMES-CNRS, Rambla de la Thermodynamique, Tecnosud, 66100 Perpignan (France); Neveu, P. [PROMES-CNRS, Rambla de la Thermodynamique, Tecnosud, 66100 Perpignan (France); Universite de Perpignan Via Domitia, 52 Avenue Paul Alduy, 66860 Perpignan (France)

2010-09-15T23:59:59.000Z

224

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

225

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

226

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

227

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

228

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...

229

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

230

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

231

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

232

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

233

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

234

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

235

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

236

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

237

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

238

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

239

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

240

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

Note: This page contains sample records for the topic "reactor cxs applied" 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

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

242

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

243

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

244

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

245

Instrumentation to Enhance Advanced Test Reactor Irradiations  

SciTech Connect

The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

J. L. Rempe; D. L. Knudson; K. G. Condie; J. E. Daw; S. C. Taylor

2009-09-01T23:59:59.000Z

246

Application of the LEPRICON Methodology to the Arkansas Nuclear One, Unit 1 Reactor  

Science Conference Proceedings (OSTI)

Applying the new LEPRICON dosimetry methodology to irradiation data from the ANO-1 reactor greatly reduced uncertainties in pressure vessel fluence estimates. Accurate estimates will enable safety engineers to better predict reactor lifetimes and will eliminate the need for overly conservative safety margins.

1986-02-19T23:59:59.000Z

247

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

248

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

249

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

250

Applied Optoelectronics | Open Energy Information  

Open Energy Info (EERE)

optical semiconductor devices, packaged optical components, optical subsystems, laser transmitters, and fiber optic transceivers. References Applied Optoelectronics1...

251

Prediction of the reactor antineutrino flux for the Double Chooz experiment  

E-Print Network (OSTI)

This thesis benchmarks the deterministic lattice code, DRAGON, against data, and then applies this code to make a prediction for the antineutrino flux from the Chooz BI and B2 reactors. Data from the destructive assay of ...

Jones, Christopher LaDon

2012-01-01T23:59:59.000Z

252

Optimal design of a high pressure organometallic chemical vapor deposition reactor  

Science Conference Proceedings (OSTI)

A team composed of material scientists, physicists, and applied mathematicians have used computer simulations as a fundamental design tool in developing a new prototype High Pressure Organometallic Chemical Vapor Deposition (HPOMCVD) reactor for use ...

K. J. Bachmann; H. T. Banks; C. Höpfner; G. M. Kepler; S. Lesure; S. D. Mccall; J. S. Scroggs

1999-04-01T23:59:59.000Z

253

Analysis and numerical optimization of gas turbine space power systems with nuclear fission reactor heat sources  

Science Conference Proceedings (OSTI)

A new three objective optimization technique is developed and applied to find the operating conditions for fission reactor heated Closed Cycle Gas Turbine (CCGT) space power systems at which maximum efficiency, minimum radiator area, and minimum total ...

Albert J. Juhasz / Jerzy Sawicki

2005-01-01T23:59:59.000Z

254

Analysis of strategies for improving uranium utilization in pressurized water reactors  

E-Print Network (OSTI)

Systematic procedures have been devised and applied to evaluate core design and fuel management strategies for improving uranium utilization in Pressurized Water Reactors operated on a once-through fuel cycle. A principal ...

Sefcik, Joseph A.

1981-01-01T23:59:59.000Z

255

The selective use of thorium and heterogeneity in uranium-efficient pressurized water reactors  

E-Print Network (OSTI)

Systematic procedures have been developed and applied to assess the uranium utilization potential of a broad range of options involving the selective use of thorium in Pressurized Water Reactors (PWRs) operating on the ...

Kamal, Altamash

1982-01-01T23:59:59.000Z

256

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

257

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

258

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

259

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

260

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

Note: This page contains sample records for the topic "reactor cxs applied" 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

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

262

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

263

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

264

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

265

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

266

NFRC Procedures for Applied Films  

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

Applied Films Applied Films Last update: 12/10/2013 07:29 PM NFRC now has a procedure for adding applied films to substrates in Optics5 and importing those applied film constructions into WINDOW5 to be used in a whole product calculation. The information presented below is provided to help simulators with this process. Feel free to contact us at WINDOWHelp@lbl.gov with questions or comments. NFRC Applied Film Procedure Applied Film Procedures (approved by NFRC) (PDF file) Approved Applied Film List (IGDB 33.0) (PDF file) NFRC Laminate Procedure Training Powerpoint with Examples (This Powerpoint presentation was used in the NFRC web based training sessions in December 2006 and January 2007) PowerPoint Presentation (PPT file) PowerPoint Presentation (PDF file) Help and Troubleshooting

267

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

268

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

269

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,...

270

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

271

Dispersion effects on membrane reactor performance  

SciTech Connect

A mathematical model has been developed that predicts the effects of design parameters, operating variables and physical properties on the performance of a membrane reactor with a permselective wall. The model consists of the full set of partial differential equations that describe the conservation of mass, momentum and chemical species, coupled with chemical kinetics and appropriate boundary conditions for the physical problem. The solution of this system is obtained by a finite-volume technique. The model was applied to study the dehydrogenation of cyclohexane. Two membrane types in tubular form were studied: a selective porous glass with low gas permeabilities and a porous alumina with very high gas permeabilities. It is concluded that gas separation and reactor performance are strongly influenced by dispersion effects only in the latter membrane reactor, while in both cases radial concentration profiles do not correspond to those obtained with plug flow. Therefore, simulations of this type of problem should be based on complex dispersion models rather than the existing ideal plug-flow ones.

Koukou, M.K.; Papayannakos, N.; Markatos, N.C. [National Technical Univ. of Athens (Greece). Dept. of Chemical Engineering

1996-09-01T23:59:59.000Z

272

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.

273

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

274

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

275

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

276

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

277

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

278

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...

279

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

280

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

Note: This page contains sample records for the topic "reactor cxs applied" from the National Library of EnergyBeta (NLEBeta).
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to obtain the most current and comprehensive results.


281

PROCESSING OF MOLTEN SALT POWER REACTOR FUEL  

SciTech Connect

ABS> Fuel reprocessing methods are being investigated for molten salt nuclear reactors which use LiF--BeF/sub 2/ salt as a solvent for UF/sub 4/ and ThF/sub 4/. A liquid HF dissolution procedure coupled with fluorination has been developed for recovery of the uranium and LiF- BeF/sub 2/ solvent salt which is highly enriched in Li/sup 7/. The recovered salt is decontaminated in the process from the major reactor poisons; namely, rare earths and neptunium. A brief investigation of alternate methods, including oxide precipitation, partial freezing, and metal reduction, indicated that such methods may give some separation of the solvent salt from reactor poisons, but they do not appear to be sufficiently quantitative for a simple processing operation. Solubilities of LiF and BeF/sub 2/ in aqueous 70t0 100% HF are presented. The BeF/sub 2/ solubility is appreciably increased in the presence of water and large amounts of LiF. Salt solubilities of 150 g/liter are attainable. Tracer experiments indicate that rare earth solubilities, relative to LiF-- BeF/sub 2/ solvent salt solubility, increase from about 10/sup -4/ mole% in 98% HF to 0.003 mole% in 80% HF. Fluorination of uranium from LiF--BeF/sub 2/ salt was demonstrated. This appears feasible also for the recovery of the relatively small ccncentration of uranium produced in the LiF- BeF/sub 2/ThF/sub 4/ blanket. A proposed chemical flowsheet is presented on the basis of this exploratory work as applied to the semicontinuous processing of a 600 Mw power reactor. (auth)

Campbell, D.O.; Cathers, G.I.

1959-04-01T23:59:59.000Z

282

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

283

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

284

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

285

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

286

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

287

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

288

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

289

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

290

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

291

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

292

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

293

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

294

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

295

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

296

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

297

Statement on Defense Nuclear Nonproliferation and Naval Reactors Activities  

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

Defense Nuclear Nonproliferation and Naval Reactors Activities Defense Nuclear Nonproliferation and Naval Reactors Activities before the House Committee on Appropriations Subcommittee on Energy & Water Development | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Congressional Testimony > Statement on Defense Nuclear

298

Statement on Defense Nuclear Nonproliferation and Naval Reactors Activities  

National Nuclear Security Administration (NNSA)

Defense Nuclear Nonproliferation and Naval Reactors Activities Defense Nuclear Nonproliferation and Naval Reactors Activities before the House Committee on Appropriations Subcommittee on Energy & Water Development | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Congressional Testimony > Statement on Defense Nuclear

299

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

300

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

Note: This page contains sample records for the topic "reactor cxs applied" 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

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

302

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

303

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

304

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

305

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

306

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

307

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

308

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...

309

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

310

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

311

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

312

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

313

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

314

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

315

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

316

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...

317

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...

318

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...

319

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...

320

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...

Note: This page contains sample records for the topic "reactor cxs applied" 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

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...

322

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...

323

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.

324

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.

325

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.

326

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...

327

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

328

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

329

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

330

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

331

The Consortium for Advanced Simulation of Light Water Reactors  

SciTech Connect

The Consortium for Advanced Simulation of Light Water Reactors (CASL) is a DOE Energy Innovation Hub for modeling and simulation of nuclear reactors. It brings together an exceptionally capable team from national labs, industry and academia that will apply existing modeling and simulation capabilities and develop advanced capabilities to create a usable environment for predictive simulation of light water reactors (LWRs). This environment, designated as the Virtual Environment for Reactor Applications (VERA), will incorporate science-based models, state-of-the-art numerical methods, modern computational science and engineering practices, and uncertainty quantification (UQ) and validation against data from operating pressurized water reactors (PWRs). It will couple state-of-the-art fuel performance, neutronics, thermal-hydraulics (T-H), and structural models with existing tools for systems and safety analysis and will be designed for implementation on both today's leadership-class computers and the advanced architecture platforms now under development by the DOE. CASL focuses on a set of challenge problems such as CRUD induced power shift and localized corrosion, grid-to-rod fretting fuel failures, pellet clad interaction, fuel assembly distortion, etc. that encompass the key phenomena limiting the performance of PWRs. It is expected that much of the capability developed will be applicable to other types of reactors. CASL's mission is to develop and apply modeling and simulation capabilities to address three critical areas of performance for nuclear power plants: (1) reduce capital and operating costs per unit energy by enabling power uprates and plant lifetime extension, (2) reduce nuclear waste volume generated by enabling higher fuel burnup, and (3) enhance nuclear safety by enabling high-fidelity predictive capability for component performance.

Ronaldo Szilard; Hongbin Zhang; Doug Kothe; Paul Turinsky

2011-10-01T23:59:59.000Z

332

The application of modern safety criteria to restarting and operating the USDOE K-Reactor  

SciTech Connect

The United States Department of Energy's (USDOE's) K-reactor, a defense production reactor located at the Savannah River Site in Aiken, South Carolina, was shut down in the summer of 1988 for safety upgrades to bring it into conformance with modern safety standards prior to restart. Over the course of the succeeding four years, all aspects of the 35-year old reactor, including hardware, operations, and analysis, were upgraded to ensure that the reactor could operate safely according to standards similar to those applied to modern nuclear reactors. This paper describes the decision making processes by which issues were identified, priorities assigned, and analysis improved to enhance reactor safety. Special emphasis is given to the probabilistic risk assessment (PRA) decision making processes used to quantify the risks and consequences of operating the K-reactor, the analytical hierarchy process (AHP) used to identify key phenomena, and modifications made to the RELAP5 computer code to make it applicable to K-reactor analysis. The success of the project was demonstrated when the K-reactor was restarted in the summer of 1992.

Dimenna, R.A.; Taylor, G.A.; Brandyberry, M.D.

1993-01-01T23:59:59.000Z

333

The application of modern safety criteria to restarting and operating the USDOE K-Reactor  

SciTech Connect

The United States Department of Energy`s (USDOE`s) K-reactor, a defense production reactor located at the Savannah River Site in Aiken, South Carolina, was shut down in the summer of 1988 for safety upgrades to bring it into conformance with modern safety standards prior to restart. Over the course of the succeeding four years, all aspects of the 35-year old reactor, including hardware, operations, and analysis, were upgraded to ensure that the reactor could operate safely according to standards similar to those applied to modern nuclear reactors. This paper describes the decision making processes by which issues were identified, priorities assigned, and analysis improved to enhance reactor safety. Special emphasis is given to the probabilistic risk assessment (PRA) decision making processes used to quantify the risks and consequences of operating the K-reactor, the analytical hierarchy process (AHP) used to identify key phenomena, and modifications made to the RELAP5 computer code to make it applicable to K-reactor analysis. The success of the project was demonstrated when the K-reactor was restarted in the summer of 1992.

Dimenna, R.A.; Taylor, G.A.; Brandyberry, M.D.

1993-06-01T23:59:59.000Z

334

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

335

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

336

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

337

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

338

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

339

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

340

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,

Note: This page contains sample records for the topic "reactor cxs applied" 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

Vehicle Technologies Office: Applied Battery Research  

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

Applied Battery Research to someone by E-mail Share Vehicle Technologies Office: Applied Battery Research on Facebook Tweet about Vehicle Technologies Office: Applied Battery...

342

BNL | Accelerators for Applied Research  

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

Accelerators for Applied Research Accelerators for Applied Research Brookhaven National Lab operates several accelerator facilities dedicated to applied research. These facilities directly address questions and concerns on a tremendous range of fields, including medical imaging, cancer therapy, computation, and space exploration. Leading scientists lend their expertise to these accelerators and offer crucial assistant to collaborating researchers, pushing the limits of science and technology. Interested in gaining access to these facilities for research? See the contact number listed for each facility. RHIC tunnel Brookhaven Linac Isotope Producer The Brookhaven Linac Isoptope Producer (BLIP)-positioned at the forefront of research into radioisotopes used in cancer treatment and diagnosis-produces commercially unavailable radioisotopes for use by the

343

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

344

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

345

CRC handbook of applied thermodynamics  

Science Conference Proceedings (OSTI)

The emphasis of this book is on applied thermodynamics, featuring the stage of development of a process rather than the logical development of thermodynamic principles. It is organized according to the types of problems encountered in industry, such as probing research, process assessment, and process development. The applied principles presented can be used in most areas of industry including oil and gas production and processing, chemical processing, power generation, polymer production, food processing, synthetic fuels production, specialty chemicals and pharmaceuticals production, bioengineered processes, etc.

Palmer, D.A. (Amoco Chemical Corp., Naperville, IL (USA). Research and Development Dept.)

1987-01-01T23:59:59.000Z

346

83rd Annual Scientific Conference of the International Association of Applied Mathematics and Mechanics  

E-Print Network (OSTI)

Orthogonal De- composition (POD) of boiling water reactor (BWR) fuel assemblies (FAs) applied to introductory Dortmund), Manuela Sander (Universität Rostock) S3.1: Modeling material failure Tue, 13:30­15:30 Chair at failure Abstract 14:30 ­ 14:50: Ingo Scheider, Yongjun Chen, Norbert Huber (Helmholtz-Zentrum Geest- hacht

Kohlenbach, Ulrich

347

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

348

Transforming criticality control methods for EBR-II fuel handling during reactor decommissioning  

SciTech Connect

A review of the Department of Energy (DOE) request to decommission the Experimental Breeder Reactor-II (EBR-II) was conducted in order to develop a scope of work and analysis method for performing the safety review of the facility. Evaluation of the current national standards, DOE orders, EBR-II nuclear safeguards and criticality control practices showed that a decommissioning policy for maintaining criticality safety during a long term fuel transfer process did not exist. The purpose of this research was to provide a technical basis for transforming the reactor from an instrumentation and measurement controlled system to a system that provides both physical constraint and administrative controls to prevent criticality accidents. Essentially, this was done by modifying the reactor core configuration, reactor operations procedures and system instrumentation to meet the safety practices of ANS-8.1-1983. Subcritical limits were determined by applying established liquid metal reactor methods for both the experimental and computational validations.

Eberle, C.S.; Dean, E.M.; Angelo, P.L.

1995-12-31T23:59:59.000Z

349

Human Reliability Analysis for Small Modular Reactors  

Science Conference Proceedings (OSTI)

Because no human reliability analysis (HRA) method was specifically developed for small modular reactors (SMRs), the application of any current HRA method to SMRs represents tradeoffs. A first- generation HRA method like THERP provides clearly defined activity types, but these activity types do not map to the human-system interface or concept of operations confronting SMR operators. A second- generation HRA method like ATHEANA is flexible enough to be used for SMR applications, but there is currently insufficient guidance for the analyst, requiring considerably more first-of-a-kind analyses and extensive SMR expertise in order to complete a quality HRA. Although no current HRA method is optimized to SMRs, it is possible to use existing HRA methods to identify errors, incorporate them as human failure events in the probabilistic risk assessment (PRA), and quantify them. In this paper, we provided preliminary guidance to assist the human reliability analyst and reviewer in understanding how to apply current HRA methods to the domain of SMRs. While it is possible to perform a satisfactory HRA using existing HRA methods, ultimately it is desirable to formally incorporate SMR considerations into the methods. This may require the development of new HRA methods. More practicably, existing methods need to be adapted to incorporate SMRs. Such adaptations may take the form of guidance on the complex mapping between conventional light water reactors and small modular reactors. While many behaviors and activities are shared between current plants and SMRs, the methods must adapt if they are to perform a valid and accurate analysis of plant personnel performance in SMRs.

Ronald L. Boring; David I. Gertman

2012-06-01T23:59:59.000Z

350

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

351

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

352

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

353

California Energy Commission Apply Today!  

E-Print Network (OSTI)

photovoltaic project in the future. Peak Demand Savings: 95 kW Energy Savings: 1,510,849 kWh Annual Energy CostCalifornia Energy Commission Apply Today! "The College implemented all of the recommended projects Programs Office (916) 654-4147 pubprog@energy.state.ca.us "CEC financing allowed us to install many

354

implementing bioenergy applied research & development  

E-Print Network (OSTI)

1 A Northern Centre for Renewable Energy implementing bioenergy applied research & development plant measures to become carbon neutral and operate on renewable energy. UNBC is uniquely positioned for Climate Solutions, and UNBC. The Green University Centre will be a model of energy efficiency

Northern British Columbia, University of

355

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

356

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

357

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

358

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.

Ritz, W.C.

1975-12-01T23:59:59.000Z

359

Particle Bed Reactor scaling relationships  

DOE Green Energy (OSTI)

Scaling relationships for Particle Bed Reactors (PBRs) are discussed. The particular applications are short duration systems, i.e., for propulsion or burst power. Particle Bed Reactors can use a wide selection of different moderators and reflectors and be designed for such a wide range of power and bed power densities. Additional design considerations include the effect of varying the number of fuel elements, outlet Mach number in hot gas channel, etc. All of these variables and options result in a wide range of reactor weights and performance. Extremely light weight reactors (approximately 1 kg/MW) are possible with the appropriate choice of moderator/reflector and power density. Such systems are very attractive for propulsion systems where parasitic weight has to be minimized.

Slovik, G.; Araj, K.; Horn, F.L.; Ludewig, H.; Benenati, R.

1987-01-01T23:59:59.000Z

360

REACTOR SYSTEM AND CONTROL VALVE  

DOE Patents (OSTI)

Valves have been developed for controlling the flow of gaseous fluid through a passage or conduit. The valves have particular application in the cooling systems of gas; cooled reactors. (R.J.S.)

Fortescue, P.; Rickard, C.; Rose, D.

1963-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor cxs applied" 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

Reactor core isolation cooling system  

DOE Patents (OSTI)

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

Cooke, F.E.

1992-12-08T23:59:59.000Z

362

Reactor core isolation cooling system  

DOE Patents (OSTI)

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

Cooke, Franklin E. (San Jose, CA)

1992-01-01T23:59:59.000Z

363

Reactor operation environmental information document  

Science Conference Proceedings (OSTI)

This volume is a reactor operation environmental information document for the Savannah River Plant. Topics include meteorology, surface hydrology, transport, environmental impacts, and radiation effects. 48 figs., 56 tabs. (KD)

Bauer, L.R.; Hayes, D.W.; Hunter, C.H.; Marter, W.L.; Moyer, R.A.

1989-12-01T23:59:59.000Z

364

Solid State Reactor Final Report  

DOE Green Energy (OSTI)

The Solid State Reactor (SSR) is an advanced reactor concept designed to take advantage of Oak Ridge National Laboratory's (ORNL's) recently developed graphite foam that has enhanced heat transfer characteristics and excellent high-temperature mechanical properties, to provide an inherently safe, self-regulated, source of heat for power and other potential applications. This work was funded by the U.S. Department of Energy's Nuclear Energy Research Initiative (NERI) program (Project No. 99-064) from August 1999 through September 30, 2002. The initial concept of utilizing the graphite foam as a basis for developing an advanced reactor concept envisioned that a suite of reactor configurations and power levels could be developed for several different applications. The initial focus was looking at the reactor as a heat source that was scalable, independent of any heat removal/power conversion process. These applications might include conventional power generation, isotope production and destruction (actinides), and hydrogen production. Having conducted the initial research on the graphite foam and having performed the scoping parametric analyses from neutronics and thermal-hydraulic perspectives, it was necessary to focus on a particular application that would (1) demonstrate the viability of the overall concept and (2) require a reasonably structured design analysis process that would synthesize those important parameters that influence the concept the most as part of a feasible, working reactor system. Thus, the application targeted for this concept was supplying power for remote/harsh environments and a design that was easily deployable, simplistic from an operational standpoint, and utilized the new graphite foam. Specifically, a 500-kW(t) reactor concept was pursued that is naturally load following, inherently safe, optimized via neutronic studies to achieve near-zero reactivity change with burnup, and proliferation resistant. These four major areas of research were undertaken: (1) establishing the design and safety-related basis via neutronic and reactor control assessments with the graphite foam as heat transfer medium; (2) evaluating the thermal performance of the graphite foam for heat removal, reactor stability, reactor operations, and overall core thermal characteristics; (3) characterizing the physical properties of the graphite foam under normal and irradiated conditions to determine any effects on structure, dimensional stability, thermal conductivity, and thermal expansion; and (4) developing a power conversion system design to match the reactor operating parameters.

Mays, G.T.

2004-03-10T23:59:59.000Z

365

REACTOR CONTROL MECHANISM  

DOE Patents (OSTI)

A quick-releasing mechanism is described which may be used to rapidiy drop a device supported from beneath during normal use, such as a safety rod in a nuclear reactor. In accordance with this invention an electrical control signal, such as may be provided by radiation detection or other alarm condition sensing devices, is delivered to an electromagnetic solenoid, the armature of which is coupled to an actuating mechanism. The solenoid is energized when the mechanism is in its upper or cocked position. In such position, the mechanism engages a plurality of retaining balls, forcing them outward into engagement with a shoulder or recess in a corresponding section of a tubular extension on the upheld device. When the control signal to the solenoid suddenly ceases, the armature drops out, allowing the actuating mechanism to move slightly but rapidly under the force of a compressed spring. The weight of the device will urge the balls inward against a beveled portion of the actuating mechanism and away from the engaging section on the tubular extension, thus allowing the upheld device to fall freely under the influence of gravity.

Lane, J.A.; Engberg, R.E.; Welch, J.M.

1959-05-12T23:59:59.000Z

366

Reactor pressure vessel nozzle  

DOE Patents (OSTI)

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

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

1994-10-04T23:59:59.000Z

367

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

368

Reactor shroud joint  

DOE Patents (OSTI)

A shroud for a nuclear reactor is described. In one embodiment, the shroud includes first and second shroud sections, and each shroud section includes a substantially cylindrical main body having a first end and a second end. With respect to each shroud section, a flange is located at the main body first end, and the flange has a plurality of bolt openings therein and a plurality of scalloped regions. The first shroud section is welded to the second shroud section, and at least some of the bolt openings in the first shroud section flange align with respective bolt openings in the second shroud section flange. In the event that the onset of inter-granular stress corrosion cracking is ever detected in the weld between the shroud section, bolts are inserted through bolt openings in the first shroud section flange and through aligned bolt openings the second shroud section flange. Each bolt, in one embodiment, has a shank section and first and second threaded end sections. Nuts are threadedly engaged to the threaded end sections and tightened against the respective flanges.

Ballas, Gary J. (San Jose, CA); Fife, Alex Blair (San Jose, CA); Ganz, Israel (San Jose, CA)

1998-01-01T23:59:59.000Z

369

Reactor shroud joint  

DOE Patents (OSTI)

A shroud for a nuclear reactor is described. In one embodiment, the shroud includes first and second shroud sections, and each shroud section includes a substantially cylindrical main body having a first end and a second end. With respect to each shroud section, a flange is located at the main body first end, and the flange has a plurality of bolt openings therein and a plurality of scalloped regions. The first shroud section is welded to the second shroud section, and at least some of the bolt openings in the first shroud section flange align with respective bolt openings in the second shroud section flange. In the event that the onset of inter-granular stress corrosion cracking is ever detected in the weld between the shroud section, bolts are inserted through bolt openings in the first shroud section flange and through aligned bolt openings the second shroud section flange. Each bolt, in one embodiment, has a shank section and first and second threaded end sections. Nuts are threadedly engaged to the threaded end sections and tightened against the respective flanges. 4 figs.

Ballas, G.J.; Fife, A.B.; Ganz, I.

1998-04-07T23:59:59.000Z

370

Novel Catalytic Membrane Reactors  

DOE Green Energy (OSTI)

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

Stuart Nemser, PhD

2010-10-01T23:59:59.000Z

371

Boiling Water Reactor Chemistry Performance Monitoring Update--2007 Edition  

Science Conference Proceedings (OSTI)

Successful operation of a nuclear plant demands careful monitoring of water chemistry, particularly in BWRs, where control of iron and copper in the reactor coolant is essential. Since the advent of hydrogen water chemistry (HWC), plant operators have successfully applied other chemistry regimes such as noble metal chemical addition (NMCA) and zinc injection to control radiation fields and provide additional mitigation for intergranular stress corrosion cracking (IGSCC). This report compiles recent BWR p...

2007-12-12T23:59:59.000Z

372

Expert system for online surveillance of nuclear reactor coolant pumps  

DOE Patents (OSTI)

An expert system for online surveillance of nuclear reactor coolant pumps. This system provides a means for early detection of pump or sensor degradation. Degradation is determined through the use of a statistical analysis technique, sequential probability ratio test, applied to information from several sensors which are responsive to differing physical parameters. The results of sequential testing of the data provide the operator with an early warning of possible sensor or pump failure.

Gross, Kenny C. (Bolingbrook, IL); Singer, Ralph M. (Naperville, IL); Humenik, Keith E. (Columbia, MD)

1993-01-01T23:59:59.000Z

373

Expert system for online surveillance of nuclear reactor coolant pumps  

DOE Patents (OSTI)

This report describes an expert system for online surveillance of nuclear reactor coolant pumps. This system provides a means for early detection of pump or sensor degradation. Degradation is determined through the use of a statistical analysis technique, sequential probability ratio test, applied to information from several sensors which are responsive to differing physical parameters. The results of sequential testing of the data provide the operator with an early warning of possible sensor or pump failure.

Gross, K.C.; Singer, R.M.; Humenik, K.E.

1992-12-31T23:59:59.000Z

374

Automatic safety rod for reactors  

DOE Patents (OSTI)

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

Germer, John H. (San Jose, CA)

1988-01-01T23:59:59.000Z

375

Possible fusion reactor. [Movable plasmas  

SciTech Connect

A scheme to improve performance characteristics of a tokamak-type fusion reactor is proposed. Basically, the tokamak-type plasma could be moved around so that the plasma could be heated by compression, brought to the region where the blanket surrounds the plasma, and moved so as to keep wall loading below the acceptable limit. This idea should be able to help to economize a fusion reactor.

Yoshikawa, S.

1976-05-01T23:59:59.000Z

376

MOLTEN FLUORIDE NUCLEAR REACTOR FUEL  

DOE Patents (OSTI)

Molten-salt reactor fuel compositions consisting of mixtures of fluoride salts are reported. In its broadest form, the composition contains an alkali fluoride such as sodium fluoride, zirconium tetrafluoride, and a uranium fluoride, the latter being the tetrafluoride or trifluoride or a mixture of the two. An outstanding property of these fuel compositions is a high coeffieient of thermal expansion which provides a negative temperature coefficient of reactivity in reactors in which they are used.

Barton, C.J.; Grimes, W.R.

1960-01-01T23:59:59.000Z

377

Fast quench reactor and method  

DOE Patents (OSTI)

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

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

1998-01-01T23:59:59.000Z

378

Fast quench reactor and method  

DOE Patents (OSTI)

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

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

2002-01-01T23:59:59.000Z

379

Physics problems of thermonuclear reactors  

SciTech Connect

From topical conference on energy; Philadelphia, Pennsylvania, USA (Nov 1973). A problem common to all controlled fusion reactors is that of the burning deuterium --tritium fuel under conditions of plasma confinement which approach the ideal limit as nearly as possible. After ignitlon, the balance between alpha- particle energy deposition and plasma losses (radiation plus thermal and particle diffusion) determines the stability or instability of the burn in toroidal systems. Tokamak systems are described both with unstable, injection-regulated burn cycles and stabilized steady-state burn conditions. In the theta-pinch reactor an unstable burn occurs, some, what regulated by high-beta plasma expansion, which is quenched by a programmed plasma decompression. The plasma expansion during the constant-pressure burn provides direct'' conversion of plasma thermonuclear heat to electrical output, in addition to the electrical power derived from the neutron energy through conventional thermal conversion equipment. The open-ended mirror reactor is characterized by a direct conversion system for recovering end-loss plasma energy and converting it to electrical energy for reinjection into the plasma. This allows a favorable reactor energy balance and an amplification factor Q (= thermonuclear energy output/injected plasma energy) which is compatible with classical collisional losses. For the three reactor types considered the ramifications of burn and confinement conditions for reactor configuration, energy balance, economy, fuel handling, materials problems, and environmentalradiological factors are considered. (auth)

Ribe, F.L.

1974-01-01T23:59:59.000Z

380

Evaluation of Torsatrons as reactors  

SciTech Connect

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

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

1994-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor cxs applied" 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

Development of Underwater Laser Cladding and Underwater Laser Seal Welding Techniques for Reactor Components (II)  

SciTech Connect

Stress corrosion cracking (SCC) is one of the major reasons to reduce the reliability of aged reactor components. Toshiba has been developing underwater laser welding onto surface of the aged components as maintenance and repair techniques. Because most of the reactor internal components to apply this underwater laser welding technique have 3-dimensional shape, effect of welding positions and welded shapes are examined and presented in this report. (authors)

Masataka Tamura; Shohei Kawano; Wataru Kouno; Yasushi Kanazawa [Toshiba Corporation (Japan)

2006-07-01T23:59:59.000Z

382

Propellant actuated nuclear reactor steam depressurization valve  

DOE Patents (OSTI)

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

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

1992-01-01T23:59:59.000Z

383

Light water reactor lower head failure analysis  

SciTech Connect

This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response.

Rempe, J.L.; Chavez, S.A.; Thinnes, G.L. [EG and G Idaho, Inc., Idaho Falls, ID (United States)] [and others

1993-10-01T23:59:59.000Z

384

Integrated systems analysis of the PIUS reactor  

SciTech Connect

Results are presented of a systems failure analysis of the PIUS plant systems that are used during normal reactor operation and postulated accidents. This study was performed to provide the NRC with an understanding of the behavior of the plant. The study applied two diverse failure identification methods, Failure Modes Effects & Criticality Analysis (FMECA) and Hazards & Operability (HAZOP) to the plant systems, supported by several deterministic analyses. Conventional PRA methods were also used along with a scheme for classifying events by initiator frequency and combinations of failures. Principal results of this study are: (a) an extensive listing of potential event sequences, grouped in categories that can be used by the NRC, (b) identification of support systems that are important to safety, and (c) identification of key operator actions.

Fullwood, F.; Kroeger, P.; Higgins, J. [Brookhaven National Lab., Upton, NY (United States)] [and others

1993-11-01T23:59:59.000Z

385

Natural Fueling of a Tokamak Fusion Reactor  

E-Print Network (OSTI)

A natural fueling mechanism that helps to maintain the main core deuterium and tritium (DT) density profiles in a tokamak fusion reactor is discussed. In H-mode plasmas dominated by ion- temperature gradient (ITG) driven turbulence, cold DT ions near the edge will naturally pinch radially inward towards the core. This mechanism is due to the quasi-neutral heat flux dominated nature of ITG turbulence and still applies when trapped and passing kinetic electron effects are included. Fueling using shallow pellet injection or supersonic gas jets is augmented by an inward pinch of could DT fuel. The natural fueling mechanism is demonstrated using the three-dimensional toroidal electromagnetic gyrokinetic turbulence code GEM and is analyzed using quasilinear theory. Profiles similar to those used for conservative ITER transport modeling that have a completely flat density profile are examined and it is found that natural fueling actually reduces the linear growth rates and energy transport.

Wan, Weigang; Chen, Yang; Perkins, Francis W

2009-01-01T23:59:59.000Z

386

Neutron initiation probability in fast burst reactor  

SciTech Connect

Based on the probability balance of neutron random events in multiply system, the four random process of neutron in prompt super-critical is described and then the equation of neutron initiation probability W(r,E,{Omega},t) is deduced. On the assumption of static, slightly prompt super-critical and the two factorial approximation, the formula of the average probability of 'one' neutron is derived which is the same with the result derived from the point model. The MC simulation using point model is applied in Godiva- II and CFBR-II, and the simulation result of one neutron initiation is well consistent with the theory that the initiation probability of Godiva- II inverted commas CFBR-II burst reactor are 0.00032, 0.00027 respectively on the ordinary burst operation. (authors)

Liu, X.; Du, J.; Xie, Q.; Fan, X. [Inst. of Nuclear Physics and Chemistry, China Academy of Engineering Physics, No.64, Mianshan Road, Mianyang, Sichuan (China)

2012-07-01T23:59:59.000Z

387

Reactor operations: Brookhaven Medical Research Reactor, Brookhaven High Flux Beam Reactor. Informal report, June 1995  

Science Conference Proceedings (OSTI)

Part one of this report gives the operating history of the Brookhaven Medical Research Reactor for the month of June. Also included are the BMRR technical safety surveillance requirements record and the summary of BMRR irradiations for the month. Part two gives the operating histories of the Brookhaven High Flux Beam Reactor and the Cold Neutron Facility at HFBR for June. Also included are the HFBR technical safety surveillance requirements record and the summary of HFBR irradiations for the month.

NONE

1995-06-01T23:59:59.000Z

388

Reactor operations: Brookhaven Medical Research Reactor, Brookhaven High Flux Beam Reactor. Informal report, July 1995  

Science Conference Proceedings (OSTI)

Part one of this report gives the operating history for the Brookhaven Medical Research Reactor for the month of July. Also included are the BMRR technical safety surveillance requirements record and the summary of BMRR irradiations for the month. Part two gives the operating histories for the Brookhaven High Flux Beam Reactor and the Cold Neutron Source Facility for the month of July. Also included are the HFBR technical safety surveillance requirements record and the summary of HFBR irradiations for the month.

NONE

1995-07-01T23:59:59.000Z

389

Neutronic reactor thermal shield  

DOE Patents (OSTI)

1. The combination with a plurality of parallel horizontal members arranged in horizontal and vertical rows, the spacing of the members in all horizontal rows being equal throughout, the spacing of the members in all vertical rows being equal throughout; of a shield for a nuclear reactor comprising two layers of rectangular blocks through which the members pass generally perpendicularly to the layers, each block in each layer having for one of the members an opening equally spaced from vertical sides of the block and located closer to the top of the block than the bottom thereof, whereby gravity tends to make each block rotate about the associated member to a position in which the vertical sides of the block are truly vertical, the openings in all the blocks of one layer having one equal spacing from the tops of the blocks, the openings in all the blocks of the other layer having one equal spacing from the tops of the blocks, which spacing is different from the corresponding spacing in the said one layer, all the blocks of both layers having the same vertical dimension or length, the blocks of both layers consisting of relatively wide blocks and relatively narrow blocks, all the narrow blocks having the same horizontal dimension or width which is less than the horizontal dimension or width of the wide blocks, which is the same throughout, each layer consisting of vertical rows of narrow blocks and wide blocks alternating with one another, each vertical row of narrow blocks of each layer being covered by a vertical row of wide blocks of the other layer which wide blocks receive the same vertical row of members as the said each vertical row of narrow blocks, whereby the rectangular perimeters of each block of each layer is completely out of register with that of each block in the other layer.

Lowe, Paul E. (Blue Ash, OH)

1976-06-15T23:59:59.000Z

390

FUSED REACTOR FUELS  

DOE Patents (OSTI)

This invention relates to a nuciear reactor fuel composition comprising (1) from about 0.01 to about 50 wt.% based on the total weight of said composition of at least one element selected from the class consisting of uranium, thorium, and plutonium, wherein said eiement is present in the form of at least one component selected from the class consisting of oxides, halides, and salts of oxygenated anions, with components comprising (2) at least one member selected from the class consisting of (a) sulfur, wherein the sulfur is in the form of at least one entity selected irom the class consisting of oxides of sulfur, metal sulfates, metal sulfites, metal halosulfonates, and acids of sulfur, (b) halogen, wherein said halogen is in the form of at least one compound selected from the class of metal halides, metal halosulfonates, and metal halophosphates, (c) phosphorus, wherein said phosphorus is in the form of at least one constituent selected from the class consisting of oxides of phosphorus, metal phosphates, metal phosphites, and metal halophosphates, (d) at least one oxide of a member selected from the class consisting of a metal and a metalloid wherein said oxide is free from an oxide of said element in (1); wherein the amount of at least one member selected from the class consisting of halogen and sulfur is at least about one at.% based on the amount of the sum of said sulfur, halogen, and phosphorus atom in said composition; and wherein the amount of said 2(a), 2(b) and 2(c) components in said composition which are free from said elements of uranium, thorium, arid plutonium, is at least about 60 wt.% based on the combined weight of the components of said composition which are free from said elements of uranium, thorium, and plutonium. (AEC)

Mayer, S.W.

1962-11-13T23:59:59.000Z

391

Computational fluid dynamic modeling of fluidized-bed polymerization reactors  

SciTech Connect

Polyethylene is one of the most widely used plastics, and over 60 million tons are produced worldwide every year. Polyethylene is obtained by the catalytic polymerization of ethylene in gas and liquid phase reactors. The gas phase processes are more advantageous, and use fluidized-bed reactors for production of polyethylene. Since they operate so close to the melting point of the polymer, agglomeration is an operational concern in all slurry and gas polymerization processes. Electrostatics and hot spot formation are the main factors that contribute to agglomeration in gas-phase processes. Electrostatic charges in gas phase polymerization fluidized bed reactors are known to influence the bed hydrodynamics, particle elutriation, bubble size, bubble shape etc. Accumulation of electrostatic charges in the fluidized-bed can lead to operational issues. In this work a first-principles electrostatic model is developed and coupled with a multi-fluid computational fluid dynamic (CFD) model to understand the effect of electrostatics on the dynamics of a fluidized-bed. The multi-fluid CFD model for gas-particle flow is based on the kinetic theory of granular flows closures. The electrostatic model is developed based on a fixed, size-dependent charge for each type of particle (catalyst, polymer, polymer fines) phase. The combined CFD model is first verified using simple test cases, validated with experiments and applied to a pilot-scale polymerization fluidized-bed reactor. The CFD model reproduced qualitative trends in particle segregation and entrainment due to electrostatic charges observed in experiments. For the scale up of fluidized bed reactor, filtered models are developed and implemented on pilot scale reactor.

Rokkam, Ram [Ames Laboratory

2012-11-02T23:59:59.000Z

392

Numerical tools applied to power reactor noise analysis Christophe Demazie`re*, Imre Pazsit  

E-Print Network (OSTI)

in collaboration with the Ringhals Nuclear Power Plant, in order to give a fingerprint of the status of these applications, calculations performed using the code are compared with at-power plant measurements. Power to the operation of the plants at the uprated power level are the reduction of the safety margins

Demazière, Christophe

393

Advanced Nuclear Reactors | Department of Energy  

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

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

394

Microsoft Word - power_reactors_briggs.doc  

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

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

395

Advanced Reactor Technology Documents | Department of Energy  

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

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

396

BOILING WATER REACTOR TRANSIENT INSTABILITY STUDIES OF RINGHALS 1 REACTOR USING TRACE COUPLED WITH PARCS.  

E-Print Network (OSTI)

??Reactor plant design often incorporates data and insight ascertained from computer code simulations of plant dynamics and reactor core behavior. Increasing utilization of data gathered… (more)

Walls, Robert

2009-01-01T23:59:59.000Z

397

Rapid starting methanol reactor system  

DOE Patents (OSTI)

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

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

1984-01-01T23:59:59.000Z

398

International Research Reactor Decommissioning Project  

SciTech Connect

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

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

2008-01-15T23:59:59.000Z

399

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

400

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 pr...

Bolonkin, Alexander

2008-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor cxs applied" 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

DOE - Office of Legacy Management -- Elk River Reactor - MN 01  

Office of Legacy Management (LM)

Elk River Reactor - MN 01 FUSRAP Considered Sites Site: Elk River Reactor (MN.01 ) Eliminated from consideration under FUSRAP - Reactor was dismantled and decommissioned by 1974...

402

DEVELOPMENT OF ELECTRO-OPTICAL INSTRUMENTATION FOR REACTOR SAFETY STUDIES  

E-Print Network (OSTI)

OPTICAL INSTRUMENTATION FOR REACTOR SAFETY STUDIES B.T.OPTICAL INSTRUMENTATION FOR REACTOR SAFETY STUDIES B.T.instru­ mentation for reactor safety studies is described.

Turko, B.T.

2010-01-01T23:59:59.000Z

403

MULTIPHASE REACTOR MODELING FOR ZINC CHLORIDE CATALYZED COAL LIQUEFACTION  

E-Print Network (OSTI)

II. c. High Yield Batch Reactor Results. . . • .Objectives •. Slurry Reactors . . . . . . .A. StudiesSystems. D. Slurry Reactor Theory. General. . Application to

Joyce, Peter James

2011-01-01T23:59:59.000Z

404

Status of the Daya Bay Reactor Neutrino Oscillation Experiment  

E-Print Network (OSTI)

Status of the Daya Bay Reactor Neutrino OscillationCheng-Ju Lin The Daya Bay reactor neutrino experiment [1] isneutrinos from the nuclear reactors at different baselines.

Lin, Cheng-Ju Stephen

2011-01-01T23:59:59.000Z

405

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel...  

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

of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research...

406

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

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

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

407

Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation  

E-Print Network (OSTI)

Wood, Completely automated nuclear reactors for long- termhandbook: Nuclear physics and reactor theory. 1993. (121 p).The analysis of any nuclear reactor system is closely

Heidet, Florent

2010-01-01T23:59:59.000Z

408

Uncertainties in the Anti-neutrino Production at Nuclear Reactors  

E-Print Network (OSTI)

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

Djurcic, Zelimir

2009-01-01T23:59:59.000Z

409

Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors  

E-Print Network (OSTI)

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

Terrani, Kurt Amir

2010-01-01T23:59:59.000Z

410

Report of the APS Neutrino Study Reactor Working Group  

E-Print Network (OSTI)

the neutrino sector. Nuclear reactors have played a criticalof experiments at nuclear reactors in the future neutrinoto measure #13 at a nuclear reactor, a study of the physics

2004-01-01T23:59:59.000Z

411

Status of the Daya Bay Reactor Neutrino Oscillation Experiment  

E-Print Network (OSTI)

neutrinos from the nuclear reactors at different baselines.will be commissioned Nuclear Reactor Anti-neutrino detectorthe nuclear power complex has two pairs of reactor cores (

Lin, Cheng-Ju Stephen

2011-01-01T23:59:59.000Z

412

Strategies for Studying High Dose Irradiation Effects in Reactor ...  

Science Conference Proceedings (OSTI)

Cladding and structural materials in fast reactors and fusion reactors will reach 200 dpa, and concepts such as the Traveling Wave Reactor could top 500 dpa.

413

Hanford Railcars Make Final Stop at B Reactor: Move Enhances...  

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

Make Final Stop at B Reactor: Move Enhances Visitor Experience at Historic Reactor Hanford Railcars Make Final Stop at B Reactor: Move Enhances Visitor Experience at Historic...

414

REACTOR DEVELOPMENT PROGRAM, PROGRESS REPORT, MAY 1961  

SciTech Connect

General research and development on water-cooled and sodium-cooled reactors are reported along with specific developments on EBWR, BORAK-V, EBR-I, and EBR-H. Thermal and fast reactor safety studies are summarized in terms of fuel-coolant chemical reactions, kinetics of oxidation and ignition of reactor materials, core meltdown studies, and a sodium vapor pressure furnace. Evaluations were made of improved fast reactors for central station power and of a 50-Mwe Prototype Organic Power Reactor (POPR). Developments in instruments, reactcr fuels and materials, reactor components, heat engineering, separations processes, and advanced reactcrs are discussed. (M.C.G.)

1961-06-15T23:59:59.000Z

415

Graphite Reactor | ornl.gov  

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

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

416

On fast reactor kinetics studies  

Science Conference Proceedings (OSTI)

The results and the program of fast reactor core time and space kinetics experiments performed and planned to be performed at the IPPE critical facility is presented. The TIMER code was taken as computation support of the experimental work, which allows transient equations to be solved in 3-D geometry with multi-group diffusion approximation. The number of delayed neutron groups varies from 6 to 8. The code implements the solution of both transient neutron transfer problems: a direct one, where neutron flux density and its derivatives, such as reactor power, etc, are determined at each time step, and an inverse one for the point kinetics equation form, where such a parameter as reactivity is determined with a well-known reactor power time variation function. (authors)

Seleznev, E. F.; Belov, A. A. [Nuclear Safety Inst. of the Russian Academy of Sciences IBRAE (Russian Federation); Matveenko, I. P.; Zhukov, A. M.; Raskach, K. F. [Inst. for Physics and Power Engineering IPPE (Russian Federation)

2012-07-01T23:59:59.000Z

417

Reactor control rod timing system  

DOE Patents (OSTI)

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

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

1982-01-01T23:59:59.000Z

418

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

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

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

419

Calculation of heating values for the high flux isotope reactor  

Science Conference Proceedings (OSTI)

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

Peterson, J.; Ilas, G. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831-6172 (United States)

2012-07-01T23:59:59.000Z

420

Calculation of Heating Values for the High Flux Isotope Reactor  

SciTech Connect

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

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

2012-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor cxs applied" 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

Actinide Burning in CANDU Reactors  

Science Conference Proceedings (OSTI)

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

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

2007-07-01T23:59:59.000Z

422

Water simulation of sodium reactors  

Science Conference Proceedings (OSTI)

The thermal hydraulic simulation of a large sodium reactor by a scaled water model is examined. The Richardson Number, friction coefficient and the Peclet Number can be closely matched with the water system at full power and the similarity is retained for buoyancy driven flows. The simulation of thermal-hydraulic conditions in a reactor vessel provided by a scaled water experiment is better than that by a scaled sodium test. Results from a correctly scaled water test can be tentatively extrapolated to a full size sodium system.

Grewal, S.S.; Gluekler, E.L.

1981-06-28T23:59:59.000Z

423

EERE Postdoctoral Research Awards: How to Apply  

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

How to Apply to someone by E-mail Share EERE Postdoctoral Research Awards: How to Apply on Facebook Tweet about EERE Postdoctoral Research Awards: How to Apply on Twitter Bookmark...

424

THREE ESSAYS ON APPLIED ECONOMICS  

E-Print Network (OSTI)

In this dissertation three essays were presented. In the first two essays we measure the consumer welfare changes caused by U.S. meat price changes. In the third essay the dynamic structure of international gasoline prices using the time series methodology is investigated. In chapter II, we investigate the U.S. consumer behavior on meat consumption depending on a linear expenditure system (LES), and then we simulate the welfare effects of a set of price changes on the U.S. meat consumption. The simulation results show that the amount of consumer welfare change for each meat is not same across the meats under the same percentage change of price. The simulation results also show that when all the prices are doubled the total amount of CV reaches almost the same amount of current total quarterly expenditures for the three meats. In chapter III, we apply the compensating variation (CV) approach for the measurement of consumer welfare losses associated with beef price changes. We applied the long-run cointegrating relationship in vector error correction model (VECM) to estimate the Marshallian demand function. Apparently, the use of long-run cointegration in VECM in deriving the direct Marshallian demand function to measure the consumer welfare change is the first attempt in the literature. This is one of the contributions of the study. The simulation results show that the amount of consumer welfare change for beef is compatible with the one derived from LES methodology. In chapter IV, an empirical framework to summarize the interdependence of four international gasoline markets (New York, U.S. Gulf Coast, Rotterdam and Singapore) is presented. For that purpose, we employ a structural VECM and directed acyclic graphs (DAGs). To solve the identification problem in structural VECM, we apply DAGs derived from contemporaneous VECM innovations. The impulse response functions show that the time period in which a shock in a market affects the other market is very short. Forecast error variance decompositions (FEVD) shows that in all markets, except the U.S. Gulf Coast market, current and past shocks in their own market explained the most of the volatility in their own market in the Short-run.

Shin, Sang-Cheol

2008-08-01T23:59:59.000Z

425

Modular Pebble Bed Reactor High Temperature Gas Reactor  

E-Print Network (OSTI)

For 1150 MW Combined Heat and Power Station Oil Refinery Hydrogen Production Desalinization Plant VHTR;Equipment Layout #12;Modular Pebble Bed Reactor Thermal Power 250 MW Core Height 10.0 m Core Diameter 3.5 m · License by Test · Expert I&C System - Hands free operation #12;MIT MPBR Specifications Thermal Power 250

426

Applied Materials | Open Energy Information  

Open Energy Info (EERE)

Materials Materials Jump to: navigation, search Name Applied Materials Address 3050 Bowers Avenue Place Santa Clara, California Zip 95054 Sector Solar Stock Symbol AMAT Website http://www.appliedmaterials.co Coordinates 37.3775749°, -121.9794416° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":37.3775749,"lon":-121.9794416,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

427

Weatherization and Intergovernmental Program: Apply for Weatherization  

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

Apply Apply for Weatherization Assistance to someone by E-mail Share Weatherization and Intergovernmental Program: Apply for Weatherization Assistance on Facebook Tweet about Weatherization and Intergovernmental Program: Apply for Weatherization Assistance on Twitter Bookmark Weatherization and Intergovernmental Program: Apply for Weatherization Assistance on Google Bookmark Weatherization and Intergovernmental Program: Apply for Weatherization Assistance on Delicious Rank Weatherization and Intergovernmental Program: Apply for Weatherization Assistance on Digg Find More places to share Weatherization and Intergovernmental Program: Apply for Weatherization Assistance on AddThis.com... Plans, Implementation, & Results Weatherization Assistance Program Weatherization Services

428

Vehicle Technologies Office: Applied Battery Research  

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

Applied Battery Research Applied battery research addresses the barriers facing the lithium-ion systems that are closest to meeting the technical energy and power requirements for...

429

Metal-fueled HWR (heavy water reactors) severe accident issues: Differences and similarities to commercial LWRs (light water reactors)  

DOE Green Energy (OSTI)

Differences and similarities in severe accident progression and phenomena between commercial Light Water Reactors (LWR) and metal-fueled isotopic production Heavy Water Reactors (HWR) are described. It is very important to distinguish between accident progression in the two systems because each reactor type behaves in a unique manner to a fuel melting accident. Some of the lessons learned as a result of the extensive commercial severe accident research are not applicable to metal-fueled heavy water reactors. A direct application of severe accident phenomena developed from oxide-fueled LWRs to metal-fueled HWRs may lead to large errors or substantial uncertainties. In general, the application of severe accident LWR concepts to HWRs should be done with the intent to define the relevant issues, define differences, and determine areas of overlap. This paper describes the relevant differences between LWR and metal-fueled HWR severe accident phenomena. Also included in the paper is a description of the phenomena that govern the source term in HWRs, the areas where research is needed to resolve major uncertainties, and areas in which LWR technology can be directly applied with few modifications.

Ellison, P.G.; Hyder, M.L.; Monson, P.R. (Westinghouse Savannah River Co., Aiken, SC (USA)); Coryell, E.W. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

1990-01-01T23:59:59.000Z

430

BNL | Our History: Reactors as Research Tools  

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

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

431

Small- and Medium-Sized Reactors  

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

to essentially the same concept. As defined by the IAEA, a small reactor has an output electrical power of 300 MWe or less while a medium-sized reactor has an electrical power...

432

Reactor Operations informal monthly report September 1994  

SciTech Connect

This paper presents operations at the MRR and HFBR reactors at Brookhaven National Laboratory for September 1994. Reactor run-times, instrumentation, mechanical maintenance, occurrence reports and safety information are listed. Irradiation summaries are included.

Junker, L.

1994-09-01T23:59:59.000Z

433

Reactivity control assembly for nuclear reactor  

DOE Patents (OSTI)

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

Bollinger, Lawrence R. (Schenectady, NY)

1984-01-01T23:59:59.000Z

434

What can Recycling in Thermal Reactors Accomplish?  

SciTech Connect

Thermal recycle provides several potential benefits when used as stop-gap, mixed, or backup recycling to recycling in fast reactors. These three roles involve a mixture of thermal and fast recycling; fast reactors are required to some degree at some time. Stop-gap uses thermal reactors only until fast reactors are adequately deployed and until any thermal-recycle-only facilities have met their economic lifetime. Mixed uses thermal and fast reactors symbiotically for an extended period of time. Backup uses thermal reactors only if problems later develop in the fast reactor portion of a recycling system. Thermal recycle can also provide benefits when used as pure thermal recycling, with no intention to use fast reactors. However, long term, the pure thermal recycling approach is inadequate to meet several objectives.

Steven Piet; Gretchen E. Matthern; Jacob J. Jacobson

2007-09-01T23:59:59.000Z

435

Stability analysis of supercritical water cooled reactors  

E-Print Network (OSTI)

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

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

2005-01-01T23:59:59.000Z

436

Research Reactors Division | Neutron Science | ORNL  

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

is responsible for operation of the High Flux Isotope Reactor. Operating at 85 MW, HFIR is the highest flux reactor-based source of neutrons for research in the United States,...

437

Challenges in the Development of Advanced Reactors  

SciTech Connect

Past generations of nuclear reactors have been successively developed and the next generation is currently being developed, demonstrating the constant progress and technical and industrial vitality of nuclear energy. In 2000 US Department of Energy launched Generation IV International Forum (GIF) which is one of the main international frameworks for the development of future nuclear systems. The six systems that were selected were: sodium cooled fast reactor, lead cooled fast reactor, supercritical water cooled reactor, very high temperature gas cooled reactor (VHTR), gas cooled fast reactor and molten salt reactor. This paper discusses some of the proposed advanced reactor concepts that are currently being researched to varying degrees in the United States, and highlights some of the major challenges these concepts must overcome to establish their feasibility and to satisfy licensing requirements.

P. Sabharwall; M.C. Teague; S.M. Bragg-Sitton; M.W. Patterson

2012-08-01T23:59:59.000Z

438

International Forum for Reactor Aging Management (IFRAM)  

SciTech Connect

The Nuclear Regulatory Commission has undertaken a program to lay the groundwork for defining proactive actions to manage degradation of materials in light water reactors (LWRs). This article discusses the international forum for reactor aging management.

Bond, Leonard J.

2010-11-01T23:59:59.000Z

439

PRESSURIZED WATER REACTOR CORE WITH PLUTONIUM BURNUP  

DOE Patents (OSTI)

A pressurized water reactor is described having a core containing Pu/sup 240/ in which the effective microscopic neutronabsorption cross section of Pu/sup 240/ in unconverted condition decreases as the time of operation of the reactor increases, in order to compensate for loss of reactivity resulting from fission product buildup during reactor operation. This means serves to improve the efficiency of the reactor operation by reducing power losses resulting from control rods and burnable poisons. (AEC)

Puechl, K.H.

1963-09-24T23:59:59.000Z

440

Digital computer operation of a nuclear reactor  

DOE Patents (OSTI)

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

Colley, R.W.

1982-06-29T23:59:59.000Z

Note: This page contains sample records for the topic "reactor cxs applied" 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

Liquid metal cooled nuclear reactor plant system  

DOE Patents (OSTI)

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

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

1993-01-01T23:59:59.000Z

442

Digital computer operation of a nuclear reactor  

DOE Patents (OSTI)

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

Colley, Robert W. (Richland, WA)

1984-01-01T23:59:59.000Z

443

Fast Reactor Curriculum Workshop - Nuclear Engineering Division...  

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

Fast Reactor Curriculum Workshop Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library About Nuclear Energy Nuclear...

444

Material Science Advances Using Test Reactor Facilities  

Science Conference Proceedings (OSTI)

Aug 2, 2010 ... About this Symposium. Meeting, 2011 TMS Annual Meeting & Exhibition. Symposium, Material Science Advances Using Test Reactor Facilities.

445

THE MATERIALS OF FAST BREEDER REACTORS  

E-Print Network (OSTI)

then circulates through a steam generator. The reactor.heat exchanger. steam generator and the primary and

Olander, Donald R.

2013-01-01T23:59:59.000Z

446

Fission energy: The integral fast reactor  

SciTech Connect

The Integral Fast Reactor (IFR) is an innovative reactor concept being developed at Argonne National Laboratory as a such next- generation reactor concept. The IFR concept has a number of specific technical advantages that collectively address the potential difficulties facing the expansion of nuclear power deployment. In particular, the IFR concept can meet all three fundamental requirements needed in a next-generation reactor as discussed below. This document discusses these requirements.

Chang, Yoon I.

1989-01-01T23:59:59.000Z

447

Fast Breeder Reactors for Energy Security  

Science Conference Proceedings (OSTI)

Technical Paper / NSF Workshop on the Research Needs of the Next Generation Nuclear Power Technology / Fission Reactor

William M. Jacobi

448

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

Science Conference Proceedings (OSTI)

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

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

449

Nozzle for electric dispersion reactor  

DOE Patents (OSTI)

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

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

1996-01-01T23:59:59.000Z

450

Nozzle for electric dispersion reactor  

DOE Patents (OSTI)

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

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

1995-01-01T23:59:59.000Z

451

Nozzle for electric dispersion reactor  

DOE Patents (OSTI)

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

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

1998-01-01T23:59:59.000Z

452

Nozzle for electric dispersion reactor  

DOE Patents (OSTI)

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

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

1998-01-01T23:59:59.000Z

453

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

454

SPECIAL TOPIC INTERNATIONAL TOKAMAK REACTOR -  

E-Print Network (OSTI)

(beyond the present generation of large tokamaks) in the world tokamak programme; (2) to assess experiment in the tokamak programme between the present generation of large tokamaks (TFTR, JET, JT-60, T-15 of the development and integration of full-scale components which can be extra- polated to a commercial reactor; (d

Abdou, Mohamed

455

High temperature catalytic membrane reactors  

DOE Green Energy (OSTI)

Current state-of-the-art inorganic oxide membranes offer the potential of being modified to yield catalytic properties. The resulting modules may be configured to simultaneously induce catalytic reactions with product concentration and separation in a single processing step. Processes utilizing such catalytically active membrane reactors have the potential for dramatically increasing yield reactions which are currently limited by either thermodynamic equilibria, product inhibition, or kinetic selectivity. Examples of commercial interest include hydrogenation, dehydrogenation, partial and selective oxidation, hydrations, hydrocarbon cracking, olefin metathesis, hydroformylation, and olefin polymerization. A large portion of the most significant reactions fall into the category of high temperature, gas phase chemical and petrochemical processes. Microporous oxide membranes are well suited for these applications. A program is proposed to investigate selected model reactions of commercial interest (i.e. dehydrogenation of ethylbenzene to styrene and dehydrogenation of butane to butadiene) using a high temperature catalytic membrane reactor. Membranes will be developed, reaction dynamics characterized, and production processes developed, culminating in laboratory-scale demonstration of technical and economic feasibility. As a result, the anticipated increased yield per reactor pass economic incentives are envisioned. First, a large decrease in the temperature required to obtain high yield should be possible because of the reduced driving force requirement. Significantly higher conversion per pass implies a reduced recycle ratio, as well as reduced reactor size. Both factors result in reduced capital costs, as well as savings in cost of reactants and energy.

Not Available

1990-03-01T23:59:59.000Z

456

Overview of Reactor and Nuclear  

E-Print Network (OSTI)

and Safety Gary Mays Nuclear Data and Criticality Safety Mike Dunn Nuclear Security Modeling Tim Valentine - Office of Environmental Management - Office of Intelligence · National Nuclear Security AdministrationOverview of Reactor and Nuclear Systems Division Cecil Parks RNS Division Director parkscv

457

REACTOR CONTROL ROD OPERATING SYSTEM  

DOE Patents (OSTI)

A nuclear reactor control rod mechanism is designed which mechanically moves the control rods into and out of the core under normal conditions but rapidly forces the control rods into the core by catapultic action in the event of an emergency. (AEC)

Miller, G.

1961-12-12T23:59:59.000Z

458

Pebble Flow Experiments For Pebble Bed Reactors  

E-Print Network (OSTI)

Pebble Flow Experiments For Pebble Bed Reactors Andrew C. Kadak1 Department of Nuclear Engineering of Technology 2nd International Topical Meeting on High Temperature Reactor Technology Institute of Nuclear in such a reactor would conform to granular flow theory which suggested rapid mixing as opposed to linear flow lines

Bazant, Martin Z.

459

Synthesis and Manipulation of Semiconductor Nanocrystals in Microfluidic Reactors  

E-Print Network (OSTI)

macroscopic flow reactors, Taylor theory demonstrates thattheory and advantages of microfluidic devices and their application as microscale chemical reactors.

Chan, Emory Ming-Yue

2006-01-01T23:59:59.000Z

460

DARPA Learning Applied to Ground Robots (LAGR)  

Science Conference Proceedings (OSTI)

DARPA Learning Applied to Ground Robots (LAGR) Project (Concluded). Summary: The National Institute of Standards ...

2012-01-04T23:59:59.000Z

Note: This page contains sample records for the topic "reactor cxs applied" 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

Domestic U.S. Reactor Conversions: Fact Sheet | National Nuclear Security  

National Nuclear Security Administration (NNSA)

Domestic U.S. Reactor Conversions: Fact Sheet | National Nuclear Security Domestic U.S. Reactor Conversions: Fact Sheet | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Fact Sheets > Domestic U.S. Reactor Conversions: Fact Sheet Fact Sheet Domestic U.S. Reactor Conversions: Fact Sheet Mar 23, 2012 The National Nuclear Security Administration (NNSA) helps convert research

462

Control of reactor coolant flow path during reactor decay heat removal  

DOE Patents (OSTI)

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

Hunsbedt, Anstein N. (Los Gatos, CA)

1988-01-01T23:59:59.000Z

463

Valorisation of forestry waste by pyrolysis in an auger reactor  

SciTech Connect

Pyrolysis of forestry waste has been carried out in an auger reactor to study the influence of operational variables on the reactor performance and the properties of the related products. Pine woodchips were used for the first time as raw material and fed continuously into the reactor. Ten experiments were carried out under inert atmosphere at: (i) different reaction temperature (1073, 973, 873, 823 and 773 K); (ii) different solid residence time (5, 3, 2 and 1.5 min); and (iii) different biomass flow rate (3.9, 4.8 and 6.9 kg/h). Results show that the greatest yields for liquid production (59%) and optimum product characterisation were obtained at the lowest temperature studied (773 K) and applying solid residence times longer than 2 min. Regarding bio-oil properties, GC/MS qualitative identification show that the most abundant compounds are volatile polar compounds, phenols and benzenediols; and very few differences can be observed among the samples regardless of the pyrolysis operating conditions. On the whole, experimental results demonstrate that complete reaction of forest woodchips can be achieved in an auger reactor in most of the experimental conditions tested. Moreover, this study presents the initial steps for the future scaling up of the auger reactor with the aim of converting it into a mobile plant which will be able to remotely process biomass such as energy crops, forestry and agricultural wastes to obtain bio-oil that, in turn, can be used as energy vector to avoid high transport costs.

Puy, Neus, E-mail: neus.puy@uab.cat [Institute of Environmental Science and Technology (ICTA), Universitat Autonoma de Barcelona, Edifici Cn - Campus de la UAB, 08193 Cerdanyola del Valles, Barcelona (Spain); Murillo, Ramon; Navarro, Maria V.; Lopez, Jose M. [Instituto de Carboquimica, CSIC, M Luesma Castan 4, 50018 Zaragoza (Spain); Rieradevall, Joan [Institute of Environmental Science and Technology (ICTA), Universitat Autonoma de Barcelona, Edifici Cn - Campus de la UAB, 08193 Cerdanyola del Valles, Barcelona (Spain); Department of Chemical Engineering, Universitat Autonoma de Barcelona, ETSE, Campus de la UAB, 08193 Cerdanyola del Valles, Barcelona (Spain); Fowler, G. [Department of Civil and Environmental Engineering, Imperial College London, South Kensington Campus, London SW7 2AZ (United Kingdom); Aranguren, Ignacio; Garcia, Tomas [Instituto de Carboquimica, CSIC, M Luesma Castan 4, 50018 Zaragoza (Spain); Bartroli, Jordi [Department of Chemistry, Universitat Autonoma de Barcelona, Edifici Cn - Campus de la UAB, 08193 Cerdanyola del Valles, Barcelona (Spain); Mastral, Ana M. [Instituto de Carboquimica, CSIC, M Luesma Castan 4, 50018 Zaragoza (Spain)

2011-06-15T23:59:59.000Z

464

Reactor and Nuclear Systems Division (RNSD)  

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

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

465

Self isolating high frequency saturable reactor  

DOE Patents (OSTI)

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

Moore, James A. (Powell, TN)

1998-06-23T23:59:59.000Z

466

Control rod drive for reactor shutdown  

DOE Patents (OSTI)

A means for rapidly shutting down or scramming a nuclear reactor, such as a liquid metal-cooled fast breeder reactor, and serves as a backup to the primary shutdown system. The control rod drive consists basically of an in-core assembly, a drive shaft and seal assembly, and a control drive mechanism. The control rod is driven into the core region of the reactor by gravity and hydraulic pressure forces supplied by the reactor coolant, thus assuring that common mode failures will not interfere with or prohibit scramming the reactor when necessary.

McKeehan, Ernest R. (Los Gatos, CA); Shawver, Bruce M. (San Jose, CA); Schiro, Donald J. (San Jose, CA); Taft, William E. (Los Gatos, CA)

1976-01-20T23:59:59.000Z

467

Fast-acting nuclear reactor control device  

DOE Patents (OSTI)

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

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

1993-01-01T23:59:59.000Z

468

Shutdown system for a nuclear reactor  

DOE Patents (OSTI)

An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion. 8 figs.

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

1984-06-05T23:59:59.000Z

469

Shutdown system for a nuclear reactor  

DOE Patents (OSTI)

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

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

1984-01-01T23:59:59.000Z

470

Shutdown system for a nuclear reactor  

DOE Patents (OSTI)

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

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

1982-01-20T23:59:59.000Z

471

Reactor monitoring and safeguards using antineutrino detectors  

E-Print Network (OSTI)

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

N. S. Bowden

2008-09-12T23:59:59.000Z

472

Reactor monitoring and safeguards using antineutrino detectors  

E-Print Network (OSTI)

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

Bowden, N S

2008-01-01T23:59:59.000Z

473

Categorical Exclusion Determinations: Bonneville Power Administration |  

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

June 4, 2013 June 4, 2013 CX-010433: Categorical Exclusion Determination Memaloose Meadows Land Acquisition CX(s) Applied: B1.25 Date: 06/04/2013 Location(s): Oregon Offices(s): Bonneville Power Administration June 3, 2013 CX-010436: Categorical Exclusion Determination Tri-Cities Maintenance Headquarters Project CX(s) Applied: B1.15 Date: 06/03/2013 Location(s): Washington, Washington Offices(s): Bonneville Power Administration June 3, 2013 CX-010435: Categorical Exclusion Determination De Moss Substation Expansion CX(s) Applied: B4.6 Date: 06/03/2013 Location(s): Oregon Offices(s): Bonneville Power Administration June 3, 2013 CX-010434: Categorical Exclusion Determination LaPine Substation Shunt Reactor Addition CX(s) Applied: B4.6 Date: 06/03/2013 Location(s): Oregon, Oregon Offices(s): Bonneville Power Administration

474

Categorical Exclusion Determinations: Wyoming | Department of Energy  

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

May 26, 2011 May 26, 2011 CX-006716: Categorical Exclusion Determination New B-1-3 Pit and Box Construction CX(s) Applied: B1.3, B6.1 Date: 05/26/2011 Location(s): Casper, Wyoming Office(s): RMOTC May 17, 2011 CX-006719: Categorical Exclusion Determination Casing Drilling Test CX(s) Applied: B1.3, B3.7, B5.12 Date: 05/17/2011 Location(s): Casper, Wyoming Office(s): RMOTC May 5, 2011 CX-005852: Categorical Exclusion Determination Stegall-Wayside 230 Kilovolt Access Road Extension CX(s) Applied: B1.13 Date: 05/05/2011 Location(s): Dawes County, Wyoming Office(s): Western Area Power Administration-Rocky Mountain Region April 29, 2011 CX-005664: Categorical Exclusion Determination Development and Testing of Compact Heat Exchange Reactors (CHER) for Synthesis of Liquid Fuels CX(s) Applied: B3.6

475

Categorical Exclusion Determinations: Idaho | Department of Energy  

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

June 12, 2012 June 12, 2012 CX-008696: Categorical Exclusion Determination Power Circuit Breaker Replacement Project CX(s) Applied: B4.6 Date: 06/12/2012 Location(s): Washington, Oregon, Idaho, Montana, Wyoming Offices(s): Bonneville Power Administration June 7, 2012 CX-008731: Categorical Exclusion Determination M-10 Emergency Pump Backup Power CX(s) Applied: B2.5 Date: 06/07/2012 Location(s): Idaho Offices(s): Idaho Operations Office June 7, 2012 CX-008730: Categorical Exclusion Determination Materials and Fuels Complex Underground and Aboveground Storage Tank Replacement CX(s) Applied: B2.5 Date: 06/07/2012 Location(s): Idaho Offices(s): Idaho Operations Office May 28, 2012 CX-008734: Categorical Exclusion Determination Infrastructure for Advanced Reactor Materials Center CX(s) Applied: B2.2

476

Categorical Exclusion Determinations: B2.2 | Department of Energy  

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

May 17, 2012 May 17, 2012 CX-008753: Categorical Exclusion Determination Nuclear Instrumentation Upgrade for University of Missouri Research Reactor Power Level Monitoring - University of Missouri CX(s) Applied: B2.2 Date: 05/17/2012 Location(s): Idaho Offices(s): Idaho Operations Office May 17, 2012 CX-008749: Categorical Exclusion Determination Reactor Power Up Rate, Compressor Replacement, Neutron Radiography Restore, Liquid Scintillation Counter - Texas Agricultural & Mechanical University CX(s) Applied: B2.2, B3.6 Date: 05/17/2012 Location(s): Idaho Offices(s): Idaho Operations Office May 17, 2012 CX-008756: Categorical Exclusion Determination Equipment Upgrade for the University of New Mexico AGN-201M Reactor - University of New Mexico CX(s) Applied: B2.2, B3.6 Date: 05/17/2012

477

Safety control circuit for a neutronic reactor  

DOE Patents (OSTI)

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

Ellsworth, Howard C. (Richland, WA)

2004-04-27T23:59:59.000Z

478

Decommssioning Of Research Reactor: Problemsand Experience  

E-Print Network (OSTI)

The study of the preparation for decommissioning the Research Reactor in Salaspils (Latvia) and the experience of decommissioning the Research Reactor in Sosny (Belarus) show that the problem of decommissioning research reactors is acute for countries that have no NPPs or their own nuclear industry. It also is associated with regulatory framework, planning and design, dismantling technologies, decontamination of radioactive equipment and materials, spent fuel and radioactive waste management, etc. 1. INTRODUCTION According to the IAEA research reactor database, there are about 300 research reactors worldwide [1]. At present over 30% of them have lifetimes of more than 35 years, 60% of more than 25 years. After the Chernobyl accident, significant efforts were made by many countries to modernize old research reactors aiming, first of all, at ensuring safe operation. However, a large number of aging research reactors will be facing shutdown in the near future. The problem of decommis...

Alexander Mikhalevich

2000-01-01T23:59:59.000Z

479

Nuclear reactor vessel fuel thermal insulating barrier  

DOE Patents (OSTI)

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

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

2013-03-19T23:59:59.000Z

480

History of Research Reactors at Brookhaven  

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

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

Note: This page contains sample records for the topic "reactor cxs applied" 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

Performance Evaluation of Advanced LLW Liquid Processing Technology: Boiling Water Reactor Liquid Processing  

Science Conference Proceedings (OSTI)

This report provides condensed information on boiling water reactor (BWR) membrane based liquid radwaste processing systems. The report presents specific details of the technology, including design, configuration, and performance. This information provides nuclear plant personnel with data useful in evaluating the merits of applying advanced processes at their plant.

2001-11-26T23:59:59.000Z

482

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

483

Application of genetic algorithm to optimize burnable poison placement in pressurized water reactors  

Science Conference Proceedings (OSTI)

An efficient and a practical genetic algorithm tool was developed and applied successfully to Burnable Poisons (BPs) placement optimization problem in the reference Three Mile Island-1 (TMI-1) core. Core BP optimization problem means developing a BP ... Keywords: burnable poison, decision variables, gadolinium, genetic algorithm, nuclear, optimization, reactor

Serkan Yilmaz; Kostadin Ivanov; Samuel Levine

2005-06-01T23:59:59.000Z

484

Licensing topical report: applicability of Division 1 regulatory guides to high-temperature gas-cooled reactors  

SciTech Connect

The application of Division 1 (power reactors) regulatory guides to high-temperature gas-cooled reactors (HTGRs) is discussed. About eighty of the Division 1 guides can be applied to any type of reactor; the remaining sixty, mostly written for light water reactors (LWRs), are divided between (1) those not applicable to the HTGR because of fundamental differences in design, (2) those applicable in intent but containing positions specific to LWRs, and (3) those written for LWRs but of sufficient generality to be applied to the HTGR without major exception. Emphasis is placed on issues which involve the unique characteristics of the HTGR. The regulatory guides evaluated are those extant as of early 1980. The positions presented are subject to periodic updating owing to the continuing modification of the guides and because the design options for the HTGR are at various stages of development. Nevertheless, this report is believed to provide a sound basis for evaluating conformance with existing Division 1 guides.

Lewis, J.H.

1980-12-01T23:59:59.000Z

485

POWER GENERATING NEUTRONIC REACTOR SYSTEM  

DOE Patents (OSTI)

This patent relates to reactor systems of the type wherein the cooiing medium is a liquid which is converted by the heat of the reaction to steam which is conveyed directly to a pnime mover such as a steam turbine driving a generatore after which it is condensed and returred to the coolant circuit. In this design, the reactor core is disposed within a tank for containing either a slurry type fuel or an aggregation of solid fuel elements such as elongated rods submerged in a liquid moderator such as heavy water. The top of the tank is provided with a nozzle which extends into an expansion chamber connected with the upper end of the tank, the coolant being maintained in the expansion chamber at a level above the nozzle and the steam being formed in the expansion chamber.

Vernon, H.C.

1958-03-01T23:59:59.000Z

486

HEAVY WATER MODERATED NEUTRONIC REACTOR  

DOE Patents (OSTI)

A nuclear reactor of the type which utilizes uranium fuel elements and a liquid coolant is described. The fuel elements are in the form of elongated tubes and are disposed within outer tubes extending through a tank containing heavy water, which acts as a moderator. The ends of the fuel tubes are connected by inlet and discharge headers, and liquid bismuth is circulated between the headers and through the fuel tubes for cooling. Helium is circulated through the annular space between the outer tubes in the tank and the fuel tubes to cool the water moderator to prevent boiling. The fuel tubes are covered with a steel lining, and suitable control means, heat exchange means, and pumping means for the coolants are provided to complete the reactor assembly.

Szilard, L.

1958-04-29T23:59:59.000Z

487

Reactor Internals Segmentation Experience Report  

Science Conference Proceedings (OSTI)

Several U.S. nuclear power plants entered decommissioning in the 1990's. Based on current information, the next group of plants will not begin this process for nearly a decade. This report provides detailed information on projects involving the segmentation of the Reactor Vessel Internal Subassemblies conducted at a number of plants undergoing decommissioning. Utilities have found this task very challenging from both a technological and radiation exposure perspective. This report documents the experience...

2007-07-10T23:59:59.000Z

488

Reactor safeguards against insider sabotage  

SciTech Connect

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

Bennett, H.A.

1982-03-01T23:59:59.000Z

489

FUEL ELEMENT FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

A fuel element particularly adapted for use in nuclear reactors of high power density is offered. It has fissionable fuel pellet segments mounted in a tubular housing and defining a central passage in the fuel element. A burnable poison element extends through the central passage, which is designed to contain more poison material at the median portion than at the end portions thereby providing a more uniform hurnup and longer reactivity life.

Bassett, C.H.

1961-05-16T23:59:59.000Z

490

Flow duct for nuclear reactors  

DOE Patents (OSTI)

Improved liquid sodium flow ducts for nuclear reactors are described wherein the improvement comprises varying the wall thickness of each of the walls of a polygonal tubular duct structure so that each of the walls is of reduced cross-section along the longitudinal center line and of a greater cross-section along wall junctions with the other walls to form the polygonal tubular configuration.

Straalsund, Jerry L. (Richland, WA)

1978-01-01T23:59:59.000Z

491

NUCLEAR REACTOR COMPENENT CLADDING MATERIAL  

DOE Patents (OSTI)

Fuel elements and coolant tubes used in nuclear reactors of the heterogeneous, water-cooled type are described, wherein the coolant tubes extend through the moderator and are adapted to contain the fuel elements. The invention comprises forming the coolant tubes and the fuel element cladding material from an alloy of aluminum and nickel, or an alloy of aluminum, nickel, alloys are selected to prevent intergranular corrosion of these components by water at temperatures up to 35O deg C.

Draley, J.E.; Ruther, W.E.

1959-01-27T23:59:59.000Z

492

The ARIES tokamak reactor study  

SciTech Connect

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

Not Available

1989-10-01T23:59:59.000Z

493

Existing reactor expansion study basis  

SciTech Connect

The latest HAPO Five Year Program review, HW-59633, forecasts substantial increases in Pu production from the eight existing Hanford reactors over the next several years. These production increases would be attained by a combination of several methods which include increased reactor power levels resulting from higher process water flow rates and coolant bulk outlet temperatures, improved time operated efficiency, higher conversion ratios, and reduced transient reactivity losses. In order to provide a realistic basis for budgeting to meet these or other increased production goals, it is necessary that a study program be undertaken to determine in general terms the plant changes required to support these forecasted levels, to evaluate the economic and technical feasibility of achieving the process conditions, and to present an integrated program for achieving these objectives. This study program will necessarily consider the interrelated effects of a number of various facets of reactor and water plant process conditions, operational requirements, and proposed development programs. The purpose of this document is to present a plan for the execution of the proposed study. Included in this outline are a review of the basic study considerations, problem assignments and schedules, and manpower and cost estimates for the performance of the study.

Heacock, H.W.

1959-06-24T23:59:59.000Z

494

Russian RBMK reactor design information  

Science Conference Proceedings (OSTI)

This document concerns the systems, design, and operations of the graphite-moderated, boiling, water-cooled, channel-type (RBMK) reactors located in the former Soviet Union (FSU). The Russian Academy of Sciences Nuclear Safety Institute (NSI) in Moscow, Russia, researched specific technical questions that were formulated by the Pacific Northwest Laboratory (PNL) and provided detailed technical answers to those questions. The Russian response was prepared in English by NSI in a question-and-answer format. This report presents the results of that technical exchange in the context they were received from the NSI organization. Pacific Northwest Laboratory is generating this document to support the US Department of Energy (DOE) community in responding to requests from FSU states, which are seeking Western technological and financial assistance to improve the safety systems of the Russian-designed reactors. This report expands upon information that was previously available to the United States through bilateral information exchanges, international nuclear society meetings, International Atomic Energy Agency (IAEA) reactor safety programs, and Research and Development Institute of Power Engineering (RDIPE) reports. The response to the PNL questions have not been edited or reviewed for technical consistency or accuracy by PNL staff or other US organizations, but are provided for use by the DOE community in the form they were received.

Not Available

1993-11-01T23:59:59.000Z

495

Prospects for Tokamak Fusion Reactors  

SciTech Connect

This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant.

Sheffield, J.; Galambos, J.

1995-04-01T23:59:59.000Z

496

Advanced Safeguards Approaches for New Fast Reactors  

SciTech Connect

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

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

2007-12-15T23:59:59.000Z

497

Oxidation Kinetics Modeling Applying Phase Field Approach  

Science Conference Proceedings (OSTI)

Presentation Title, Oxidation Kinetics Modeling Applying Phase Field Approach ... chemical reaction rates will increase exponentially and environmental attack ...

498

Researcher, Los Alamos National Laboratory - Applied Physics...  

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Applied Physics Division | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response...