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1

The High Flux Beam Reactor at Brookhaven National Laboratory  

SciTech Connect (OSTI)

Brookhaven National Laboratory`s High Flux Beam Reactor (HFBR) was built because of the need of the scientist to always want `more`. In the mid-50`s the Brookhaven Graphite reactor was churning away producing a number of new results when the current generation of scientists, led by Donald Hughes, realized the need for a high flux reactor and started down the political, scientific and engineering path that led to the BFBR. The effort was joined by a number of engineers and scientists among them, Chemick, Hastings, Kouts, and Hendrie, who came up with the novel design of the HFBR. The two innovative features that have been incorporated in nearly all other research reactors built since are: (i) an under moderated core arrangement which enables the thermal flux to peak outside the core region where beam tubes can be placed, and (ii) beam tubes that are tangential to the core which decrease the fast neutron background without affecting the thermal beam intensity. Construction began in the fall of 1961 and four years later, at a cost of $12 Million, criticality was achieved on Halloween Night, 1965. Thus began 30 years of scientific accomplishments.

Shapiro, S.M.

1994-12-31T23:59:59.000Z

2

Decommissioning of the high flux beam reactor at Brookhaven Lab  

SciTech Connect (OSTI)

The high-flux beam reactor (HFBR) at the Brookhaven National Laboratory was a heavy water cooled and moderated reactor that achieved criticality on Oct. 31, 1965. It operated at a power level of 40 megawatts. An equipment upgrade in 1982 allowed operations at 60 megawatts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 megawatts. The HFBR was shut down in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of groundwater from wells located adjacent to the reactor's spent fuel pool. The reactor remained shut down for almost three years for safety and environmental reviews. In November 1999 the United States Dept. of Energy decided to permanently shut down the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome, which still contains the irradiated reactor vessel, is presently under 24/7 surveillance for safety. Detailed dosimetry performed for the HFBR decommissioning during 1996-2009 is described in the paper. (authors)

Hu, J.P. [National Synchrotron Light Source, Brookhaven Laboratory, Upton, NY 11973 (United States); Reciniello, R.N. [Radiological Control Div., Brookhaven Laboratory, Upton, NY 11973 (United States); Holden, N.E. [National Nuclear Data Center, Brookhaven Laboratory, Upton, NY 11973 (United States)

2011-07-01T23:59:59.000Z

3

EIS-0291: High Flux Beam Reactor (HFBR) Transition Project at the Brookhaven National Laboratory, Upton, New York  

Broader source: Energy.gov [DOE]

The EIS evaluates the range of reasonable alternatives and their impacts regarding the future management of the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL).

4

INDEPENDENT VERIFICATION SURVEY OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT OUTSIDE AREAS BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK  

SciTech Connect (OSTI)

5098-SR-03-0 FINAL REPORT- INDEPENDENT VERIFICATION SURVEY OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT OUTSIDE AREAS, BROOKHAVEN NATIONAL LABORATORY

P.C. Weaver

2010-12-15T23:59:59.000Z

5

LETTER REPORT INDEPENDENT VERIFICATION OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT FAN HOUSE, BUILDING 704 BNL  

SciTech Connect (OSTI)

5098-LR-01-0 -LETTER REPORT INDEPENDENT VERIFICATION OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT FAN HOUSE, BUILDING 704 BROOKHAVEN NATIONAL LABORATORY

P.C. Weaver

2010-10-22T23:59:59.000Z

6

Three-dimensional calculations of neutron streaming in the beam tubes of the ORNL HFIR (High Flux Isotope Reactor) Reactor  

SciTech Connect (OSTI)

The streaming of neutrons through the beam tubes in High Flux Isotope Reactor at Oak Ridge National Laboratory has resulted in a reduction of the fracture toughness of the reactor vessel. As a result, an evaluation of vessel integrity was undertaken in order to determine if the reactor can be operated again. As a part of this evaluation, three-dimensional neutron transport calculations were performed to obtain fluxes at points of interest in the wall of the vessel. By comparing the calculated and measured activation of dosimetry specimens from the vessel surveillance program, it was determined that the calculated flux shape was satisfactory to transpose the surveillance data to the locations in the vessel. A bias factor was applied to correct for the average C/E ratio of 0.69. 8 refs., 7 figs., 3 tabs.

Childs, R.L.; Rhoades, W.A.; Williams, L.R.

1988-01-01T23:59:59.000Z

7

TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 2 DF WASTE LINE REMOVAL, BNL  

SciTech Connect (OSTI)

5098-SR-02-0 PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 2 DF WASTE LINE REMOVAL, BROOKHAVEN NATIONAL LABORATORY

P.C. Weaver

2010-07-09T23:59:59.000Z

8

Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression  

DOE Patents [OSTI]

The invention is a laser or particle-beam-driven fusion reactor system which takes maximum advantage of both the very short pulsed nature of the energy release of inertial confinement fusion (ICF) and the very small volumes within which the thermonuclear burn takes place. The pulsed nature of ICF permits dynamic direct energy conversion schemes such as magnetohydrodynamic (MHD) generation and magnetic flux compression; the small volumes permit very compact blanket geometries. By fully exploiting these characteristics of ICF, it is possible to design a fusion reactor with exceptionally high power density, high net electric efficiency, and low neutron-induced radioactivity. The invention includes a compact blanket design and method and apparatus for obtaining energy utilizing the compact blanket.

Lasche, G.P.

1983-09-29T23:59:59.000Z

9

PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 1, BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK  

SciTech Connect (OSTI)

5098-SR-05-0 PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 1 BROOKHAVEN NATIONAL LABORATORY

E.M. Harpenau

2010-12-15T23:59:59.000Z

10

PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 5, BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK  

SciTech Connect (OSTI)

5098-SR-04-0 PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 5, BROOKHAVEN NATIONAL LABORATORY

P.C. Weaver

2010-11-03T23:59:59.000Z

11

Tensile and impact testing of an HFBR (High Flux Beam Reactor) control rod follower  

SciTech Connect (OSTI)

The Materials Technology Group of the Department of Nuclear Energy (DNE) at Brookhaven National Laboratory (BNL) undertook a program to machine and test specimens from a control rod follower from the High Flux Beam Reactor (HFBR). Tensile and Charpy impact specimens were machined and tested from non-irradiated aluminum alloys in addition to irradiated 6061-T6 from the HFBR. The tensile test results on irradiated material showed a two-fold increase in tensile strength to a maximum of 100.6 ksi. The impact resistance of the irradiated material showed a six-fold decrease in values (3 in-lb average) compared to similar non-irradiated material. Fracture toughness (K{sub I}) specimens were tested on an unirradiated compositionally and dimensionally similar (to HFBR follower) 6061 T-6 material with K{sub max} values of 24.8 {plus minus} 1.0 Ksi{radical}in (average) being obtained. The report concludes that the specimens produced during the program yielded reproducible and believable results and that proper quality assurance was provided throughout the program. 9 figs., 6 tabs.

Czajkowski, C.J.; Schuster, M.H.; Roberts, T.C.; Milian, L.W.

1989-08-01T23:59:59.000Z

12

Determining Reactor Neutrino Flux  

E-Print Network [OSTI]

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

Jun Cao

2012-03-08T23:59:59.000Z

13

Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression  

DOE Patents [OSTI]

A high-power-density laser or charged-particle-beam fusion reactor system maximizes the directed kinetic energy imparted to a large mass of liquid lithium by a centrally located fusion target. A fusion target is embedded in a large mass of lithium, of sufficient radius to act as a tritium breeding blanket, and provided with ports for the access of beam energy to implode the target. The directed kinetic energy is converted directly to electricity with high efficiency by work done against a pulsed magnetic field applied exterior to the lithium. Because the system maximizes the blanket thickness per unit volume of lithium, neutron-induced radioactivities in the reaction chamber wall are several orders of magnitude less than is typical of other fusion reactor systems.

Lasche, George P. (Arlington, VA)

1988-01-01T23:59:59.000Z

14

Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression  

DOE Patents [OSTI]

A high-power-density-laser or charged-particle-beam fusion reactor system maximizes the directed kinetic energy imparted to a large mass of liquid lithium by a centrally located fusion target. A fusion target is embedded in a large mass of lithium, of sufficient radius to act as a tritium breeding blanket, and provided with ports for the access of beam energy to implode the target. The directed kinetic energy is converted directly to electricity with high efficiency by work done against a pulsed magnetic field applied exterior to the lithium. Because the system maximizes the blanket thickness per unit volume of lithium, neutron-induced radioactivities in the reaction chamber wall are several orders of magnitude less than is typical of other fusion reactor systems. 25 figs.

Lasche, G.P.

1987-02-20T23:59:59.000Z

15

Neutron Scattering Science User Office, neutronusers@ornl.gov or (865) 574-4600. Proposals for beam time at Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR)  

E-Print Network [OSTI]

Neutron Scattering Science User Office, neutronusers@ornl.gov or (865) 574-4600. Proposals for beam Wildgruber, wildgrubercu@ornl.gov. VISION CallforProposals neutrons.ornl.gov Neutron Scattering Science - Oak time at Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR) and Spallation Neutron Source

Pennycook, Steve

16

Final Report Independent Verification Survey of the High Flux Beam Reactor, Building 802 Fan House Brookhaven National Laboratory Upton, New York  

SciTech Connect (OSTI)

On May 9, 2011, ORISE conducted verification survey activities including scans, sampling, and the collection of smears of the remaining soils and off-gas pipe associated with the 802 Fan House within the HFBR (High Flux Beam Reactor) Complex at BNL. ORISE is of the opinion, based on independent scan and sample results obtained during verification activities at the HFBR 802 Fan House, that the FSS (final status survey) unit meets the applicable site cleanup objectives established for as left radiological conditions.

Harpeneau, Evan M. [Oak Ridge Institute for Science and Education, Oak Ridge, TN (United States). Independent Environmental Assessment and Verification Program

2011-06-24T23:59:59.000Z

17

SUMMARY AND RESULTS LETTER REPORT – INDEPENDENT VERIFICATION OF THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PROJECT, PHASE 3: TRENCHES 2, 3, AND 4 BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK  

SciTech Connect (OSTI)

5098-LR-02-0 SUMMARY AND RESULTS LETTER REPORT – INDEPENDENT VERIFICATION OF THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PROJECT, PHASE 3 TRENCHES 2, 3, AND 4 BROOKHAVEN NATIONAL LABORATORY

E.M. Harpenau

2010-11-15T23:59:59.000Z

18

TYPE A VERIFICATION REPORT FOR THE HIGH FLUX BEAM REACTOR STACK AND GROUNDS, BROOKHAVEN NATIONAL LABORATORY, UPTON, NEW YORK DCN 5098-SR-08-0  

SciTech Connect (OSTI)

The U.S. Department of Energy (DOE) Order 458.1 requires independent verification (IV) of DOE cleanup projects (DOE 2011). The Oak Ridge Institute for Science and Education (ORISE) has been designated as the responsible organization for IV of the High Flux Beam Reactor (HFBR) Stack and Grounds area at Brookhaven National Laboratory (BNL) in Upton, New York. The IV evaluation may consist of an in-process inspection with document and data reviews (Type A Verification) or a confirmatory survey of the site (Type B Verification). DOE and ORISE determined that a Type A verification of the documents and data for the HFBR Stack and Grounds: Survey Units (SU) 6, 7, and 8 was appropriate based on the initial survey unit classification, the walkover surveys, and the final analytical results provided by the Brookhaven Science Associates (BSA).

Evan Harpenau

2011-11-30T23:59:59.000Z

19

Department of Energy's High Flux Beam Reactor (HFBR), September 15--19, 1980: An independent on-site safety review  

SciTech Connect (OSTI)

The intent of this on-site safety review was to make a broad management assessment of HFBR operations, rather than conduct a detailed in-depth audit. The result of the review should only be considered as having identified trends or indications. The Team's observations and recommendations for the most part are based upon licensed reactor facility practices used to meet industry standards. These standards form the basis for many of the comments in this report. The Team believes that a uniform minimum standard of performance should be achieved in the operation of DOE reactors. In order to assure that this is accomplished, clear standards are necessary. Consistent with the past AEC and ERDA policy, the team has used the standards of the commercial nuclear power industry. It is recognized that this approach is conservative in that the HFBR reactor has a significantly greater degree of inherent safety (low pressure, temperature, power, etc.) than a licensed reactor.

Not Available

1981-02-01T23:59:59.000Z

20

Beam characterization at the Neutron Radiography Reactor  

SciTech Connect (OSTI)

The quality of a neutron-imaging beam directly impacts the quality of radiographic images produced using that beam. Fully characterizing a neutron beam, including determination of the beam's effective length-to-diameter ratio, neutron flux profile, energy spectrum, potential image quality, and beam divergence, is vital for producing quality radiographic images. This paper provides a characterization of the east neutron imaging beamline at the Idaho National Laboratory Neutron Radiography Reactor (NRAD). The experiments which measured the beam's effective length-to-diameter ratio and potential image quality are based on American Society for Testing and Materials (ASTM) standards. An analysis of the image produced by a calibrated phantom measured the beam divergence. The energy spectrum measurements consist of a series of foil irradiations using a selection of activation foils, compared to the results produced by a Monte Carlo n-Particle (MCNP) model of the beamline. The NRAD has an effective collimation ratio greater than 125, a beam divergence of 0.3 +_ 0.1 degrees, and a gold foil cadmium ratio of 2.7. The flux profile has been quantified and the facility is an ASTM Category 1 radiographic facility. Based on bare and cadmium covered foil activation results, the neutron energy spectrum used in the current MCNP model of the radiography beamline over-samples the thermal region of the neutron energy spectrum.

Sarah W. Morgan; Jeffrey C. King; Chad L. Pope

2013-12-01T23:59:59.000Z

Note: This page contains sample records for the topic "flux beam reactor" 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

High Flux Isotope Reactor power upgrade status  

SciTech Connect (OSTI)

A return to 100-MW operation is being planned for the High Flux Isotope Reactor (HFIR). Recent improvements in fuel element manufacturing procedures and inspection equipment will be exploited to reduce hot spot and hot streak factors sufficiently to permit the power upgrade without an increase in primary coolant pressure. Fresh fuel elements already fabricated for future use are being evaluated individually for power upgrade potential based on their measured coolant channel dimensions.

Rothrock, R.B.; Hale, R.E. [Oak Ridge National Lab., TN (United States); Cheverton, R.D. [Delta-21 Resources Inc., Oak Ridge, TN (United States)

1997-03-01T23:59:59.000Z

22

High Flux Isotope Reactor cold neutron source reference design concept  

SciTech Connect (OSTI)

In February 1995, Oak Ridge National Laboratory`s (ORNL`s) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH{sub 2}) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH{sub 2} cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept.

Selby, D.L.; Lucas, A.T.; Hyman, C.R. [and others

1998-05-01T23:59:59.000Z

23

Three-dimensional discrete ordinates radiation transport calculations of neutron fluxes for beginning-of-cycle at several pressure vessel surveillance positions in the high flux isotope reactor  

SciTech Connect (OSTI)

The objective of this research was to determine improved thermal, epithermal, and fast fluxes and several responses at mechanical test surveillance location keys 2, 4, 5, and 7 of the pressure vessel of the Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) for the beginning of the fuel cycle. The purpose of the research was to provide essential flux data in support of radiation embrittlement studies of the pressure vessel shell and beam tubes at some of the important locations.

Pace, J.V. III; Slater, C.O.; Smith, M.S.

1993-11-01T23:59:59.000Z

24

BROOKHAVEN NATIONAL LABORATORY'S HIGH FLUX BEAM REACTOR  

E-Print Network [OSTI]

irradiation experiments with a variety of neutron energy spectra for neutron activation analysis, producing HISTORY #12;2 Erecting the dome in 1962. The building was designed as "confinement" to prevent leakage of radioactive materials into the environment. The interior of the building is maintained at a pressure slightly

Ohta, Shigemi

25

Spheromak reactor with poloidal flux-amplifying transformer  

DOE Patents [OSTI]

An inductive transformer in the form of a solenoidal coils aligned along the major axis of a flux core induces poloidal flux along the flux core's axis. The current in the solenoidal coil is then reversed resulting in a poloidal flux swing and the conversion of a portion of the poloidal flux to a toroidal flux in generating a spheromak plasma wherein equilibrium approaches a force-free, minimum Taylor state during plasma formation, independent of the initial conditions or details of the formation. The spheromak plasma is sustained with the Taylor state maintained by oscillating the currents in the poloidal and toroidal field coils within the plasma-forming flux core. The poloidal flux transformer may be used either as an amplifier stage in a moving plasma reactor scenario for initial production of a spheromak plasma or as a method for sustaining a stationary plasma and further heating it. The solenoidal coil embodiment of the poloidal flux transformer can alternately be used in combination with a center conductive cylinder aligned along the length and outside of the solenoidal coil. This poloidal flux-amplifying inductive transformer approach allows for a relaxation of demanding current carrying requirements on the spheromak reactor's flux core, reduces plasma contamination arising from high voltage electrode discharge, and improves the efficiency of poloidal flux injection.

Furth, Harold P. (Princeton, NJ); Janos, Alan C. (East Windsor, NJ); Uyama, Tadao (Osaka, JP); Yamada, Masaaki (Lawrenceville, NJ)

1987-01-01T23:59:59.000Z

26

High flux isotope reactor cold source preconceptual design study report  

SciTech Connect (OSTI)

In February 1995, the deputy director of Oak Ridge National Laboratory (ORNL) formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced Neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. The anticipated cold source will consist of a cryogenic LH{sub 2} moderator plug, a cryogenic pump system, a refrigerator that uses helium gas as a refrigerant, a heat exchanger to interface the refrigerant with the hydrogen loop, liquid hydrogen transfer lines, a gas handling system that includes vacuum lines, and an instrumentation and control system to provide constant system status monitoring and to maintain system stability. The scope of this project includes the development, design, safety analysis, procurement/fabrication, testing, and installation of all of the components necessary to produce a working cold source within an existing HFIR beam tube. This project will also include those activities necessary to transport the cold neutron beam to the front face of the present HFIR beam room. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and research and development (R and D), (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the preconceptual phase and establishes the concept feasibility. The information presented includes the project scope, the preliminary design requirements, the preliminary cost and schedule, the preliminary performance data, and an outline of the various plans for completing the project.

Selby, D.L.; Bucholz, J.A.; Burnette, S.E. [and others

1995-12-01T23:59:59.000Z

27

Reactor Neutrino Flux Uncertainty Suppression on Multiple Detector Experiments  

E-Print Network [OSTI]

This publication provides a coherent treatment for the reactor neutrino flux uncertainties suppression, specially focussed on the latest $\\theta_{13}$ measurement. The treatment starts with single detector in single reactor site, most relevant for all reactor experiments beyond $\\theta_{13}$. We demonstrate there is no trivial error cancellation, thus the flux systematic error can remain dominant even after the adoption of multi-detector configurations. However, three mechanisms for flux error suppression have been identified and calculated in the context of Double Chooz, Daya Bay and RENO sites. Our analysis computes the error {\\it suppression fraction} using simplified scenarios to maximise relative comparison among experiments. We have validated the only mechanism exploited so far by experiments to improve the precision of the published $\\theta_{13}$. The other two newly identified mechanisms could lead to total error flux cancellation under specific conditions and are expected to have major implications o...

Cucoanes, Andi; Cabrera, Anatael; Fallot, Muriel; Onillon, Anthony; Obolensky, Michel; Yermia, Frederic

2015-01-01T23:59:59.000Z

28

Neutron beam characterization at the Neutron Radiography Reactor (NRAD)  

SciTech Connect (OSTI)

The Neutron Radiography Reactor (NRAD) is a 250-kW TRIGA Reactor operated by Argonne National Laboratory and is located near Idaho Falls, Idaho. The reactor and its facilities regarding radiography are detailed in another paper at this conference; this paper summarizes neutron flux measurements and calculations that have been performed to better understand and potentially improve the neutronics characteristics of the reactor.

Imel, G.R.; Urbatsch, T.; Pruett, D.P.; Ross, J.R.

1990-01-01T23:59:59.000Z

29

Reactor Neutrino Flux Uncertainty Suppression on Multiple Detector Experiments  

E-Print Network [OSTI]

This publication provides a coherent treatment for the reactor neutrino flux uncertainties suppression, specially focussed on the latest $\\theta_{13}$ measurement. The treatment starts with single detector in single reactor site, most relevant for all reactor experiments beyond $\\theta_{13}$. We demonstrate there is no trivial error cancellation, thus the flux systematic error can remain dominant even after the adoption of multi-detector configurations. However, three mechanisms for flux error suppression have been identified and calculated in the context of Double Chooz, Daya Bay and RENO sites. Our analysis computes the error {\\it suppression fraction} using simplified scenarios to maximise relative comparison among experiments. We have validated the only mechanism exploited so far by experiments to improve the precision of the published $\\theta_{13}$. The other two newly identified mechanisms could lead to total error flux cancellation under specific conditions and are expected to have major implications on the global $\\theta_{13}$ knowledge today. First, Double Chooz, in its final configuration, is the only experiment benefiting from a negligible reactor flux error due to a $\\sim$90\\% geometrical suppression. Second, Daya Bay and RENO could benefit from their partial geometrical cancellation, yielding a potential $\\sim$50\\% error suppression, thus significantly improving the global $\\theta_{13}$ precision today. And third, we illustrate the rationale behind further error suppression upon the exploitation of the inter-reactor error correlations, so far neglected. So, our publication is a key step forward in the context of high precision neutrino reactor experiments providing insight on the suppression of their intrinsic flux error uncertainty, thus affecting past and current experimental results, as well as the design of future experiments.

Andi Cucoanes; Pau Novella; Anatael Cabrera; Muriel Fallot; Anthony Onillon; Michel Obolensky; Frederic Yermia

2015-01-02T23:59:59.000Z

30

CRAD, Engineering- Oak Ridge National Laboratory High Flux Isotope Reactor  

Broader source: Energy.gov [DOE]

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Engineering Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

31

CRAD, Maintenance- Oak Ridge National Laboratory High Flux Isotope Reactor  

Broader source: Energy.gov [DOE]

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Maintenance Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

32

CRAD, Training- Oak Ridge National Laboratory High Flux Isotope Reactor  

Broader source: Energy.gov [DOE]

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Training Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

33

CRAD, Management- Oak Ridge National Laboratory High Flux Isotope Reactor  

Broader source: Energy.gov [DOE]

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Management in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

34

E-Print Network 3.0 - argonne high flux reactor Sample Search...  

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

for: argonne high flux reactor Page: << < 1 2 3 4 5 > >> 1 Thirteenth National School on Neutron and X-ray Scattering Summary: Neutron Source and High Flux Isotope Reactor...

35

Conversion feasibility studies for the Grenoble high flux reactor  

SciTech Connect (OSTI)

Feasibility studies for conversion of the High Flux Reactor (RHF) at Grenoble France have been performed at the Argonne National Laboratory in cooperation with the Institut Laue-Langevin (ILL). The uranium densities required for conversion of the RHF to reduced enrichment fuels were computed to be 7.9 g/cm{sup 3} with 20% enrichment, 4.8 g/cm{sup 3} with 29% enrichment, and 2.8 g/cm{sup 3} with 45% enrichment. Thermal flux reductions at the peak in the heavy water reflector were computed to be 3% with 45% enriched fuel and 7% with 20% enriched fuel. In each case, the reactor's 44 day cycle length was preserved and no changes were made in the fuel element geometry. If the cladding thickness could be reduced from 0.38 mm to 0.30 mm, the required uranium density with 20% enrichment would be about 6.0 g/cm{sup 3} and the thermal flux reduction at the peak in the heavy water reflector would be about 7%. Significantly higher uranium densities are required in the RHF than in heavy water reactors with more conventional designs because the neutron spectrum is much harder in the RHF. Reduced enrichment fuels with the uranium densities required for use in the RHF are either not available or are not licensable at the present time. 6 refs., 6 figs., 3 tabs.

Mo, S.C.; Matos, J.E.

1989-01-01T23:59:59.000Z

36

High Flux Metallic Membranes for Hydrogen Recovery and Membrane Reactors  

SciTech Connect (OSTI)

We made and tested over 250 new alloys for use as lower cost, higher flux hydrogen extraction membrane materials. Most of these were intermetallic, or contained significant intermetallic content, particularly based on B2 alloy compositions with at least one refractory component; B2 intermetallics resemble BCC alloys, in structure, but the atoms have relatively fixed positions, with one atom at the corners of the cube, the other at the centers. The target materals we were looking for would contain little or no expensive elements, no strongly toxic or radioactive elements, would have high flux to hydrogen, while being fabricable, brazable, and relatively immune to hydrogen embrittlement and corrosion in operation. The best combination of properties of the membrane materials we developed was, in my opinion, a Pd-coated membrane consisting of V -9 atomic % Pd. This material was relatively cheap, had 5 times the flux of Pd under the same pressure differential, was reasonably easy to fabricate and braze, and not bad in terms of embrittlement. Based on all these factors we project, about 1/3 the cost of Pd, on an area basis for a membrane designed to last 20 years, or 1/15 the cost on a flux basis. Alternatives to this membrane replaced significant fractions of the Pd with Ni and or Co. The cost for these membranes was lower, but so was the flux. We produced successful brazed products from the membrane materials, and made them into flat sheets. We tested, unsuccessfully, several means of fabricating thematerials into tubes, and eventually built a membrane reactor using a new, flat-plate design: a disc and doughnut arrangement, a design that seems well- suited to clean hydrogen production from coal. The membranes and reactor were tested successfully at Western Research. A larger equipment company (Chart Industries) produced similar results using a different flat-plate reactor design. Cost projections of the membrane are shown to be attractive.

Buxbaum, Robert

2010-06-30T23:59:59.000Z

37

Irradiation research capabilities at HFIR (High Flux Isotope Reactor) and ANS (Advanced Neutron Source)  

SciTech Connect (OSTI)

A variety of materials irradiation facilities exist in the High Flux Isotope Reactor (HFIR) and are planned for the Advanced Neutron Source (ANS) reactor. In 1986 the HFIR Irradiation Facilities Improvement (HIFI) project began modifications to the HFIR which now permit the operation of two instrumented capsules in the target region and eight capsules of 46-mm OD in the RB region. Thus, it is now possible to perform instrumented irradiation experiments in the highest continuous flux of thermal neutrons available in the western world. The new RB facilities are now large enough to permit neutron spectral tailoring of experiments and the modified method of access to these facilities permit rotation of experiments thereby reducing fluence gradients in specimens. A summary of characteristics of irradiation facilities in HFIR is presented. The ANS is being designed to provide the highest thermal neutron flux for beam facilities in the world. Additional design goals include providing materials irradiation and transplutonium isotope production facilities as good, or better than, HFIR. The reference conceptual core design consists of two annular fuel elements positioned one above the other instead of concentrically as in the HFIR. A variety of materials irradiation facilities with unprecedented fluxes are being incorporated into the design of the ANS. These will include fast neutron irradiation facilities in the central hole of the upper fuel element, epithermal facilities surrounding the lower fuel element, and thermal facilities in the reflector tank. A summary of characteristics of irradiation facilities presently planned for the ANS is presented. 2 tabs.

Thoms, K.R.

1990-01-01T23:59:59.000Z

38

Determining Reactor Flux from Xenon-136 and Cesium-135 in Spent Fuel  

E-Print Network [OSTI]

The ability to infer the reactor flux from spent fuel or seized fissile material would enhance the tools of nuclear forensics and nuclear nonproliferation significantly. We show that reactor flux can be inferred from the ratios of xenon-136 to xenon-134 and cesium-135 to cesium-137. If the average flux of a reactor is known, the flux inferred from measurements of spent fuel could help determine whether that spent fuel was loaded as a blanket or close to the mid-plane of the reactor. The cesium ratio also provides information on reactor shutdowns during the irradiation of fuel, which could prove valuable for identifying the reactor in question through comparisons with satellite reactor heat monitoring data. We derive analytic expressions for these correlations and compare them to experimental data and to detailed reactor burn simulations. The enrichment of the original uranium fuel affects the correlations by up to 3 percent, but only at high flux.

A. C. Hayes; Gerard Jungman

2012-05-30T23:59:59.000Z

39

Calculation of Heating Values for the High Flux Isotope Reactor  

SciTech Connect (OSTI)

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

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

2012-01-01T23:59:59.000Z

40

Design and optimization of a high thermal flux research reactor via Kriging-based algorithm  

E-Print Network [OSTI]

In response to increasing demands for the services of research reactors, a 5 MW LEU-fueled research reactor core is developed and optimized to provide high thermal flux within specified limits upon thermal hydraulic ...

Kempf, Stephanie Anne

2011-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "flux beam reactor" 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

Fabrication of control rods for the High Flux Isotope Reactor  

SciTech Connect (OSTI)

The High Flux Isotope Reactor (HFIR) is a research-type nuclear reactor that was designed and built in the early 1960s and has been in continuous operation since its initial criticality in 1965. Under current plans, the HFIR is expected to continue in operation until 2035. This report updates ORNL/TM-9365, Fabrication Procedure for HFIR Control Plates, which was mainly prepared in the early 1970's but was not issued until 1984, and reflects process changes, lessons learned in the latest control rod fabrication campaign, and suggested process improvements to be considered in future campaigns. Most of the personnel involved with the initial development of the processes and in part campaigns have retired or will retire soon. Because their unlikely availability in future campaigns, emphasis has been placed on providing some explanation of why the processes were selected and some discussions about the importance of controlling critical process parameters. Contained in this report is a description of the function of control rods in the reactor, the brief history of the development of control rod fabrication processes, and a description of procedures used in the fabrication of control rods. A listing of the controlled documents and procedures used in the last fabrication campaigns is referenced in Appendix A.

Sease, J.D.

1998-03-01T23:59:59.000Z

42

High Flux Isotope Reactor (HFIR) | U.S. DOE Office of Science...  

Office of Science (SC) Website

(SUF) Division SUF Home About User Facilities User Facilities Dev X-Ray Light Sources Neutron Scattering Facilities High Flux Isotope Reactor (HFIR) Lujan Neutron Scattering...

43

Low-Enriched Uranium Fuel Design with Two-Dimensional Grading for the High Flux Isotope Reactor  

SciTech Connect (OSTI)

An engineering design study of the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel is ongoing at Oak Ridge National Laboratory. The computational models developed during fiscal year 2010 to search for an LEU fuel design that would meet the requirements for the conversion and the results obtained with these models are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating HEU fuel core. The results obtained indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations under the assumption that the operating power for the reactor fueled with LEU can be increased from the current value of 85 MW to 100 MW.

Ilas, Germina [ORNL; Primm, Trent [ORNL

2011-05-01T23:59:59.000Z

44

FFTF (Fast Flux Test Facility) reactor shutdown system reliability reevaluation  

SciTech Connect (OSTI)

The reliability analysis of the Fast Flux Test Facility reactor shutdown system was reevaluated. Failure information based on five years of plant operating experience was used to verify original reliability numbers or to establish new ones. Also, system modifications made subsequent to performance of the original analysis were incorporated into the reevaluation. Reliability calculations and sensitivity analyses were performed using a commercially available spreadsheet on a personal computer. The spreadsheet was configured so that future failures could be tracked and compared with expected failures. A number of recommendations resulted from the reevaluation including both increased and decreased surveillance intervals. All recommendations were based on meeting or exceeding existing reliability goals. Considerable cost savings will be incurred upon implementation of the recommendations.

Pierce, B.F.

1986-07-01T23:59:59.000Z

45

High Flux Isotope Reactor system RELAP5 input model  

SciTech Connect (OSTI)

A thermal-hydraulic computational model of the High Flux Isotope Reactor (HFIR) has been developed using the RELAP5 program. The purpose of the model is to provide a state-of-the art thermal-hydraulic simulation tool for analyzing selected hypothetical accident scenarios for a revised HFIR Safety Analysis Report (SAR). The model includes (1) a detailed representation of the reactor core and other vessel components, (2) three heat exchanger/pump cells, (3) pressurizing pumps and letdown valves, and (4) secondary coolant system (with less detail than the primary system). Data from HFIR operation, component tests, tests in facility mockups and the HFIR, HFIR specific experiments, and other pertinent experiments performed independent of HFIR were used to construct the model and validate it to the extent permitted by the data. The detailed version of the model has been used to simulate loss-of-coolant accidents (LOCAs), while the abbreviated version has been developed for the operational transients that allow use of a less detailed nodalization. Analysis of station blackout with core long-term decay heat removal via natural convection has been performed using the core and vessel portions of the detailed model.

Morris, D.G.; Wendel, M.W.

1993-01-01T23:59:59.000Z

46

A NOVEL MICROMEGAS DETECTOR FOR IN-CORE NUCLEAR REACTOR NEUTRON FLUX MEASUREMENTS  

E-Print Network [OSTI]

1 A NOVEL MICROMEGAS DETECTOR FOR IN-CORE NUCLEAR REACTOR NEUTRON FLUX MEASUREMENTS S. ANDRIAMONJE Talence Cedex, France Future fast nuclear reactors designed for energy production and transmutation to neutron detection inside nuclear reactor is given. The advantage of this detector over conventional

Paris-Sud XI, Université de

47

A Review of Proposed Upgrades to the High Flux Isotope Reactor and Potential Impacts to Reactor Vessel Integrity  

SciTech Connect (OSTI)

The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) was scheduled in October 2000 to implement design upgrades that include the enlargement of the HB-2 and HB-4 beam tubes. Higher dose rates and higher radiation embrittlement rates were predicted for the two beam-tube nozzles and surrounding vessel areas. ORNL had performed calculations for the upgraded design to show that vessel integrity would be maintained at acceptable levels. Pacific Northwest National Laboratory (PNNL) was requested by the U.S. Department of Energy Headquarters (DOE/HQ) to perform an independent peer review of the ORNL evaluations. PNNL concluded that the calculated probabilities of failure for the HFIR vessel during hydrostatic tests and for operational conditions as estimated by ORNL are an acceptable basis for selecting pressures and test intervals for hydrostatic tests and for justifying continued operation of the vessel. While there were some uncertainties in the embrittlement predictions, the ongoing efforts at ORNL to measure fluence levels at critical locations of the vessel wall and to test materials from surveillance capsules should be effective in dealing with embrittlement uncertainties. It was recommended that ORNL continue to update their fracture mechanics calculations to reflect methods and data from ongoing research for commercial nuclear power plants. Such programs should provide improved data for vessel fracture mechanics calculations.

Simonen, Fredric A.

2001-05-31T23:59:59.000Z

48

Emulation of reactor irradiation damage using ion beams  

SciTech Connect (OSTI)

The continued operation of existing light water nuclear reactors and the development of advanced nuclear reactor depend heavily on understanding how damage by radiation to levels degrades materials that serve as the structural components in reactor cores. The first high dose ion irradiation experiments on a ferritic-martensitic steel showing that ion irradiation closely emulates the full radiation damage microstructure created in-reactor are described. Ferritic-martensitic alloy HT9 (heat 84425) in the form of a hexagonal fuel bundle duct (ACO-3) accumulated 155 dpa at an average temperature of 443°C in the Fast Flux Test Facility (FFTF). Using invariance theory as a guide, irradiation of the same heat was conducted using self-ions (Fe++) at 5 MeV at a temperature of 460°C and to a dose of 188 displacements per atom. The void swelling was nearly identical between the two irradiations and the size and density of precipitates and loops following ion irradiation are within a factor of two of those for neutron irradiation. The level of agreement across all of the principal microstructure changes between ion and reactor irradiations establishes the capability of tailoring ion irradiations to emulate the reactor-irradiated microstructure.

G. S. Was; Z. Jiao; E. Beckett; A. M. Monterrosa; O. Anderoglu; B. H. Sencer; M. Hackett

2014-10-01T23:59:59.000Z

49

Study of a multi-beam accelerator driven thorium reactor  

SciTech Connect (OSTI)

The primary advantages that accelerator driven systems have over critical reactors are: (1) Greater flexibility regarding the composition and placement of fissile, fertile, or fission product waste within the blanket surrounding the target, and (2) Potentially enhanced safety brought about by operating at a sufficiently low value of the multiplication factor to preclude reactivity induced events. The control of the power production can be achieved by vary the accelerator beam current. Furthermore, once the beam is shut off the system shuts down. The primary difference between the operation of an accelerator driven system and a critical system is the issue of beam interruptions of the accelerator. These beam interruptions impose thermo-mechanical loads on the fuel and mechanical components not found in critical systems. Studies have been performed to estimate an acceptable number of trips, and the value is significantly less stringent than had been previously estimated. The number of acceptable beam interruptions is a function of the length of the interruption and the mission of the system. Thus, for demonstration type systems and interruption durations of 1sec < t < 5mins, and t > 5mins 2500/yr and 50/yr are deemed acceptable. However, for industrial scale power generation without energy storage type systems and interruption durations of t < 1sec., 1sec < t < 10secs., 10secs < t < 5mins, and t > 5mins, the acceptable number of interruptions are 25000, 2500, 250, and 3 respectively. However, it has also been concluded that further development is required to reduce the number of trips. It is with this in mind that the following study was undertaken. The primary focus of this study will be the merit of a multi-beam target system, which allows for multiple spallation sources within the target/blanket assembly. In this manner it is possible to ameliorate the effects of sudden accelerator beam interruption on the surrounding reactor, since the remaining beams will still be supplying source neutrons. The proton beam will be assumed to have an energy of 1 GeV, and the target material will be natural lead, which will also be the coolant for the reactor assembly. Three proton beam arrangements will be considered, first a single beam (the traditional arrangement) with an entry at the assembly center, two more options will consist of three and six entry locations. The reactor fuel assembly parameters will be based on those of the S-PRISM fast reactor proposed by GE, and the fuel composition and type will be based on that proposed by Aker Solutions for use in their accelerator driven thorium reactor. The following table summarizes the parameters to be used in this study. The isotopic composition of the fertile material is 100% Th-232, and the plutonium isotopic distribution corresponds to that characteristic of the discharge from a typical LWR, following five years of decay. Thus, the isotopic distribution for the plutonium is; Pu-238 2.5%, Pu-239 53.3%, Pu-240 25.1%, Pu-241 11.8%, and Pu-242 7.3%.

Ludewig, H.; Aronson, A.

2011-03-01T23:59:59.000Z

50

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

51

Plasma focus ion beam fluence and flux—For various gases  

SciTech Connect (OSTI)

A recent paper derived benchmarks for deuteron beam fluence and flux in a plasma focus (PF) [S. Lee and S. H. Saw, Phys. Plasmas 19, 112703 (2012)]. In the present work we start from first principles, derive the flux equation of the ion beam of any gas; link to the Lee Model code and hence compute the ion beam properties of the PF. The results show that, for a given PF, the fluence, flux, ion number and ion current decrease from the lightest to the heaviest gas except for trend-breaking higher values for Ar fluence and flux. The energy fluence, energy flux, power flow, and damage factors are relatively constant from H{sub 2} to N{sub 2} but increase for Ne, Ar, Kr and Xe due to radiative cooling and collapse effects. This paper provides much needed benchmark reference values and scaling trends for ion beams of a PF operated in any gas.

Lee, S. [Centre for Plasma Research, INTI International University, 71800 Nilai (Malaysia) [Centre for Plasma Research, INTI International University, 71800 Nilai (Malaysia); Institute for Plasma Focus Studies, 32 Oakpark Drive, Chadstone 3148 (Australia); Physics Department, University of Malaya (Malaysia); Saw, S. H. [Centre for Plasma Research, INTI International University, 71800 Nilai (Malaysia) [Centre for Plasma Research, INTI International University, 71800 Nilai (Malaysia); Institute for Plasma Focus Studies, 32 Oakpark Drive, Chadstone 3148 (Australia)

2013-06-15T23:59:59.000Z

52

The determination of neutron flux in the Texas A & M triga reactor during pulse and steady-state operations  

E-Print Network [OSTI]

-state operation. Neutron flux measurement during a pulse presents an additional problem in that the flux levels vary during the rapid rise and fall in reactor power. The power level transient of the reactor was followed, using the current output of a boron-10... as a flux monitor only at low power. levels or neutron fluxes. The antimony flux monitor in the steady-state flux measurement showed the same type of increase in flux magni- tude as that of the pulse measurement. The high flux values at steady...

O'Donnell, John Joseph

2012-06-07T23:59:59.000Z

53

High Flux Isotope Reactor named Nuclear Historic Landmark | ornl...  

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

late 1950s as a production reactor to meet anticipated demand for transuranic isotopes ("heavy" elements such as plutonium and curium). HFIR today is a DOE Office of Science User...

54

Development of High Flux Isotope Reactor (HFIR) subcriticality monitoring methods  

SciTech Connect (OSTI)

Use of subcritical source multiplication measurements during refueling has been investigated as a possible replacement for out-of-reactor subcriticality measurements formerly made on fresh HFIR fuel elements at the ORNL Critical Experiment Facility. These measurements have been used in the past for preparation of estimated critical rod positions, and as a partial verification, prior to reactor startup, that the requirements for operational shutdown margin would be met. Results of subcritical count rate data collection during recent HFIR refuelings and supporting calculations are described illustrating the intended measurement method and its expected uncertainty. These results are compared to historical uses of the out-of-reactor core measurements and their accuracy requirements, and a planned in-reactor test is described which will establish the sensitivity of the method and calibrate it for future routine use during HFIR refueling. 2 refs., 1 fig., 2 tabs.

Rothrock, R.B.

1991-01-01T23:59:59.000Z

55

Homogeneous fast-flux isotope-production reactor  

DOE Patents [OSTI]

A method is described for producing tritium in a liquid metal fast breeder reactor. Lithium target material is dissolved in the liquid metal coolant in order to facilitate the production and removal of tritium.

Cawley, W.E.; Omberg, R.P.

1982-08-19T23:59:59.000Z

56

High flux isotope reactor: Quarterly report October through December 1986  

SciTech Connect (OSTI)

Two routine cycles of operation of the HFIR reactor were completed during the quarter. The shutdowns to end these cycles were both scheduled. The end-of-cycle 287 shutdown was extended indefinitely to investigate the embrittlement of reactor vessel materials due to radiation damage. The reactor remains down at the end of the quarter. Following the scheduled end-of-cycle 287 shutdown period, subsequent shutdown time was designated as unscheduled. The two scheduled shutdowns, fourth quarter downtime resulting from a third quarter scheduled shutdown, and the extended unscheduled shutdown account for the low 44.2% on-stream time for the quarter. The scheduled control plate replacement and vessel internals inspection was completed at the end-of-cycle 287. The inspection revealed a blister on control cylinder 9. This flaw was attributed to a manufacturing defect.

Corbett, B.L.; Farrar, M.B.

1987-04-01T23:59:59.000Z

57

High flux isotope reactor. Quarterly report, January-March 1982  

SciTech Connect (OSTI)

Routine reactor operation with four end-of-cycle shutdowns and one scheduled shutdown for training purposes resulted in an on-stream time of 92.1% for the quarter. The outer control plates were changed. The control plate track guide bearings, the control plate extension tubes, and the shock absorbers were replaced and a semiannual core component inspection was made. Cracks were discovered in the outermost ring of the beryllium reflector.

Corbett, B.L.; Poteet, K.H.

1982-06-01T23:59:59.000Z

58

High Flux Isotope Reactor quarterly report, April-June 1986  

SciTech Connect (OSTI)

Four routine cycles of operation were completed during the second quarter. Four scheduled end-of-cycle shutdowns, three other scheduled shutdowns, and two unscheduled shutdowns resulted in an on-stream time of 90.7%. The control plates and cylinder were replaced during the end-of-cycle 281 shutdown. Control plate set 2 and previously unirradiated cylinder No. 9 were inserted into the reactor.

Corbett, B.L.; Farrar, M.B.

1986-08-01T23:59:59.000Z

59

Apparatus for high flux photocatalytic pollution control using a rotating fluidized bed reactor  

DOE Patents [OSTI]

An apparatus based on optimizing photoprocess energetics by decoupling of the process energy efficiency from the DRE for target contaminants. The technique is applicable to both low- and high-flux photoreactor design and scale-up. An apparatus for high-flux photocatalytic pollution control is based on the implementation of multifunctional metal oxide aerogels and other media in conjunction with a novel rotating fluidized particle bed reactor.

Tabatabaie-Raissi, Ali; Muradov, Nazim Z.; Martin, Eric

2003-06-24T23:59:59.000Z

60

Awareness, Preference, Utilization, and Messaging Research for the Spallation Neutron Source and High Flux Isotope Reactor  

SciTech Connect (OSTI)

Oak Ridge National Laboratory (ORNL) offers the scientific community unique access to two types of world-class neutron sources at a single site - the Spallation Neutron Source (SNS) and the High Flux Isotope Reactor (HFIR). The 85-MW HFIR provides one of the highest steady-state neutron fluxes of any research reactor in the world, and the SNS is one of the world's most intense pulsed neutron beams. Management of these two resources is the responsibility of the Neutron Sciences Directorate (NScD). NScD commissioned this survey research to develop baseline information regarding awareness of and perceptions about neutron science. Specific areas of investigative interest include the following: (1) awareness levels among those in the scientific community about the two neutron sources that ORNL offers; (2) the level of understanding members of various scientific communities have regarding benefits that neutron scattering techniques offer; and (3) any perceptions that negatively impact utilization of the facilities. NScD leadership identified users of two light sources in North America - the Advanced Photon Source (APS) at Argonne National Laboratory and the National Synchrotron Light Source (NSLS) at Brookhaven National Laboratory - as key publics. Given the type of research in which these scientists engage, they would quite likely benefit from including the neutron techniques available at SNS and HFIR among their scientific investigation tools. The objective of the survey of users of APS, NSLS, SNS, and HFIR was to explore awareness of and perceptions regarding SNS and HFIR among those in selected scientific communities. Perceptions of SNS and FHIR will provide a foundation for strategic communication plan development and for developing key educational messages. The survey was conducted in two phases. The first phase included qualitative methods of (1) key stakeholder meetings; (2) online interviews with user administrators of APS and NSLS; and (3) one-on-one interviews and traditional and online focus groups with scientists. The latter include SNS, HFIR, and APS users as well as scientists at ORNL, some of whom had not yet used HFIR and/or SNS. These approaches informed development of the second phase, a quantitative online survey. The survey consisted of 16 questions and 7 demographic categorizations, 9 open-ended queries, and 153 pre-coded variables and took an average time of 18 minutes to complete. The survey was sent to 589 SNS/HFIR users, 1,819 NSLS users, and 2,587 APS users. A total of 899 individuals provided responses for this study: 240 from NSLS; 136 from SNS/HFIR; and 523 from APS. The overall response rate was 18%.

Bryant, Rebecca [Bryant Research, LLC; Kszos, Lynn A [ORNL

2011-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "flux beam reactor" 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

Modelling the Electron Beam Welding of Nuclear Reactor Pressure Vessel Steel  

E-Print Network [OSTI]

Modelling the Electron Beam Welding of Nuclear Reactor Pressure Vessel Steel Christopher J. Duffy fabrication of thick-section steel for critical components such as reactor pressure vessels. Electron beam weld tests performed by Rolls-Royce and The Welding Institute of SA 508 Grade 3 and SA 508 Grade 4N

Cambridge, University of

62

Electron beams in material testing for fusion reactors  

SciTech Connect (OSTI)

The Plasma Material Test Facility at Sandia National Laboratory is currently involved in a multi-year program of thermally testing materials and high heat flux components for eventual use in fusion reactors. The major test system in the facility is the EB-1200 System, which uses two EH600S Von Ardenne electron guns to provide up to 1.2 MW of heating for large material samples and components. In this paper, the advantages of using electron guns for this type of testing will be discussed. Gun performance characteristics will be presented, as well as electronic techniques used to provide the rapidly varying power levels and complex scan patterns required for complicated targets. Diagnostic capabilities of importance to the experimental program will also be discussed.

Ganley, J.T.; McDonald, J.M.; Youchison, D.L. [Sandia National Laboratories, Albuquerque, NM (United States)

1995-12-31T23:59:59.000Z

63

FRP Retrofit of the Ring-Beam of a Nuclear Reactor Containment Structure  

E-Print Network [OSTI]

SP·215-18 FRP Retrofit of the Ring-Beam of a Nuclear Reactor Containment Structure by M. Demers. A for the storage of the moderately contaminated nuclear reactor. The enforcement of more rigorous environmental. 1. HISTORY 1.1 Decommissioning of the Reactor The Gentilly-I nuclear power plant, located

64

High Flux Isotope Reactor (HFIR) | Nuclear Science | ORNL  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth7-1D: Vegetation ProposedUsingFun withconfinement plasmas in the Madison SymmetricHigh Carbon| ArgonneHigh Flux

65

CRAD, Maintenance- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR  

Broader source: Energy.gov [DOE]

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Maintenance Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor.

66

CRAD, Conduct of Operations- Oak Ridge National Laboratory High Flux Isotope Reactor  

Broader source: Energy.gov [DOE]

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February, 2007 assessment of the Conduct of Operations Program in preparation for restart of the Oak Ridge National Laboratory, High Flux Isotope Reactor.

67

CRAD, Conduct of Operations- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR  

Broader source: Energy.gov [DOE]

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February, 2007 assessment of the Conduct of Operations Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory, High Flux Isotope Reactor.

68

CRAD, Management- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR  

Broader source: Energy.gov [DOE]

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Management portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor.

69

CRAD, Emergency Management- Oak Ridge National Laboratory High Flux Isotope Reactor  

Broader source: Energy.gov [DOE]

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Emergency Management Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

70

CRAD, Engineering- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR  

Broader source: Energy.gov [DOE]

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Engineering Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor.

71

CRAD, Environmental Protection- Oak Ridge National Laboratory High Flux Isotope Reactor  

Broader source: Energy.gov [DOE]

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Environmental Compliance Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

72

CRAD, Configuration Management- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR  

Broader source: Energy.gov [DOE]

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Configuration Management Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory, High Flux Isotope Reactor.

73

CRAD, Emergency Management- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR  

Broader source: Energy.gov [DOE]

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Emergency Management Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor.

74

CRAD, Configuration Management- Oak Ridge National Laboratory High Flux Isotope Reactor  

Broader source: Energy.gov [DOE]

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Configuration Management Program in preparation for restart of the Oak Ridge National Laboratory, High Flux Isotope Reactor.

75

CRAD, Training- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR  

Broader source: Energy.gov [DOE]

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Training Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor.

76

CRAD, Nuclear Safety- Oak Ridge National Laboratory High Flux Isotope Reactor  

Broader source: Energy.gov [DOE]

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Nuclear Safety Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

77

CRAD, Occupational Safety & Health- Oak Ridge National Laboratory High Flux Isotope Reactor  

Broader source: Energy.gov [DOE]

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Industrial Safety and Hygiene Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

78

CRAD, Radiological Controls- Oak Ridge National Laboratory High Flux Isotope Reactor  

Broader source: Energy.gov [DOE]

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Radiation Protection Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

79

CRAD, Safety Basis- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR  

Broader source: Energy.gov [DOE]

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Safety Basis portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor.

80

CRAD, Safety Basis- Oak Ridge National Laboratory High Flux Isotope Reactor  

Broader source: Energy.gov [DOE]

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Safety Basis in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

Note: This page contains sample records for the topic "flux beam reactor" 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

CRAD, Occupational Safety & Health- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR  

Broader source: Energy.gov [DOE]

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Occupational Safety and Health Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor.

82

CRAD, Quality Assurance- Oak Ridge National Laboratory High Flux Isotope Reactor  

Broader source: Energy.gov [DOE]

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Quality Assurance Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

83

Lessons Learned in the Update of a Safety Limit for the High Flux Isotope Reactor  

SciTech Connect (OSTI)

A recent unreviewed safety question (USQ) regarding a portion of the High Flux Isotope Reactor (HFIR) transient decay heat removal analysis focused on applicability of a heat transfer correlation at the low flow end of reactor operations. During resolution of this issue, review of the correlations used to establish the safety limit (SL) on reactor flux-to-flow ratio revealed the need to change the magnitude of the SL at the low flow end of reactor operations and the need to update the hot spot fuel damage criteria to incorporate current knowledge involving parallel channel flow stability. Because of the original safety design strategy for the reactor, resolution of the issues for the flux-to-flow ratio involved reevaluation of all key process variable SLs and limiting control settings (LCSs) using the current version of the heat transfer analysis code for the reactor. Goals of the work involved updating and upgrading the SL analysis where necessary, while preserving the safety design strategy for the reactor. Changes made include revisions to the safety design criteria at low flows to address the USQ, update of the process- and analysis input-variable uncertainty considerations, and upgrade of the safety design criteria at high flow. The challenges faced during update/upgrade of this SL and LCS are typical of the problems found in the integration of safety into the design process for a complex facility. In particular, the problems addressed in the area of instrument uncertainties provide valuable lessons learned for establishment and configuration control of SLs for large facilities.

Cook, David Howard [ORNL

2009-01-01T23:59:59.000Z

84

Global neutrino data and recent reactor fluxes: status of three-flavour oscillation parameters  

E-Print Network [OSTI]

We present the results of a global neutrino oscillation data analysis within the three-flavour framework. We include latest results from the MINOS long-baseline experiment (including electron neutrino appearance as well as anti-neutrino data), updating all relevant solar (SK II+III), atmospheric (SK I+II+III) and reactor (KamLAND) data. Furthermore, we include a recent re-calculation of the anti-neutrino fluxes emitted from nuclear reactors. These results have important consequences for the analysis of reactor experiments and in particular for the status of the mixing angle $\\theta_{13}$. In our recommended default analysis we find from the global fit that the hint for non-zero $\\theta_{13}$ remains weak, at 1.8$\\sigma$ for both neutrino mass hierarchy schemes. However, we discuss in detail the dependence of these results on assumptions concerning the reactor neutrino analysis.

Thomas Schwetz; Mariam Tórtola; J. W. F. Valle

2011-03-28T23:59:59.000Z

85

Feasibility analyses for HEU to LEU fuel conversion of the LAUE Langivin Institute (ILL) High Flux Reactor (RHF).  

SciTech Connect (OSTI)

The High Flux Reactor (RHF) of the Laue Langevin Institute (ILL) based in Grenoble, France is a research reactor designed primarily for neutron beam experiments for fundamental science. It delivers one of the most intense neutron fluxes worldwide, with an unperturbed thermal neutron flux of 1.5 x 10{sup 15} n/cm{sup 2}/s in its reflector. The reactor has been conceived to operate at a nuclear power of 57 MW but currently operates at 52 MW. The reactor currently uses a Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most worldwide research and test reactors have already started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on a mixture of uranium and molybdenum (UMo) is expected to allow the conversion of compact high performance reactors like the RHF. This report presents the results of reactor design, performance and steady state safety analyses for conversion of the RHF from the use of HEU fuel to the use of UMo LEU fuel. The objective of this work was to show that is feasible, under a set of manufacturing assumptions, to design a new RHF fuel element that could safely replace the HEU element currently used. The new proposed design has been developed to maximize performance, minimize changes and preserve strong safety margins. Neutronics and thermal-hydraulics models of the RHF have been developed and qualified by benchmark against experiments and/or against other codes and models. The models developed were then used to evaluate the RHF performance if LEU UMo were to replace the current HEU fuel 'meat' without any geometric change to the fuel plates. Results of these direct replacement analyses have shown a significant degradation of the RHF performance, in terms of both neutron flux and cycle length. Consequently, ANL and ILL have collaborated to investigate alternative designs. A promising candidate design has been selected and studied, increasing the total amount of fuel without changing the external plate dimensions by relocating the burnable poison. In this way, changes required in the fuel element are reasonably small. With this new design, neutronics analyses have shown that performance could be maintained at a high level: 2 day decrease of cycle length (to 47.5 days at 58.3 MW) and 1-2% decrease of brightness in the cold and hot sources in comparison to the current typical operation. In addition, studies have shown that the thermal-hydraulic and shutdown margins for the proposed LEU design would satisfy technical specifications.

Stevens, J.; Tentner. A.; Bergeron, A.; Nuclear Engineering Division

2010-08-19T23:59:59.000Z

86

Seismic, high wind, tornado, and probabilistic risk assessments of the High Flux Isotope Reactor  

SciTech Connect (OSTI)

Natural phenomena analyses were performed on the High Flux Isotope Reactor (HFIR) Deterministic and probabilistic evaluations were made to determine the risks resulting from earthquakes, high winds, and tornadoes. Analytic methods in conjunction with field evaluations and an earthquake experience data base evaluation methods were used to provide more realistic results in a shorter amount of time. Plant modifications completed in preparation for HFIR restart and potential future enhancements are discussed. 5 figs.

Harris, S.P.; Stover, R.L.; Hashimoto, P.S.; Dizon, J.O. (EQE, Inc., San Francisco, CA (USA); Oak Ridge National Lab., TN (USA); EQE, Inc., San Francisco, CA (USA))

1989-01-01T23:59:59.000Z

87

Countercurrent flow limited (CCFL) heat flux in the high flux isotope reactor (HFIR) fuel element  

SciTech Connect (OSTI)

The countercurrent flow (CCF) performance in the fuel element region of the HFIR is examined experimentally and theoretically. The fuel element consists of two concentric annuli filled with aluminum clad fuel plates of 1.27 mm thickness separated by 1.27 mm flow channels. The plates are curved as they go radially outward to accomplish constant flow channel width and constant metal-to-coolant ratio. A full-scale HFIR fuel element mock-up is studied in an adiabatic air-water CCF experiment. A review of CCF models for narrow channels is presented along with the treatment of CCFs in system of parallel channels. The experimental results are related to the existing models and a mechanistic model for the annular'' CCF in a narrow channel is developed that captures the data trends well. The results of the experiment are used to calculate the CCFL heat flux of the HFIR fuel assembly. It was determined that the HFIR fuel assembly can reject 0.62 Mw of thermal power in the CCFL situation. 31 refs., 17 figs.

Ruggles, A.E.

1990-10-12T23:59:59.000Z

88

External event Probabilistic Risk Assessment for the High Flux Isotope Reactor (HFIR)  

SciTech Connect (OSTI)

The High Flux Isotope Reactor (HFIR) is a high performance isotope production and research reactor which has been in operation at Oak Ridge National Laboratory (ORNL) since 1965. In late 1986 the reactor was shut down as a result of discovery of unexpected neutron embrittlement of the reactor vessel. In January of 1988 a level 1 Probabilistic Risk Assessment (PRA) (excluding external events) was published as part of the response to the many reviews that followed the shutdown and for use by ORNL to prioritize action items intended to upgrade the safety of the reactor. A conservative estimate of the core damage frequency initiated by internal events for HFIR was 3.11 {times} 10{sup {minus}4}. In June 1989 a draft external events initiated PRA was published. The dominant contributions from external events came from seismic, wind, and fires. The overall external event contribution to core damage frequency is about 50% of the internal event initiated contribution and is dominated by seismic events.

Flanagan, G.F.; Johnson, D.H.; Buttemer, D.; Perla, H.F.; Chien, S.H. (Oak Ridge National Lab., TN (USA))

1989-01-01T23:59:59.000Z

89

Beyond ITER: Neutral beams for a demonstration fusion reactor (DEMO) (invited)  

SciTech Connect (OSTI)

In the development of magnetically confined fusion as an economically sustainable power source, International Tokamak Experimental Reactor (ITER) is currently under construction. Beyond ITER is the demonstration fusion reactor (DEMO) programme in which the physics and engineering aspects of a future fusion power plant will be demonstrated. DEMO will produce net electrical power. The DEMO programme will be outlined and the role of neutral beams for heating and current drive will be described. In particular, the importance of the efficiency of neutral beam systems in terms of injected neutral beam power compared to wallplug power will be discussed. Options for improving this efficiency including advanced neutralisers and energy recovery are discussed.

McAdams, R., E-mail: roy.mcadams@ccfe.ac.uk [EURATOM/CCFE Association, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom)

2014-02-15T23:59:59.000Z

90

A neutronic feasibility study for LEU conversion of the high flux isotope reactor (HFIR).  

SciTech Connect (OSTI)

A neutronic feasibility study was performed to determine the uranium densities that would be required to convert the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) from HEU (93%) to LEU (<20%)fuel. The LEU core that was studied is the same as the current HEU core, except for potential changes in the design of the fuel plates. The study concludes that conversion of HFIR from HEU to LEU fuel would require an advanced fuel with a uranium density of 6-7 gU/cm{sup 3} in the inner fuel element and 9-10 gU/cm{sup 3} in the outer fuel element to match the cycle length of the HEU core. LEU fuel with uranium density up to 4.8 gU/cm{sup 3} is currently qualified for research reactor use. Modifications in fuel grading and burnable poison distribution are needed to produce an acceptable power distribution.

Mo, S. C.

1998-01-14T23:59:59.000Z

91

Establishing Specifications for Low Enriched Uranium Fuel Operations Conducted Outside the High Flux Isotope Reactor Site  

SciTech Connect (OSTI)

The National Nuclear Security Administration (NNSA) has funded staff at Oak Ridge National Laboratory (ORNL) to study the conversion of the High Flux Isotope Reactor (HFIR) from the current, high enriched uranium fuel to low enriched uranium fuel. The LEU fuel form is a metal alloy that has never been used in HFIR or any HFIR-like reactor. This report provides documentation of a process for the creation of a fuel specification that will meet all applicable regulations and guidelines to which UT-Battelle, LLC (UTB) the operating contractor for ORNL - must adhere. This process will allow UTB to purchase LEU fuel for HFIR and be assured of the quality of the fuel being procured.

Pinkston, Daniel [ORNL; Primm, Trent [ORNL; Renfro, David G [ORNL; Sease, John D [ORNL

2010-10-01T23:59:59.000Z

92

Reactor Physics Studies of Reduced-Tantaulum-Content Control and Safety Elements for the High Flux Isotope Reactor  

SciTech Connect (OSTI)

Some of the unirradiated High Flux Isotope Reactor (HFIR) control elements discharged during the late 1990s were observed to have cladding damage--local swelling or blistering. The cladding damage was limited to the tantalum/europium interface of the element and is thought to result from interaction of hydrogen and europium to form a compound of lower density than europium oxide, thus leading to a ''blistering'' of the control plate cladding. Reducing the tantalum loading in the control plates should help preclude this phenomena. The impact of the change to the control plates on the operation of the reactor was assessed. Regarding nominal, steady-state reactor operation, the impact of the change in the power distribution in the core due to reduced tantalum content was calculated and found to be insignificant. The magnitude and impact of the change in differential control element worth was calculated, and the differential worths of reduced tantalum elements vs the current elements from equivalent-burnup critical configurations were determined to be unchanged within the accuracy of the computational method and relevant experimental measurements. The location of the critical control elements symmetric positions for reduced tantalum elements was found to be 1/3 in. less withdrawn relative to existing control elements regardless of the value of fuel cycle burnup (time in the fuel cycle). The magnitude and impact of the change in the shutdown margin (integral rod worth) was assessed and found to be unchanged. Differential safety element worth values for the reduced-tantalum-content elements were calculated for postulated accident conditions and were found to be greater than values currently assumed in HFIR safety analyses.

Primm, R.T., III

2003-11-01T23:59:59.000Z

93

Reactor physics input to the safety analysis report for the High Flux Isotope Reactor  

SciTech Connect (OSTI)

HFIR specific, few group neutron and coupled neutron-gamma libraries have been prepared. These are based on data from ENDF/B-V and beginning-of-life (BOL) conditions. The neutron library includes actinide data for curium target rods. Six critical experiments, collectively designated HFIR critical experiment 4, were analyzed. Calculated k-effective was 2% high at BOL-typical conditions but was 1.0 at end-of-life-typical conditions. The local power density distributions were calculated for each of the critical experiments. The axially averaged values at a given radius were frequently within experimental error. However at individual points, the calculated local power densities were significantly different from the experimentally derived values (several times greater than experimental uncertainty). A reassessment of the foil activation data with transport theory techniques seems desirable. Using the results of the critical experiments study, a model of current HFIR configuration was prepared. As with the critical experiments, BOL k-effective was high (3%). However, end-of-life k-effective was high (2%). The end-of-life concentrations of fission products were compared to those generated using the ORIGEN code. Agreement was generally good through differences in the inventories of some important nuclides, Xe and I, need to be understood. End-of-cycle curium target isotopics based on measured, discharged target rods were compared to calculated values and agreement was good. Axial flux plots at various irradiation positions were generated. Time-dependent power distributions based on two-dimensional calculations were provided.

Primm, R.T. III.

1992-03-01T23:59:59.000Z

94

A high-speed data acquisition system to measure low-level current from self-powered flux detectors in CANDU nuclear reactors  

E-Print Network [OSTI]

A high-speed data acquisition system to measure low-level current from self-powered flux detectors in CANDU nuclear reactors

Lawrence, C B

1982-01-01T23:59:59.000Z

95

Photo-neutralization of Negative Ion Beam for Future Fusion Reactor  

SciTech Connect (OSTI)

An exploratory study of negative ion beam photo-neutralization for future fusion reactors is explained. A refolded Fabry-Perot cavity system is proposed, with which a 60% neutralisation efficiency could be reached with low electric power consumption. The system would make use of sophisticated optical-cavity locking systems, which have been developed recently for gravitational-wave optical detection. The ITER Neutral beam Injector is taken as an example.

Chaibi, W.; Simonin, A. [CEA, IRFM, F-13108 St Paul lez Durance (France); Blondel, C.; Cabaret, L.; Delsart, C.; Drag, C. [Laboratoire Aime Cotton, CNRS-Universite Paris XI, Bat 505 Campus d'Orsay 91405 Orsay (France)

2009-03-12T23:59:59.000Z

96

The Intense Slow Positron Beam Facility at the NC State University PULSTAR Reactor  

SciTech Connect (OSTI)

An intense slow positron beam is in its early stages of operation at the 1-MW open-pool PULSTAR research reactor at North Carolina State University. The positron beam line is installed in a beam port that has a 30-cmx30-cm cross sectional view of the core. The positrons are created in a tungsten converter/moderator by pair-production using gamma rays produced in the reactor core and by neutron capture reactions in cadmium cladding surrounding the tungsten. Upon moderation, slow ({approx}3 eV) positrons that are emitted from the moderator are electrostatically extracted, focused and magnetically guided until they exit the reactor biological shield with 1-keV energy, approximately 3-cm beam diameter and an intensity exceeding 6x10{sup 8} positrons per second. A magnetic beam switch and transport system has been installed and tested that directs the beam into one of two spectrometers. The spectrometers are designed to implement state-of-the-art PALS and DBS techniques to perform positron and positronium annihilation studies of nanophases in matter.

Hawari, Ayman I.; Moxom, Jeremy; Hathaway, Alfred G.; Brown, Benjamin [Nuclear Engineering/Nuclear Reactor Program, North Carolina State University, P.O. Box 7909, Raleigh NC 27695 (United States); Gidley, David W.; Vallery, Richard [Physics Department, University of Michigan, 450 Church Street, Ann Arbor MI 48109 (United States); Xu, Jun [Chemical and Analytical Sciences Division, Oak Ridge National Laboratory, Oak Ridge TN 37831 (United States)

2009-03-10T23:59:59.000Z

97

Large break loss-of-coolant accident analyses for the high flux isotope reactor  

SciTech Connect (OSTI)

The US Department of Energy's High Flux Isotope Reactor (HFIR) was analyzed to evaluate it's response to a spectrum of loss-of-coolant accidents (LOCAs) with potential for leading to core damage. The MELCOR severe accident analysis code (version 1.7.1) was used to evaluate the overall dynamic response of HFIR. Before conducting LOCA analyses, the steady-state thermal-hydraulic parameters evaluated by MELCOR for various loop sections were verified against steady-state operating data. Thereafter, HFIR depressurization tests were simulated to evaluate the system pressure change for a given depletion in coolant inventory. Interesting and important safety-related phenomena were observed. The current analyses (which should be considered preliminary) that occur over a period from 1 to 3 seconds do not lead to core wide fuel melting. Core fluid flashing during the initial rapid depressurization does cause fuel temperature excursions due to adiabatic-like heatup. 3 refs., 4 figs.

Taleyarkhan, R.P. (Oak Ridge National Lab., TN (USA))

1989-01-01T23:59:59.000Z

98

Studies of Plutonium-238 Production at the High Flux Isotope Reactor  

SciTech Connect (OSTI)

The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) is a versatile 85 MW{sub th}, pressurized, light water-cooled and -moderated research reactor. The core consists of two fuel elements, an inner fuel element (IFE) and an outer fuel element (OFE), each constructed of involute fuel plates containing high-enriched-uranium (HEU) fuel ({approx}93 wt% {sup 235}U/U) in the form of U{sub 3}O{sub 8} in an Al matrix and encapsulated in Al-6061 clad. An over-moderated flux trap is located in the center of the core, a large beryllium reflector is located on the outside of the core, and two control elements (CE) are located between the fuel and the reflector. The flux trap and reflector house numerous experimental facilities which are used for isotope production, material irradiation, and cold/thermal neutron scattering. Over the past five decades, the US Department of Energy (DOE) and its agencies have been producing radioisotope power systems used by the National Aeronautics and Space Administration (NASA) for unmanned, long-term space exploration missions. Plutonium-238 is used to power Radioisotope Thermoelectric Generators (RTG) because it has a very long half-life (t{sub 1/2} {approx} 89 yr.) and it generates about 0.5 watts/gram when it decays via alpha emission. Due to the recent shortage and uncertainty of future production, the DOE has proposed a plan to the US Congress to produce {sup 238}Pu by irradiating {sup 237}Np as early as in fiscal year 2011. An annual production rate of 1.5 to 2.0 kg of {sup 238}Pu is expected to satisfy these needs and could be produced in existing national nuclear facilities like HFIR and the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Reactors at the Savannah River Site were used in the past for {sup 238}Pu production but were shut down after the last production in 1988. The nation's {sup 237}Np inventory is currently stored at INL. A plan for producing {sup 238}Pu at US research reactor facilities such as the High Flux Isotope Reactor at ORNL has been initiated by the US DOE and NASA for space exploration needs. Two Monte Carlo-based depletion codes, TRITON (ORNL) and VESTA (IRSN), were used to study the {sup 238}Pu production rates with varying target configurations in a typical HFIR fuel cycle. Preliminary studies have shown that approximately 11 grams and within 15 to 17 grams of {sup 238}Pu could be produced in the first irradiation cycle in one small and one large VXF facility, respectively, when irradiating fresh target arrays as those herein described. Important to note is that in this study we discovered that small differences in assumptions could affect the production rates of Pu-238 observed. The exact flux at a specific target location can have a significant impact upon production, so any differences in how the control elements are modeled as a function of exposure, will also cause differences in production rates. In fact, the surface plot of the large VXF target Pu-238 production shown in Figure 3 illustrates that the pins closest to the core can potentially have production rates as high as 3 times those of pins away from the core, thus implying that a cycle-to-cycle rotation of the targets may be well advised. A methodology for generating spatially-dependent, multi-group self-shielded cross sections and flux files with the KENO and CENTRM codes has been created so that standalone ORIGEN-S inputs can be quickly constructed to perform a variety of {sup 238}Pu production scenarios, i.e. combinations of the number of arrays loaded and the number of irradiation cycles. The studies herein shown with VESTA and TRITON/KENO will be used to benchmark the standalone ORIGEN.

Lastres, Oscar [University of Tennessee, Knoxville (UTK)] [University of Tennessee, Knoxville (UTK); Chandler, David [University of Tennessee, Knoxville (UTK) & Oak Ridge National Laboratory (ORNL)] [University of Tennessee, Knoxville (UTK) & Oak Ridge National Laboratory (ORNL); Jarrell, Joshua J [ORNL] [ORNL; Maldonado, G. Ivan [University of Tennessee, Knoxville (UTK)] [University of Tennessee, Knoxville (UTK)

2011-01-01T23:59:59.000Z

99

STATUS OF HIGH FLUX ISOTOPE REACTOR IRRADIATION OF SILICON CARBIDE/SILICON CARBIDE JOINTS  

SciTech Connect (OSTI)

Development of silicon carbide (SiC) joints that retain adequate structural and functional properties in the anticipated service conditions is a critical milestone toward establishment of advanced SiC composite technology for the accident-tolerant light water reactor (LWR) fuels and core structures. Neutron irradiation is among the most critical factors that define the harsh service condition of LWR fuel during the normal operation. The overarching goal of the present joining and irradiation studies is to establish technologies for joining SiC-based materials for use as the LWR fuel cladding. The purpose of this work is to fabricate SiC joint specimens, characterize those joints in an unirradiated condition, and prepare rabbit capsules for neutron irradiation study on the fabricated specimens in the High Flux Isotope Reactor (HFIR). Torsional shear test specimens of chemically vapor-deposited SiC were prepared by seven different joining methods either at Oak Ridge National Laboratory or by industrial partners. The joint test specimens were characterized for shear strength and microstructures in an unirradiated condition. Rabbit irradiation capsules were designed and fabricated for neutron irradiation of these joint specimens at an LWR-relevant temperature. These rabbit capsules, already started irradiation in HFIR, are scheduled to complete irradiation to an LWR-relevant dose level in early 2015.

Katoh, Yutai [ORNL; Koyanagi, Takaaki [ORNL; Kiggans, Jim [ORNL; Cetiner, Nesrin [ORNL; McDuffee, Joel [ORNL

2014-09-01T23:59:59.000Z

100

A system analysis computer model for the High Flux Isotope Reactor (HFIRSYS Version 1)  

SciTech Connect (OSTI)

A system transient analysis computer model (HFIRSYS) has been developed for analysis of small break loss of coolant accidents (LOCA) and operational transients. The computer model is based on the Advanced Continuous Simulation Language (ACSL) that produces the FORTRAN code automatically and that provides integration routines such as the Gear`s stiff algorithm as well as enabling users with numerous practical tools for generating Eigen values, and providing debug outputs and graphics capabilities, etc. The HFIRSYS computer code is structured in the form of the Modular Modeling System (MMS) code. Component modules from MMS and in-house developed modules were both used to configure HFIRSYS. A description of the High Flux Isotope Reactor, theoretical bases for the modeled components of the system, and the verification and validation efforts are reported. The computer model performs satisfactorily including cases in which effects of structural elasticity on the system pressure is significant; however, its capabilities are limited to single phase flow. Because of the modular structure, the new component models from the Modular Modeling System can easily be added to HFIRSYS for analyzing their effects on system`s behavior. The computer model is a versatile tool for studying various system transients. The intent of this report is not to be a users manual, but to provide theoretical bases and basic information about the computer model and the reactor.

Sozer, M.C.

1992-04-01T23:59:59.000Z

Note: This page contains sample records for the topic "flux beam reactor" 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

Density of Gadolinium Nitrate Solutions for the High Flux Isotope Reactor  

SciTech Connect (OSTI)

In late 1992, the High Flux Isotope Reactor (HFIR) was planning to switch the solution contained in the poison injection tank from cadmium nitrate to gadolinium nitrate. The poison injection system is an emergency system used to shut down the reactor by adding a neutron poison to the cooling water. This system must be able to supply a minimum of 69 pounds of gadolinium to the reactor coolant system in order to guarantee that the reactor would become subcritical. A graph of the density of gadolinium nitrate solutions over a concentration range of 5 to 30 wt% and a temperature range of 15 to 40{sup o}C was prepared. Routine density measurements of the solution in the poison injection tank are made by HFIR personnel, and an adaptation of the original graph is used to determine the gadolinium nitrate concentration. In late 2008, HFIR personnel decided that the heat tracing that was present on the piping for the poison injection system could be removed without any danger of freezing the solution; however, the gadolinium nitrate solution might get as cold as 5{sup o}C. This was outside the range of the current density-concentration correlation, so the range needed to be expanded. This report supplies a new density-concentration correlation that covers the extended temperature range. The correlation is given in new units, which greatly simplifies the calculation that is required to determine the pounds of gadolinium in the tank solution. The procedure for calculating the amount of gadolinium in the HFIR poison injection system is as follows: (1) Calculate the usable volume in the system; (2) Measure the density of the solution; (3) Calculate the gadolinium concentration using the following equation: Gd(lb/ft{sup 3}) = measured density (g/mL) x 34.681 - 34.785; (4) Calculate the amount of gadolinium in the system using the following equation: Amount of Gd(lb) = Gd concentration (lb/ft{sup 3}) x usable volume (ft{sup 3}). The equation in step 3 is exact for a temperature of 5{sup o}C, and overestimates the gadolinium concentration at all higher temperatures. This guarantees that the calculation is conservative, in that the actual concentration will be at least as high as that calculated. If an additional safety factor is desired, it is recommended that an administrative control limit be set that is higher than the required minimum amount of gadolinium.

Taylor, Paul Allen [ORNL; Lee, Denise L [ORNL

2009-05-01T23:59:59.000Z

102

Reactivity Accountability Attributed to Reflector Poisons in the High Flux Isotope Reactor  

SciTech Connect (OSTI)

The objective of this study is to develop a methodology to predict the reactivity impact as a function of outage time between cycles of 3He, 6Li, and other poisons in the High Flux Isotope Reactor s (HFIR) beryllium reflector. The reactivity worth at startup of the HFIR has been incorrectly predicted in the past after the reactor has been shut-down for long periods of time. The incorrect prediction was postulated to be due to the erroneous calculation of 3He buildup in the beryllium reflector. It is necessary to develop a better estimate of the start-of-cycle symmetric critical control element positions since if the estimated and actual symmetrical critical control element positions differ by more than $1.55 in reactivity (approximately one-half inch in control element startup position), HFIR is to be shutdown and a technical evaluation is performed to resolve the discrepancy prior to restart. 3He is generated and depleted during operation, but during an outage, the depletion of 3He ceases because it is a stable isotope. 3He is born from the radioactive decay of tritium, and thus the concentration of 3He increases during shutdown. SCALE, specifically the TRITON and CSAS5 control modules including the KENO V.A, COUPLE, and ORIGEN functional modules were utilized in this study. An equation relating the down time (td) to the change in symmetric control element position was generated and validated against measurements for approximately 40 HFIR operating cycles. The newly-derived correlation was shown to improve accuracy of predictions for long periods of down time.

Chandler, David [ORNL; Maldonado, G Ivan [ORNL; Primm, Trent [ORNL

2009-12-01T23:59:59.000Z

103

Investigation of failed fuel detection and location using a flux tilting method in a fast breeder reactor  

SciTech Connect (OSTI)

Detection and location of failed fuel in a liquid-metal fast breeder reactor (LMFBR) are very important both for safety and availability. When a fuel failure is detected, it is desirable to identify the failed subassembly quickly to reduce plant shutdown time. The flux tilting method is expected to effectively identify the defective subassembly. The feasibility of the flux tilting method is investigated for an LMFBR with a 100-MW (electric) homogeneous core. A numerical simulation is performed to estimate the viability of the flux tilting method, and a combination of the flux tilting method and the sipping method is found to be very effective in identifying the failed subassembly. In this paper a functional scheme for a computer-aided failed fuel detection and location system is discussed as part of a future on-line support system.

Hamada, M.; Uehara, K.; Muramatsu, K. (Japan Atomic Power Co., 1-6-1 Otemachi, Chiyoda-ku, Tokyo 100 (JP)); Kamei, T.; Tamaoki, T.; Yamaoka, M.; Sonada, Y.; Sano, Y. (Toshiba Corp., Nuclear Engineering Lab., 4-1 Ukishima-cho, Kawasaki-ku, Kawasaki 210 (JP))

1992-04-01T23:59:59.000Z

104

On RELAP5-simulated High Flux Isotope Reactor reactivity transients: Code change and application  

SciTech Connect (OSTI)

This paper presents a new and innovative application for the RELAP5 code (hereafter referred to as the code''). The code has been used to simulate several transients associated with the (presently) draft version of the High-Flux Isotope Reactor (HFIR) updated safety analysis report (SAR). This paper investigates those thermal-hydraulic transients induced by nuclear reactivity changes. A major goal of the work was to use an existing RELAP5 HFIR model for consistency with other thermal-hydraulic transient analyses of the SAR. To achieve this goal, it was necessary to incorporate a new self-contained point kinetics solver into the code because of a deficiency in the point-kinetics reactivity model of the Mod 2.5 version of the code. The model was benchmarked against previously analyzed (known) transients. Given this new code, four event categories defined by the HFIR probabilistic risk assessment (PRA) were analyzed: (in ascending order of severity) a cold-loop pump start; run-away shim-regulating control cylinder and safety plate withdrawal; control cylinder ejection; and generation of an optimum void in the target region. All transients are discussed. Results of the bounding incredible event transient, the target region optimum void, are shown. Future plans for RELAP5 HFIR applications and recommendations for code improvements are also discussed.

Freels, J.D.

1993-01-01T23:59:59.000Z

105

Extraction of gadolinium from high flux isotope reactor control plates. [Alternative method  

SciTech Connect (OSTI)

Gadolinium-153 is an important radioisotope used in the diagnosis of various bone disorders. Recent medical and technical developments in the detection and cure of osteoporosis, a bone disease affecting an estimated 50 million people, have greatly increased the demand for this isotope. The Oak Ridge National Laboratory (ORNL) has produced /sup 153/Gd since 1980 primarily through the irradiation of a natural europium-oxide powder followed by the chemical separation of the gadolinium fraction from the europium material. Due to the higher demand for /sup 153/Gd, an alternative production method to supplement this process has been investigated. This process involves the extraction of gadolinium from the europium-bearing region of highly radioactive, spent control plates used at the High Flux Isotope Reactor (HFIR) with a subsequent re-irradiation of the extracted material for the production of the /sup 153/Gd. Based on the results of experimental and calculational analyses, up to 25 grams of valuable gadolinium (greater than or equal to60% enriched in /sup 152/Gd) resides in the europium-bearing region of the HFIR control components of which 70% is recoverable. At a specific activity yield of 40 curies of /sup 153/Gd for each gram of gadolinium re-irradiated, 700 one-curie sources can be produced from each control plate assayed.

Kohring, M.W.

1987-04-01T23:59:59.000Z

106

On RELAP5-simulated High Flux Isotope Reactor reactivity transients: Code change and application  

SciTech Connect (OSTI)

This paper presents a new and innovative application for the RELAP5 code (hereafter referred to as ``the code``). The code has been used to simulate several transients associated with the (presently) draft version of the High-Flux Isotope Reactor (HFIR) updated safety analysis report (SAR). This paper investigates those thermal-hydraulic transients induced by nuclear reactivity changes. A major goal of the work was to use an existing RELAP5 HFIR model for consistency with other thermal-hydraulic transient analyses of the SAR. To achieve this goal, it was necessary to incorporate a new self-contained point kinetics solver into the code because of a deficiency in the point-kinetics reactivity model of the Mod 2.5 version of the code. The model was benchmarked against previously analyzed (known) transients. Given this new code, four event categories defined by the HFIR probabilistic risk assessment (PRA) were analyzed: (in ascending order of severity) a cold-loop pump start; run-away shim-regulating control cylinder and safety plate withdrawal; control cylinder ejection; and generation of an optimum void in the target region. All transients are discussed. Results of the bounding incredible event transient, the target region optimum void, are shown. Future plans for RELAP5 HFIR applications and recommendations for code improvements are also discussed.

Freels, J.D.

1993-07-01T23:59:59.000Z

107

Neutron flux estimations based on niobium impurities in reactor pressure vessel steel  

SciTech Connect (OSTI)

The use of (ppm level) niobium impurities in reactor pressure vessel (RPV) steel for neutron flux estimations based on the reaction {sup 93}Nb (n,n{prime}) {sup 93m}Nb has been reported previously. The method has now been further investigated and refined. Small niobium fractions in RPV steel ({approx} ppm) and plating ({approx} 1%) materials have been separated by ion exchange chromatography in one to three steps. The measured Nb fractions in samples from some four pressure vessel (RPV) base materials were 1 to 3 ppm. The purification of tens of milligrams of RPV material provides sufficient amounts of niobium for mass determination with a highly sensitive (10{sup {minus}5} ppm) Inductively Coupled Plasma Mass Spectrometer (ICP-MS). The {sup 93m}Nb and small remaining {sup 54}Mn activities were measured with a Calibrated Liquid Scintillation Counter (LSC) based on dual label technique and almost 100% efficiency to {sup 93m}Nb. One purification is needed for plating materials ({approx}1% Nb) and two purifications of about one gram of steel with Nb impurities in order to resolve the needed activities ({approx}10 Bq {sup 93m}Nb/{mu}g Nb). The achieved accuracy of the measured specific {sup 93m}Nb activities was about {+-} 3% (1{sigma}) in irradiated RPV plating materials and about {+-} 4% for Nb ppm impurities.

Baers, L.B.; Hasanen, E.K. [Technical Research Centre of Finland, Espoo (Finland). Reactor Lab.

1994-12-31T23:59:59.000Z

108

Hydrogen Cylinder Storage Array Explosion Evaluations at the High Flux Isotope Reactor  

SciTech Connect (OSTI)

The safety analysis for a recently-installed cold neutron source at the High Flux Isotope Reactor (HFIR) involved evaluation of potential explosion consequences from accidental hydrogen jet releases that could occur from an array of hydrogen cylinders. The scope of the safety analysis involved determination of the release rate of hydrogen, the total quantity of hydrogen assumed to be involved in the explosion, the location of an ignition point or center of the explosion from receptors of interest, and the peak overpressure at the receptors. To evaluate the total quantity of hydrogen involved in the explosion, a 2D model was constructed of the jet concentration and a radial-axial integral over the jet cloud from the centerline to the flammability limit of 4% was used to determine the hydrogen mass to be used as a source term. The location of the point source was chosen as the peak of the jet centerline concentration profile. Consequences were assessed using a combination of three methods for estimating local overpressure as a function of explosion source strength and distance: the Baker-Strehlow method, the TNT-equivalence method, and the TNO method. Results from the explosions were assessed using damage estimates in screening tables for buildings and industrial equipment.

Cook, David Howard [ORNL] [ORNL; Griffin, Frederick P [ORNL] [ORNL; Hyman III, Clifton R [ORNL] [ORNL

2010-01-01T23:59:59.000Z

109

Tritium trapping in silicon carbide in contact with solid breeder under high flux isotope reactor irradiation  

SciTech Connect (OSTI)

The trapping of tritium in silicon carbide (SiC) injected from ceramic breeding materials was examined via tritium measurements using imaging plate (IP) techniques. Monolithic SiC in contact with ternary lithium oxide (lithium titanate and lithium aluminate) as a ceramic breeder was irradiated in the High Flux Isotope Reactor (HFIR) in Oak Ridge, Tennessee, USA. The distribution of photo-stimulated luminescence (PSL) of tritium in SiC was successfully obtained, which separated the contribution of 14C ß-rays to the PSL. The tritium incident from ceramic breeders was retained in the vicinity of the SiC surface even after irradiation at 1073 K over the duration of ~3000 h, while trapping of tritium was not observed in the bulk region. The PSL intensity near the SiC surface in contact with lithium titanate was higher than that obtained with lithium aluminate. The amount of the incident tritium and/or the formation of a Li2SiO3 phase on SiC due to the reaction with lithium aluminate under irradiation likely were responsible for this observation.

H. Katsui; Y. Katoh; A. Hasegawa; M. Shimada; Y. Hatano; T. Hinoki; S. Nogami; T. Tanaka; S. Nagata; T. Shikama

2013-11-01T23:59:59.000Z

110

Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual report for FY 2009  

SciTech Connect (OSTI)

This report documents progress made during FY 2009 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Studies are reported of the application of a silicon coating to surrogates for spheres of uranium-molybdenum alloy. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. A description of the progress in developing a finite element thermal hydraulics model of the LEU core is provided.

Chandler, David [ORNL; Freels, James D [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Primm, Trent [ORNL; Sease, John D [ORNL; Guida, Tracey [University of Pittsburgh; Jolly, Brian C [ORNL

2010-02-01T23:59:59.000Z

111

Inverse Beta Decay in a Nonequilibrium Antineutrino Flux from a Nuclear Reactor  

E-Print Network [OSTI]

The evolution of the reactor antineutrino spectrum toward equilibrium above the inverse beta-decay threshold during the reactor operating period and the decay of residual antineutrino radiation after reactor shutdown are considered. It is found that, under certain conditions, these processes can play a significant role in experiments seeking neutrino oscillations.

V. I. Kopeikin; L. A. Mikaelyan; V. V. Sinev

2001-10-23T23:59:59.000Z

112

Unsteady momentum fluxes in two-phase flow and the vibration of nuclear reactor components  

E-Print Network [OSTI]

The steady and unsteady components of the momentum flux in a twophase flow have been measured at the exit of a vertical pipe. Measured momentum flux data has been machine processed by standard random vibration techniques ...

Yih, Tien Sieh

1968-01-01T23:59:59.000Z

113

The use of PRA (Probabilistic Risk Assessment) in the management of safety issues at the High Flux Isotope Reactor  

SciTech Connect (OSTI)

The High Flux Isotope reactor (HFIR) is a high performance isotope production and research reactor which has been in operation at Oak Ridge National Laboratory (ORNL) since 1965. In late 1986 the reactor was shut down as a result of discovery of unexpected neutron embrittlement of the reactor vessel. In January of 1988, a level 1 Probabilistic Risk Assessment (PRA) (excluding external events) was published as part of the response to the many reviews that followed the shutdown and for use by ORNL to prioritize action items intended to upgrade the safety of the reactor. A conservative estimate of the core damage frequency initiated by internal events for HFIR was 3.11 {times} 10{sup {minus}4}. In June 1989 a draft external events initiated PRA was published. The dominant contributions from external events came from seismic, wind, and fires. The overall external event contribution to core damage frequency is about 138% of the internal event initiated contribution and is dominated by wind initiators. The PRA has provided a basis for the management of a wide range of safety and operation issues at the HFIR. 3 refs., 4 figs., 2 tabs.

Flanagan, G.F.

1990-01-01T23:59:59.000Z

114

ORNL - Restart of the High Flux Isotope Reactor 2-07  

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

personnel are provided, and adequate facilities and equipment are available to ensure engineering services are adequate to support reactor and CS operations. The engineering...

115

ORNL - Restart of the High Flux Isotope Reactor 2-07  

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

and adequate facilities and equipment are available to ensure services are adequate to conduct and support reactor operations with the hydrogen-moderated CS. Functions,...

116

ORNL - Restart of the High Flux Isotope Reactor 2-07  

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

reactor operations and CS system operations staff and management are available to conduct and support safe operations with the hydrogen-moderated CS. (CR - 1, CR - 4, CR -...

117

Radiobiologic Significance of Response of Intratumor Quiescent Cells In Vivo to Accelerated Carbon Ion Beams Compared With {gamma}-Rays and Reactor Neutron Beams  

SciTech Connect (OSTI)

Purpose: To clarify the radiosensitivity of intratumor quiescent cells in vivo to accelerated carbon ion beams and reactor neutron beams. Methods and Materials: Squamous cell carcinoma VII tumor-bearing mice were continuously given 5-bromo-2'-deoxyuridine to label all intratumor proliferating cells. Next, they received accelerated carbon ion or {gamma}-ray high-dose-rate (HDR) or reduced-dose-rate (RDR) irradiation. Other tumor-bearing mice received reactor thermal or epithermal neutrons with RDR irradiation. Immediately after HDR and RDR irradiation or 12 h after HDR irradiation, the response of quiescent cells was assessed in terms of the micronucleus frequency using immunofluorescence staining for 5-bromo-2'-deoxyuridine. The response of the total (proliferating plus quiescent) tumor cells was determined from the 5-bromo-2'-deoxyuridine nontreated tumors. Results: The difference in radiosensitivity between the total and quiescent cell populations after {gamma}-ray irradiation was markedly reduced with reactor neutron beams or accelerated carbon ion beams, especially with a greater linear energy transfer (LET) value. Clearer repair in quiescent cells than in total cells through delayed assay or a decrease in the dose rate with {gamma}-ray irradiation was efficiently inhibited with carbon ion beams, especially with a greater LET. With RDR irradiation, the radiosensitivity to accelerated carbon ion beams with a greater LET was almost similar to that to reactor thermal and epithermal neutron beams. Conclusion: In terms of tumor cell-killing effect as a whole, including quiescent cells, accelerated carbon ion beams, especially with greater LET values, are very useful for suppressing the dependency on the heterogeneity within solid tumors, as well as depositing the radiation dose precisely.

Masunaga, Shin-ichiro [Particle Radiation Oncology Research Center, Research Reactor Institute, Kyoto University, Kumatori, Osaka (Japan)], E-mail: smasuna@rri.kyoto-u.ac.jp; Ando, Koichi; Uzawa, Akiko; Hirayama, Ryoichi; Furusawa, Yoshiya; Koike, Sachiko [Heavy-Ion Radiobiology Research Group, Research Center for Charged Particle Therapy, National Institute of Radiological Sciences, Chiba (Japan); Sakurai, Yoshinori [Department of Physics, Faculty of Medicine, Sapporo Medical University, Sapporo (Japan); Nagata, Kenji; Suzuki, Minoru; Kashino, Genro [Particle Radiation Oncology Research Center, Research Reactor Institute, Kyoto University, Kumatori, Osaka (Japan); Kinashi, Yuko [Department of Radiation Safety and Control, Research Reactor Institute, Kyoto University, Kumatori, Osaka (Japan); Tanaka, Hiroki; Maruhashi, Akira [Department of Radiation Medical Physics, Research Reactor Institute, Kyoto University, Kumatori, Osaka (Japan); Ono, Koji [Particle Radiation Oncology Research Center, Research Reactor Institute, Kyoto University, Kumatori, Osaka (Japan)

2008-01-01T23:59:59.000Z

118

DESIGN STUDY FOR A LOW-ENRICHED URANIUM CORE FOR THE HIGH FLUX ISOTOPE REACTOR, ANNUAL REPORT FOR FY 2010  

SciTech Connect (OSTI)

This report documents progress made during FY 2010 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current level. Studies are reported of support to a thermal hydraulic test loop design, the implementation of finite element, thermal hydraulic analysis capability, and infrastructure tasks at HFIR to upgrade the facility for operation at 100 MW. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. Continuing development in the definition of the fuel fabrication process is described.

Cook, David Howard [ORNL; Freels, James D [ORNL; Ilas, Germina [ORNL; Jolly, Brian C [ORNL; Miller, James Henry [ORNL; Primm, Trent [ORNL; Renfro, David G [ORNL; Sease, John D [ORNL; Pinkston, Daniel [ORNL

2011-02-01T23:59:59.000Z

119

Establishing a Cost Basis for Converting the High Flux Isotope Reactor from High Enriched to Low Enriched Uranium Fuel  

SciTech Connect (OSTI)

Under the auspices of the Global Threat Reduction Initiative Reduced Enrichment for Research and Test Reactors Program, the National Nuclear Security Administration /Department of Energy (NNSA/DOE) has, as a goal, to convert research reactors worldwide from weapons grade to non-weapons grade uranium. The High Flux Isotope Reactor (HFIR) at Oak Ridge National Lab (ORNL) is one of the candidates for conversion of fuel from high enriched uranium (HEU) to low enriched uranium (LEU). A well documented business model, including tasks, costs, and schedules was developed to plan the conversion of HFIR. Using Microsoft Project, a detailed outline of the conversion program was established and consists of LEU fuel design activities, a fresh fuel shipping cask, improvements to the HFIR reactor building, and spent fuel operations. Current-value costs total $76 million dollars, include over 100 subtasks, and will take over 10 years to complete. The model and schedule follows the path of the fuel from receipt from fuel fabricator to delivery to spent fuel storage and illustrates the duration, start, and completion dates of each subtask to be completed. Assumptions that form the basis of the cost estimate have significant impact on cost and schedule.

Primm, Trent [ORNL; Guida, Tracey [University of Pittsburgh

2010-02-01T23:59:59.000Z

120

Design Study for a Low-enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2007  

SciTech Connect (OSTI)

This report documents progress made during fiscal year 2007 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low enriched uranium fuel (LEU). Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. A high volume fraction U/Mo-in-Al fuel could attain the same neutron flux performance as with the current, HEU fuel but materials considerations appear to preclude production and irradiation of such a fuel. A diffusion barrier would be required if Al is to be retained as the interstitial medium and the additional volume required for this barrier would degrade performance. Attaining the high volume fraction (55 wt. %) of U/Mo assumed in the computational study while maintaining the current fuel plate acceptance level at the fuel manufacturer is unlikely, i.e. no increase in the percentage of plates rejected for non-compliance with the fuel specification. Substitution of a zirconium alloy for Al would significantly increase the weight of the fuel element, the cost of the fuel element, and introduce an as-yet untried manufacturing process. A monolithic U-10Mo foil is the choice of LEU fuel for HFIR. Preliminary calculations indicate that with a modest increase in reactor power, the flux performance of the reactor can be maintained at the current level. A linearly-graded, radial fuel thickness profile is preferred to the arched profile currently used in HEU fuel because the LEU fuel media is a metal alloy foil rather than a powder. Developments in analysis capability and nuclear data processing techniques are underway with the goal of verifying the preliminary calculations of LEU flux performance. A conceptual study of the operational cost of an LEU fuel fabrication facility yielded the conclusion that the annual fuel cost to the HFIR would increase significantly from the current, HEU fuel cycle. Though manufacturing can be accomplished with existing technology, several engineering proof-of-principle tests would be required. The RERTR program is currently conducting a series of generic fuel qualification tests at the Advanced Test Reactor. A review of these tests and a review of the safety basis for the current, HEU fuel cycle led to the identification of a set of HFIR-specific fuel qualification tests. Much additional study is required to formulate a HFIR-specific fuel qualification plan from this set. However, one such test - creating a graded fuel profile across a flat foil - has been initiated with promising results.

Primm, Trent [ORNL; Ellis, Ronald James [ORNL; Gehin, Jess C [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Sease, John D [ORNL

2007-11-01T23:59:59.000Z

Note: This page contains sample records for the topic "flux beam reactor" from the National Library of EnergyBeta (NLEBeta).
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121

Self-corrected Sensors Based On Atomic Absorption Spectroscopy For Atom Flux Measurements In Molecular Beam Epitaxy  

SciTech Connect (OSTI)

A high sensitivity atom flux sensor based on atomic absorption spectroscopy has been designed and implemented to control electron beam evaporators and effusion cells in a molecular beam epitaxy system. Using a high-resolution spectrometer and a two-dimensional charge coupled device (CCD) detector in a double-beam configuration, we employ a non-resonant line or a resonant line with lower absorbance from the same hollow cathode lamp as the reference for nearly perfect background correction and baseline drift removal. This setup also significantly shortens the warm-up time needed compared to other sensor technologies and drastically reduces the noise coming from the surrounding environment. In addition, the high-resolution spectrometer allows the most sensitive resonant line to be isolated and used to provide excellent signal-to-noise ratio.

Du, Yingge; Droubay, Timothy C.; Liyu, Andrey V.; Li, Guosheng; Chambers, Scott A.

2014-04-24T23:59:59.000Z

122

Neutronic Analysis of an Advanced Fuel Design Concept for the High Flux Isotope Reactor  

SciTech Connect (OSTI)

This study presents the neutronic analysis of an advanced fuel design concept for the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) that could significantly extend the current fuel cycle length under the existing design and safety criteria. A key advantage of the fuel design herein proposed is that it would not require structural changes to the present HFIR core, in other words, maintaining the same rated power and fuel geometry (i.e., fuel plate thickness and coolant channel dimensions). Of particular practical importance, as well, is the fact that the proposed change could be justified within the bounds of the existing nuclear safety basis. The simulations herein reported employed transport theory-based and exposure-dependent eigenvalue characterization to help improve the prediction of key fuel cycle parameters. These parameters were estimated by coupling a benchmarked three-dimensional MCNP5 model of the HFIR core to the depletion code ORIGEN via the MONTEBURNS interface. The design of an advanced HFIR core with an improved fuel loading is an idea that evolved from early studies by R. D. Cheverton, formerly of ORNL. This study contrasts a modified and increased core loading of 12 kg of 235U against the current core loading of 9.4 kg. The simulations performed predict a cycle length of 39 days for the proposed fuel design, which represents a 50% increase in the cycle length in response to a 25% increase in fissile loading, with an average fuel burnup increase of {approx}23%. The results suggest that the excess reactivity can be controlled with the present design and arrangement of control elements throughout the core's life. Also, the new power distribution is comparable or even improved relative to the current power distribution, displaying lower peak to average fission rate densities across the inner fuel element's centerline and bottom cells. In fact, the fission rate density in the outer fuel element also decreased at these key locations for the proposed design. Overall, it is estimated that the advanced core design could increase the availability of the HFIR facility by {approx}50% and generate {approx}33% more neutrons annually, which is expected to yield sizeable savings during the remaining life of HFIR, currently expected to operate through 2014. This study emphasizes the neutronics evaluation of a new fuel design. Although a number of other performance parameters of the proposed design check favorably against the current design, and most of the core design features remain identical to the reference, it is acknowledged that additional evaluations would be required to fully justify the thermal-hydraulic and thermal-mechanical performance of a new fuel design, including checks for cladding corrosion performance as well as for industrial and economic feasibility.

Xoubi, Ned [ORNL; Primm, Trent [ORNL; Maldonado, G. Ivan [University of Tennessee, Knoxville (UTK)

2009-01-01T23:59:59.000Z

123

Low-Enriched Uranium Fuel Conversion Activities for the High Flux Isotope Reactor, Annual Report for FY 2011  

SciTech Connect (OSTI)

This report describes progress made during FY11 in ORNL activities to support converting the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum (UMo) alloy. With both radial and axial contouring of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current levels achieved with HEU fuel. Studies are continuing to demonstrate that the fuel thermal safety margins can be preserved following conversion. Studies are also continuing to update other aspects of the reactor steady state operation and accident response for the effects of fuel conversion. Technical input has been provided to Oregon State University in support of their hydraulic testing program. The HFIR conversion schedule was revised and provided to the GTRI program. In addition to HFIR conversion activities, technical support was provided directly to the Fuel Fabrication Capability program manager.

Renfro, David G [ORNL; Cook, David Howard [ORNL; Freels, James D [ORNL; Griffin, Frederick P [ORNL; Ilas, Germina [ORNL; Sease, John D [ORNL; Chandler, David [ORNL

2012-03-01T23:59:59.000Z

124

Modified floor response spectra for the Brookhaven National Laboratory High Flux Beam Reactor (HFBR)  

SciTech Connect (OSTI)

This report documents the modified floor response spectra that will be used for future seismic evaluations of structures, systems, and components (SSC`s) within the HFBR and the technical basis for the modified floor response spectra. When used within this report, the term ``current spectra`` refers to the 1978 spectra developed in BNL Informal Report BNL-26019.

Morante, R.J.; Skonieczny, J.

1994-03-01T23:59:59.000Z

125

Compact and high-particle-flux thermal-lithium-beam probe system for measurement of two-dimensional electron density profile  

SciTech Connect (OSTI)

A compact and high-particle-flux thermal-lithium-beam source for two-dimensional measurement of electron density profiles has been developed. The thermal-lithium-beam oven is heated by a carbon heater. In this system, the maximum particle flux of the thermal lithium beam was ?4 × 10{sup 19} m{sup ?2} s{sup ?1} when the temperature of the thermal-lithium-beam oven was 900 K. The electron density profile was evaluated in the small tokamak device HYBTOK-II. The electron density profile was reconstructed using the thermal-lithium-beam probe data and this profile was consistent with the electron density profile measured with a Langmuir electrostatic probe. We confirm that the developed thermal-lithium-beam probe can be used to measure the two-dimensional electron density profile with high time and spatial resolutions.

Shibata, Y., E-mail: shibata.yoshihide@jaea.go.jp; Manabe, T.; Ohno, N.; Takagi, M. [Graduate School of Engineering, Nagoya University, Furo-cho, Chikusa-ku, Nagoya, Aichi 464-8603 (Japan); Kajita, S. [EcoTopia Science Institute, Nagoya University, Furo-cho, Chikusa-ku, Nagoya, Aichi 464-8603 (Japan); Tsuchiya, H.; Morisaki, T. [National Institute for Fusion Science, Oroshi, Toki, Gifu 509-5292 (Japan)

2014-09-15T23:59:59.000Z

126

Determination of the theoretical feasibility for the transmutation of europium isotopes from high flux isotope reactor control cylinders  

SciTech Connect (OSTI)

The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) is a 100 MWth light-water research reactor designed and built in the 1960s primarily for the production of transuranic isotopes. The HFIR is equipped with two concentric cylindrical blade assemblies, known as control cylinders, that are used to control reactor power. These control cylinders, which become highly radioactive from neutron exposure, are periodically replaced as part of the normal operation of the reactor. The highly radioactive region of the control cylinders is composed of europium oxide in an aluminum matrix. The spent HFIR control cylinders have historically been emplaced in the ORNL Waste Area Grouping (WAG) 6. The control cylinders pose a potential radiological hazard due to the long lived radiotoxic europium isotopes {sup 152}Eu, {sup 154}Eu, and {sup 155}Eu. In a 1991 health evaluation of WAG 6 (ERD 1991) it was shown that these cylinders were a major component of the total radioactivity in WAG 6 and posed a potential exposure hazard to the public in some of the postulated assessment scenarios. These health evaluations, though preliminary and conservative in nature, illustrate the incentive to investigate methods for permanent destruction of the europium radionuclides. When the cost of removing the control cylinders from WAG 6, performing chemical separations and irradiating the material in HFIR are factored in, the option of leaving the control cylinders in place for decay must be considered. Other options, such as construction of an engineered barrier around the disposal silos to reduce the chance of migration, should also be analyzed.

Elam, K.R.; Reich, W.J.

1995-09-01T23:59:59.000Z

127

Report of the ANS Project Feasibility Workshop for a High Flux Isotope Reactor-Center for Neutron Research Facility  

SciTech Connect (OSTI)

The Advanced Neutron Source (ANS) Conceptual Design Report (CDR) and its subsequent updates provided definitive design, cost, and schedule estimates for the entire ANS Project. A recent update to this estimate of the total project cost for this facility was $2.9 billion, as specified in the FY 1996 Congressional data sheet, reflecting a line-item start in FY 1995. In December 1994, ANS management decided to prepare a significantly lower-cost option for a research facility based on ANS which could be considered during FY 1997 budget deliberations if DOE or Congressional planners wished. A cost reduction for ANS of about $1 billion was desired for this new option. It was decided that such a cost reduction could be achieved only by a significant reduction in the ANS research scope and by maximum, cost-effective use of existing High Flux Isotope Reactor (HFIR) and ORNL facilities to minimize the need for new buildings. However, two central missions of the ANS -- neutron scattering research and isotope production-were to be retained. The title selected for this new option was High Flux Isotope Reactor-Center for Neutron Research (HFIR-CNR) because of the project`s maximum use of existing HFIR facilities and retention of selected, central ANS missions. Assuming this shared-facility requirement would necessitate construction work near HFIR, it was specified that HFIR-CNR construction should not disrupt normal operation of HFIR. Additional objectives of the study were that it be highly credible and that any material that might be needed for US Department of Energy (DOE) and Congressional deliberations be produced quickly using minimum project resources. This requirement made it necessary to rely heavily on the ANS design, cost, and schedule baselines. A workshop methodology was selected because assessment of each cost and/or scope-reduction idea required nearly continuous communication among project personnel to ensure that all ramifications of propsed changes.

Peretz, F.J.; Booth, R.S. [comp.

1995-07-01T23:59:59.000Z

128

Fission-chamber-compensated self-powered detector for in-core flux measurement and reactor control  

SciTech Connect (OSTI)

An apparatus is described for in-core flux measurement and nuclear reactor control consisting of: a self-powered rhodium neutron detector for producing an output signal corresponding to reactor power level; first amplifier means having an input for receiving the neutron detector output signal, and the first amplifier means producing a corresponding first amplifier output signal (V/sub 1/); a fission chamber for producing an output signal corresponding to reactor power level; second amplifier means having an input for receiving the fission chamber output signal, the second amplifier means producing a corresponding second amplifier output signal; first differentiating means coupled to the output of the second amplifier means including a resistor and a capacitor coupled to provide a time constant T/sub 1//1n 2 where T/sub 1/ is the half-life of the ground state of Rh/sup 104/; third amplifier means having an input for receiving the second amplifier means output signal, the third amplifier means having a voltage gain of (1-c)/c; second differentiating means coupled to the output of the third amplifying means and including a capacitor and resistor providing a time constant T/sub 2//1n 2 where T/sub 2/ is the half-life of the isomeric state of Rh/sup 104/; and fourth amplifier means including corresponding inputs for receiving the output of the first amplifier means, a processed output from the first differentiating circuit, and a processed output from the second differentiating circuit, and for producing a corresponding summed output signal.

Neissel, J.P.

1986-04-15T23:59:59.000Z

129

Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility  

SciTech Connect (OSTI)

Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility.

Not Available

1992-07-01T23:59:59.000Z

130

Fabrication procedures for manufacturing High Flux Isotope Reactor fuel elements - 2  

SciTech Connect (OSTI)

The original fabrication procedures written in 1968 delineated the manufacturing procedures at that time. Since 1968, there have been a number of procedural changes. This rewrite of the fabrication procedures incorporates these changes. The entire fuel core of this reactor is made up of two fuel elements. Each element consists of one annular array of fuel plates. These annuli are identified as the inner and outer fuel elements, since one fits inside the other. The inner element consists of 171 identical fuel plates, and the outer element contains 369 identical fuel plates differing slightly from those in the inner element. Both sets of fuel plates contain U{sub 3}O{sub 8} powder as the fuel, dispersed in an aluminum powder matrix and clad with aluminum. Procedures for manufacturing and inspection of the fuel elements are described and illustrated.

Knight, R.W.; Morin, R.A.

1999-12-01T23:59:59.000Z

131

Development of CFD models to support LEU Conversion of ORNL s High Flux Isotope Reactor  

SciTech Connect (OSTI)

The US Department of Energy s National Nuclear Security Administration (NNSA) is participating in the Global Threat Reduction Initiative to reduce and protect vulnerable nuclear and radiological materials located at civilian sites worldwide. As an integral part of one of NNSA s subprograms, Reduced Enrichment for Research and Test Reactors, HFIR is being converted from the present HEU core to a low enriched uranium (LEU) core with less than 20% of U-235 by weight. Because of HFIR s importance for condensed matter research in the United States, its conversion to a high-density, U-Mo-based, LEU fuel should not significantly impact its existing performance. Furthermore, cost and availability considerations suggest making only minimal changes to the overall HFIR facility. Therefore, the goal of this conversion program is only to substitute LEU for the fuel type in the existing fuel plate design, retaining the same number of fuel plates, with the same physical dimensions, as in the current HFIR HEU core. Because LEU-specific testing and experiments will be limited, COMSOL Multiphysics was chosen to provide the needed simulation capability to validate against the HEU design data and previous calculations, and predict the performance of the proposed LEU fuel for design and safety analyses. To achieve it, advanced COMSOL-based multiphysics simulations, including computational fluid dynamics (CFD), are being developed to capture the turbulent flows and associated heat transfer in fine detail and to improve predictive accuracy [2].

Khane, Vaibhav B [ORNL] [ORNL; Jain, Prashant K [ORNL] [ORNL; Freels, James D [ORNL] [ORNL

2012-01-01T23:59:59.000Z

132

STARTUP REACTIVITY ACCOUNTABILITY ATTRIBUTED TO ISOTOPIC TRANSMUTATIONS IN THE IRRADIATED BERYLLIUM REFLECTOR OF THE HIGH FLUX ISTOTOPE REACTOR  

SciTech Connect (OSTI)

The objective of this study is to develop a methodology to predict the reactivity impact as a function of outage time between cycles of 3He, 6Li, and other poisons in the High Flux Isotope Reactor s (HFIR) beryllium reflector. The reactivity worth at startup of the HFIR has been incorrectly predicted in the past after the reactor has been shut-down for long periods of time. The incorrect prediction was postulated to be due to the erroneous calculation of 3He buildup in the beryllium reflector. It is necessary to develop a better estimate of the start-of-cycle symmetric critical control element positions since if the estimated and actual symmetrical critical control element positions differ by more than $1.55 in reactivity (approximately one-half inch in control element startup position), HFIR is to be shutdown and a technical evaluation is performed to resolve the discrepancy prior to restart. 3He is generated and depleted during operation, but during an outage, the depletion of 3He ceases because it is a stable isotope. 3He is born from the radioactive decay of tritium, and thus the concentration of 3He increases during shutdown. The computer program SCALE, specifically the TRITON and CSAS5 control modules including the KENO V.A, COUPLE, and ORIGEN functional modules were utilized in this study. An equation relating the down time (td) to the change in symmetric control element position was generated and validated against measurements for approximately 40 HFIR operating cycles. The newly-derived correlation was shown to improve accuracy of predictions for long periods of down time.

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

2010-01-01T23:59:59.000Z

133

Validation of a Monte Carlo based depletion methodology via High Flux Isotope Reactor HEU post-irradiation examination measurements  

SciTech Connect (OSTI)

The purpose of this study is to validate a Monte Carlo based depletion methodology by comparing calculated post-irradiation uranium isotopic compositions in the fuel elements of the High Flux Isotope Reactor (HFIR) core to values measured using uranium mass-spectrographic analysis. Three fuel plates were analyzed: two from the outer fuel element (OFE) and one from the inner fuel element (IFE). Fuel plates O-111-8, O-350-1, and I-417-24 from outer fuel elements 5-O and 21-O and inner fuel element 49-I, respectively, were selected for examination. Fuel elements 5-O, 21-O, and 49-1 were loaded into HFIR during cycles 4, 16, and 35, respectively (mid to late 1960s). Approximately one year after each of these elements were irradiated, they were transferred to the High Radiation Level Examination Laboratory (HRLEL) where samples from these fuel plates were sectioned and examined via uranium mass-spectrographic analysis. The isotopic composition of each of the samples was used to determine the atomic percent of the uranium isotopes. A Monte Carlo based depletion computer program, ALEPH, which couples the MCNP and ORIGEN codes, was utilized to calculate the nuclide inventory at the end-of-cycle (EOC). A current ALEPH/MCNP input for HFIR fuel cycle 400 was modified to replicate cycles 4, 16, and 35. The control element withdrawal curves and flux trap loadings were revised, as well as the radial zone boundaries and nuclide concentrations in the MCNP model. The calculated EOC uranium isotopic compositions for the analyzed plates were found to be in good agreement with measurements, which reveals that ALEPH/MCNP can accurately calculate burn-up dependent uranium isotopic concentrations for the HFIR core. The spatial power distribution in HFIR changes significantly as irradiation time increases due to control element movement. Accurate calculation of the end-of-life uranium isotopic inventory is a good indicator that the power distribution variation as a function of space and time is accurately calculated, i.e. an integral check. Hence, the time dependent heat generation source terms needed for reactor core thermal hydraulic analysis, if derived from this methodology, have been shown to be accurate for highly enriched uranium (HEU) fuel.

Chandler, David [ORNL; Maldonado, G Ivan [ORNL; Primm, Trent [ORNL

2010-01-01T23:59:59.000Z

134

Multi-scale Control and Enhancement of Reactor Boiling Heat Flux by Reagents and Nanoparticles  

SciTech Connect (OSTI)

The phenomenological characterization of the use of non-invasive and passive techniques to enhance the boiling heat transfer in water has been carried out in this extended study. It provides fundamental enhanced heat transfer data for nucleate boiling and discusses the associated physics with the aim of addressing future and next-generation reactor thermal-hydraulic management. It essentially addresses the hypothesis that in phase-change processes during boiling, the primary mechanisms can be related to the liquid-vapor interfacial tension and surface wetting at the solidliquid interface. These interfacial characteristics can be significantly altered and decoupled by introducing small quantities of additives in water, such as surface-active polymers, surfactants, and nanoparticles. The changes are fundamentally caused at a molecular-scale by the relative bulk molecular dynamics and adsorption-desorption of the additive at the liquid-vapor interface, and its physisorption and electrokinetics at the liquid-solid interface. At the micro-scale, the transient transport mechanisms at the solid-liquid-vapor interface during nucleation and bubblegrowth can be attributed to thin-film spreading, surface-micro-cavity activation, and micro-layer evaporation. Furthermore at the macro-scale, the heat transport is in turn governed by the bubble growth and distribution, macro-layer heat transfer, bubble dynamics (bubble coalescence, collapse, break-up, and translation), and liquid rheology. Some of these behaviors and processes are measured and characterized in this study, the outcomes of which advance the concomitant fundamental physics, as well as provide insights for developing control strategies for the molecular-scale manipulation of interfacial tension and surface wetting in boiling by means of polymeric reagents, surfactants, and other soluble surface-active additives.

R.M. Manglik; A. Athavale; D.S. Kalaikadal; A. Deodhar; U. Verma

2011-09-02T23:59:59.000Z

135

Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2006  

SciTech Connect (OSTI)

Neutronics and thermal-hydraulics studies show that, for equivalent operating power [85 MW(t)], a low-enriched uranium (LEU) fuel cycle based on uranium-10 wt % molybdenum (U-10Mo) metal foil with radially, “continuously graded” fuel meat thickness results in a 15% reduction in peak thermal flux in the beryllium reflector of the High Flux Isotope Reactor (HFIR) as compared to the current highly enriched uranium (HEU) cycle. The uranium-235 content of the LEU core is almost twice the amount of the HEU core when the length of the fuel cycle is kept the same for both fuels. Because the uranium-238 content of an LEU core is a factor of 4 greater than the uranium-235 content, the LEU HFIR core would weigh 30% more than the HEU core. A minimum U-10Mo foil thickness of 84 ?m is required to compensate for power peaking in the LEU core although this value could be increased significantly without much penalty. The maximum U-10Mo foil thickness is 457?m. Annual plutonium production from fueling the HFIR with LEU is predicted to be 2 kg. For dispersion fuels, the operating power for HFIR would be reduced considerably below 85 MW due to thermal considerations and due to the requirement of a 26-d fuel cycle. If an acceptable fuel can be developed, it is estimated that $140 M would be required to implement the conversion of the HFIR site at Oak Ridge National Laboratory from an HEU fuel cycle to an LEU fuel cycle. To complete the conversion by fiscal year 2014 would require that all fuel development and qualification be completed by the end of fiscal year 2009. Technological development areas that could increase the operating power of HFIR are identified as areas for study in the future.

Primm, R. T. [ORNL] [ORNL; Ellis, R. J. [ORNL] [ORNL; Gehin, J. C. [ORNL] [ORNL; Clarno, K. T. [ORNL] [ORNL; Williams, K. A. [ORNL] [ORNL; Moses, D. L. [ORNL] [ORNL

2006-11-01T23:59:59.000Z

136

SIPHORE: Conceptual Study of a High Efficiency Neutral Beam Injector Based on Photo-detachment for Future Fusion Reactors  

SciTech Connect (OSTI)

An innovative high efficiency neutral beam injector concept for future fusion reactors is under investigation (simulation and R and D) between several laboratories in France, the goal being to perform a feasibility study for the neutralization of intense high energy (1 MeV) negative ion (NI) beams by photo-detachment.The objective of the proposed project is to put together the expertise of three leading groups in negative ion quantum physics, high power stabilized lasers and neutral beam injectors to perform studies of a new injector concept called SIPHORE (SIngle gap PHOto-neutralizer energy REcovery injector), based on the photo-detachment of negative ions and energy recovery of unneutralised ions; the main feature of SIPHORE being the relevance for the future Fusion reactors (DEMO), where high injector efficiency (up to 70-80%), technological simplicity and cost reduction are key issues to be addressed.The paper presents the on-going developments and simulations around this project, such as, a new concept of ion source which would fit with this injector topology and which could solve the remaining uniformity issue of the large size ion source, and, finally, the presentation of the R and D program in the laboratories (LAC, ARTEMIS) around the photo-neutralization for Siphore.

Simonin, A.; Christin, L.; Esch, H. de; Garibaldi, P.; Grand, C.; Villecroze, F. [IRFM, CEA Cadarache, IRFM, St. Paul-lez-Durance (France); Blondel, C.; Delsart, C.; Drag, C.; Vandevraye, M. [LAC :Aime-Cotton Laboratory, Univ. Paris-sud, Orsay (France); Brillet, A.; Chaibi, W. [ARTEMIS Laboratory, Cote-d'azur Observatory, Nice (France)

2011-09-26T23:59:59.000Z

137

Effect of flux addition on the microstructure and hardness of TiC-reinforced ferrous surface composite layers fabricated by high-energy electron beam irradiation  

SciTech Connect (OSTI)

Surface composites reinforced with TiC particulates were fabricated by high-energy electron-beam irradiation. In order to investigate the effects of flux addition on the TiC dispersion in surface composite layers, four kinds of powder mixtures were made by mixing TiC with 5, 10, 20, and 40 wt% of the flux components (MgO-CaO). To fabricate TiC-reinforced surface composites, the TiC-flux mixtures were deposited evenly on a plain carbon steel substrate, which was subjected to electron-beam irradiation. Microstructural analysis was conducted using X-ray diffraction and Moessbauer spectroscopy as well as optical and scanning electron microscopy. The microstructure of the surface composites was composed of a melted region, an interfacial region, a coarse-grained heat-affected zone (HAZ), a fine-grained HAZ, and an unaltered original substrate region. TiC agglomerates and residual pores were found in the melted region of materials processed without flux, but the number of agglomerates and pores are significantly decreased in materials processed with a considerable amount of flux. As a result of irradiation, TiC particles were homogeneously distributed throughout the melted region of 2.5 mm in thickness, whose hardness was greatly increased. The optimum flux amount, which resulted in surface composites containing homogeneously dispersed TiC particles, was found to be in the range of 10 to 20% to obtain excellent surface composites.

Choo, S.H.; Lee, S.; Kwon, S.J.

1999-12-01T23:59:59.000Z

138

Design of a low enrichment, enhanced fast flux core for the Massachusetts Institute of Technology Research Reactor  

E-Print Network [OSTI]

Worldwide, there is limited test reactor capacity to perform the required irradiation experiments on advanced fast reactor materials and fuel designs. This is particularly true in the U.S., which no longer has an operating ...

Ellis, Tyler Shawn

2009-01-01T23:59:59.000Z

139

Assumptions and Criteria for Performing a Feasability Study of the Conversion of the High Flux Isotope Reactor Core to Use Low-Enriched Uranium Fuel  

SciTech Connect (OSTI)

A computational study will be initiated during fiscal year 2006 to examine the feasibility of converting the High Flux Isotope Reactor from highly enriched uranium fuel to low-enriched uranium. The study will be limited to steady-state, nominal operation, reactor physics and thermal-hydraulic analyses of a uranium-molybdenum alloy that would be substituted for the current fuel powder--U{sub 3}O{sub 8} mixed with aluminum. The purposes of this document are to (1) define the scope of studies to be conducted, (2) define the methodologies to be used to conduct the studies, (3) define the assumptions that serve as input to the methodologies, (4) provide an efficient means for communication with the Department of Energy and American research reactor operators, and (5) expedite review and commentary by those parties.

Primm, R.T., III; Ellis, R.J.; Gehin, J.C.; Moses, D.L.; Binder, J.L.; Xoubi, N. (U. of Cincinnati)

2006-02-01T23:59:59.000Z

140

Control of power characteristics of ion flow in plasma-etching reactor based on beam-plasma discharge  

SciTech Connect (OSTI)

It is shown that on the basis of the earlier revealed effect of generating the ion flow in the beam-plasma discharge from the discharge axis, a plasma processing reactor can be created for low-energy etching of semiconductor structures. The possibility of easily controlling the density and energy of ion flow by means of varying the potential of the discharge collector is demonstrated. The charge compensation of the ion flow incident on the nonconducting surface is implemented using the modulation of the potential of the substrate holder as well as the plasma-potential modulation.

Isaev, N. V.; Klykov, I. L.; Shustin, E. G., E-mail: shustin@ms.ire.rssi.ru [Russian Academy of Sciences, Kotel'nikov Institute of Radio Engineering and Electronics (Fryazino Branch) (Russian Federation)

2011-12-15T23:59:59.000Z

Note: This page contains sample records for the topic "flux beam reactor" 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

Research reactors - an overview  

SciTech Connect (OSTI)

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

West, C.D.

1997-03-01T23:59:59.000Z

142

The intense slow positron beam facility at the PULSTAR reactor and applications in nano-materials study  

SciTech Connect (OSTI)

An intense slow positron beam has been established at the PULSTAR nuclear research reactor of North Carolina State University. The slow positrons are generated by pair production in a tungsten moderator from gammarays produced in the reactor core and by neutron capture reactions in cadmium. The moderated positrons are electrostatically extracted and magnetically guided out of the region near the core. Subsequently, the positrons are used in two spectrometers that are capable of performing positron annihilation lifetime spectroscopy (PALS) and positron Doppler broadening spectroscopy (DBS) to probe the defect and free volume properties of materials. One of the spectrometers (e{sup +}-PALS) utilizes an rf buncher to produce a pulsed beam and has a timing resolution of 277 ps. The second spectrometer (Ps-PALS) uses a secondary electron timing technique and is dedicated to positronium lifetime measurements with an approximately 1 ns timing resolution. PALS measurements have been conducted in the e{sup +}-PALS spectrometer on a series of nano-materials including organic photovoltaic thin films, membranes for filtration, and polymeric fibers. These studies have resulted in understanding some critical issues related to the development of the examined nano-materials.

Liu, Ming; Moxom, Jeremy; Hawari, Ayman I. [Nuclear Reactor Program, Department of Nuclear Engineering, North Carolina State University, P.O. Box 7909, Raleigh, NC 27695 (United States); Gidley, David W. [Department of Physics, University of Michigan, 450 Church Street, Ann Arbor MI 48109 (United States)

2013-04-19T23:59:59.000Z

143

Source Terms for HFIR Beam Tube Shielding Analyses, and a Complete Shielding Analysis of the HB-3 Tube  

SciTech Connect (OSTI)

The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory is in the midst of a massive upgrade program to enhance experimental facilities. The reactor presently has four horizontal experimental beam tubes, all of which will be replaced or redesigned. The HB-2 beam tube will be enlarged to support more guide tubes, while the HB-4 beam tube will soon include a cold neutron source.

Bucholz, J.A.

2000-07-01T23:59:59.000Z

144

Instrumentation for Neutron Scattering at the Missouri University Research Reactor Paul F. Miceli  

E-Print Network [OSTI]

Instrumentation for Neutron Scattering at the Missouri University Research Reactor Paul F. Miceli Research Reactor (MURR) provides significant thermal neutron flux, which enables neutron scattering]. There are presently 5 instruments located on the beam port floor that are dedicated to neutron scattering: (1) TRIAX

Montfrooij, Wouter

145

CRAD, Fire Protection - Oak Ridge National Laboratory High Flux...  

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

Fire Protection - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Fire Protection - Oak Ridge National Laboratory High Flux Isotope Reactor February 2006 A section of...

146

CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope...  

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

Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C to DOE G...

147

CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope...  

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

Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of...

148

CRAD, Nuclear Safety - Oak Ridge National Laboratory High Flux...  

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

Nuclear Safety - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Nuclear Safety - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of...

149

CRAD, Safety Basis - Oak Ridge National Laboratory High Flux...  

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

Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Safety Basis - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR February 2007 A...

150

CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope...  

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

Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR...

151

CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope...  

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

Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR...

152

CRAD, Management- Oak Ridge National Laboratory High Flux Isotope...  

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

Management- Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Management- Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C...

153

Type B investigation of the iridium contamination event at the High Flux Isotope Reactor on September 7, 1993  

SciTech Connect (OSTI)

On the title date, at ORNL, area radiation alarms sounded during a routine transfer of a shielding cask (containing 60 Ci{sup 192}Ir) from the HFIR pool side to a transport truck. Small amounts of Ir were released from the cask onto the reactor bay floor. The floor was cleaned, and the cask was shipped to a hot cell at Building 3047 on Oct. 3, 1993. The event was caused by rupture of one of the Ir target rods after it was loaded into the cask for normal transport operations; the rupture was the result of steam generation in the target rod soon after it was placed in the cask (water had entered the target rod through a tiny defect in a weld while it was in the reactor under pressure). While the target rods were in the reactor and reactor pool, there was sufficient cooling to prevent steam generation; when the target rod was loaded into the dry transport cask, the temperature increased enough to result in boiling of the trapped water and produced high enough pressure to result in rupture. The escaping steam ejected some of the Ir pellets. The event was reported as Occurrence Report Number ORO--MMES-X10HFIR-1993-0030, dated Sept. 8, 1993. Analysis indicated that the following conditions were probable causes: less than adequate welding procedures, practices, or techniques, material controls, or inspection methods, or combination thereof, could have led to weld defects, affecting the integrity of target rod IR-75; less than adequate secondary containment in the cask allowed Ir pellets to escape.

Not Available

1994-03-01T23:59:59.000Z

154

Design and characterization of 2.45 GHz electron cyclotron resonance plasma source with magnetron magnetic field configuration for high flux of hyperthermal neutral beam  

SciTech Connect (OSTI)

A 2.45 GHz electron cyclotron resonance (ECR) source with a magnetron magnetic field configuration was developed to meet the demand of a hyperthermal neutral beam (HNB) flux on a substrate of more than 1x10{sup 15} cm{sup -2} s{sup -1} for industrial applications. The parameters of the operating pressure, ion density, electron temperature, and distance between the neutralization plate and the substrate for the HNB source are specified in a theoretical analysis. The electron temperature and the ion density are measured to characterize the ECR HNB source using a Langmuir probe and optical emission spectroscopy. The parameters of the ECR HNB source are in good agreement with the theoretically specified parameters.

Kim, Seong Bong [Department of Physics, Pohang University of Science and Technology, San 31, Hyoja-dong, Nam-gu, Pohang 790-784 (Korea, Republic of); Convergence Plasma Research Center, National Fusion Research Institute, Gwahangno 113, Yuseong-gu, Daejeon 305-333 (Korea, Republic of); Kim, Dae Chul; Yoo, Suk Jae [Convergence Plasma Research Center, National Fusion Research Institute, Gwahangno 113, Yuseong-gu, Daejeon 305-333 (Korea, Republic of); Namkung, Won; Cho, Moohyun [Department of Physics, Pohang University of Science and Technology, San 31, Hyoja-dong, Nam-gu, Pohang 790-784 (Korea, Republic of)

2010-08-15T23:59:59.000Z

155

Monte-Carlo Modeling of Parameters of a Subcritical Cascade Reactor Based on MSBR and LMFBR Technologies  

E-Print Network [OSTI]

Parameters of a subcritical cascade reactor driven by a proton accelerator and based on a primary lead-bismuth target, main reactor constructed analogously to the molten salt breeder (MSBR) reactor core and a booster-reactor analogous to the core of the BN-350 liquid metal cooled fast breeder reactor (LMFBR). It is shown by means of Monte-Carlo modeling that the reactor under study provides safe operation modes (k_{eff}=0.94-0.98), is apable to transmute effectively radioactive nuclear waste and reduces by an order of magnitude the requirements on the accelerator beam current. Calculations show that the maximal neutron flux in the thermal zone is 10^{14} cm^{12}\\cdot s^_{-1}, in the fast booster zone is 5.12\\cdot10^{15} cm^{12}\\cdot s{-1} at k_{eff}=0.98 and proton beam current I=2.1 mA.

Bznuni, S A; Zhamkochyan, V M; Polanski, A; Sosnin, A N; Khudaverdyan, A H

2001-01-01T23:59:59.000Z

156

Effusive molecular beam-sampled Knudsen flow reactor coupled to vacuum ultraviolet single photon ionization mass spectrometry using an external free radical source  

SciTech Connect (OSTI)

A new apparatus using vacuum ultraviolet single photon ionization mass spectrometry (VUV SPIMS) of an effusive molecular beam emanating from a Knudsen flow reactor is described. It was designed to study free radical-molecule kinetics over a significant temperature range (300–630 K). Its salient features are: (1) external free radical source, (2) counterpropagating molecular beam and diffuse VUV photon beam meeting in a crossed-beam ion source of a quadrupole mass spectrometer with perpendicular ion extraction, (3) analog detection of the photocurrent of the free radical molecular cation, and (4) possibility of detecting both free radicals and closed shell species in the same apparatus and under identical reaction conditions owing to the presence of photoelectrons generated by the photoelectric effect of the used VUV-photons. The measured thermal molecular beam-to-background ratio was 6.35 ± 0.39 for Ar and 10.86 ± 1.59 for i-C{sub 4}H{sub 10} at 300 K, a factor of 2.52 and 1.50 smaller, respectively, than predicted from basic gas-dynamic considerations. Operating parameters as well as the performance of key elements of the instrument are presented and discussed. Coupled to an external free radical source a steady-state specific exit flow of 1.6 × 10{sup 11} and 5.0 × 10{sup 11} molecule s{sup ?1} cm{sup ?3} of C{sub 2}H{sub 5}{sup •} (ethyl) and t-C{sub 4}H{sub 9}{sup •} (t-butyl) free radicals have been detected using VUV SPIMS at their molecular ion m/z 29 and 57, respectively, at 300 K.

Leplat, N.; Rossi, M. J. [Laboratory of Atmospheric Chemistry (LAC), Paul Scherrer Institute (PSI), CH-5232 Villigen PSI (Switzerland)] [Laboratory of Atmospheric Chemistry (LAC), Paul Scherrer Institute (PSI), CH-5232 Villigen PSI (Switzerland)

2013-11-15T23:59:59.000Z

157

CRAD, Safety Basis - Oak Ridge National Laboratory High Flux...  

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

Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Safety Basis - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C to DOE G...

158

Neutronic reactor  

DOE Patents [OSTI]

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

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

1983-01-01T23:59:59.000Z

159

Beam Characterization at the Neutron Radiography Facility  

SciTech Connect (OSTI)

The quality of a neutron imaging beam directly impacts the quality of radiographic images produced using that beam. Fully characterizing a neutron beam, including determination of the beam’s effective length-to-diameter ratio, neutron flux profile, energy spectrum, image quality, and beam divergence, is vital for producing quality radiographic images. This project characterized the east neutron imaging beamline at the Idaho National Laboratory Neutron Radiography Reactor (NRAD). The experiments which measured the beam’s effective length-to-diameter ratio and image quality are based on American Society for Testing and Materials (ASTM) standards. An analysis of the image produced by a calibrated phantom measured the beam divergence. The energy spectrum measurements consist of a series of foil irradiations using a selection of activation foils, compared to the results produced by a Monte Carlo n-Particle (MCNP) model of the beamline. Improvement of the existing NRAD MCNP beamline model includes validation of the model’s energy spectrum and the development of enhanced image simulation methods. The image simulation methods predict the radiographic image of an object based on the foil reaction rate data obtained by placing a model of the object in front of the image plane in an MCNP beamline model.

Sarah Morgan; Jeffrey King

2013-01-01T23:59:59.000Z

160

Physics design of a 100 keV acceleration grid system for the diagnostic neutral beam for international tokamak experimental reactor  

SciTech Connect (OSTI)

This paper describes the physics design of a 100 keV, 60 A H{sup -} accelerator for the diagnostic neutral beam (DNB) for international tokamak experimental reactor (ITER). The accelerator is a three grid system comprising of 1280 apertures, grouped in 16 groups with 80 apertures per beam group. Several computer codes have been used to optimize the design which follows the same philosophy as the ITER Design Description Document (DDD) 5.3 and the 1 MeV heating and current drive beam line [R. Hemsworth, H. Decamps, J. Graceffa, B. Schunke, M. Tanaka, M. Dremel, A. Tanga, H. P. L. De Esch, F. Geli, J. Milnes, T. Inoue, D. Marcuzzi, P. Sonato, and P. Zaccaria, Nucl. Fusion 49, 045006 (2009)]. The aperture shapes, intergrid distances, and the extractor voltage have been optimized to minimize the beamlet divergence. To suppress the acceleration of coextracted electrons, permanent magnets have been incorporated in the extraction grid, downstream of the cooling water channels. The electron power loads on the extractor and the grounded grids have been calculated assuming 1 coextracted electron per ion. The beamlet divergence is calculated to be 4 mrad. At present the design for the filter field of the RF based ion sources for ITER is not fixed, therefore a few configurations of the same have been considered. Their effect on the transmission of the electrons and beams through the accelerator has been studied. The OPERA-3D code has been used to estimate the aperture offset steering constant of the grounded grid and the extraction grid, the space charge interaction between the beamlets and the kerb design required to compensate for this interaction. All beamlets in the DNB must be focused to a single point in the duct, 20.665 m from the grounded grid, and the required geometrical aimings and aperture offsets have been calculated.

Singh, M. J. [ITER-India, Institute for Plasma Research, Bhat, Gandhinagar, Gujarat 382428 (India); De Esch, H. P. L. [CEA-Cadarache, IRFM, F-13108 Saint-Paul-lez-Durance (France)

2010-01-15T23:59:59.000Z

Note: This page contains sample records for the topic "flux beam reactor" 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

Method of Fission Product Beta Spectra Measurements for Predicting Reactor Anti-neutrino Emission  

E-Print Network [OSTI]

The nuclear fission process that occurs in the core of nuclear reactors results in unstable, neutron rich fission products that subsequently beta decay and emit electron anti-neutrinos. These reactor neutrinos have served neutrino physics research from the initial discovery of the neutrino to current precision measurements of neutrino mixing angles. The prediction of the absolute flux and energy spectrum of the emitted reactor neutrinos hinges upon a series of seminal papers based on measurements performed in the 1970s and 1980s. The steadily improving reactor neutrino measurement techniques and recent re-considerations of the agreement between the predicted and observed reactor neutrino flux motivates revisiting the underlying beta spectra measurements. A method is proposed to use an accelerator proton beam delivered to an engineered target to yield a neutron field tailored to reproduce the neutron energy spectrum present in the core of an operating nuclear reactor. Foils of the primary reactor fissionable isotopes placed in this tailored neutron flux will ultimately emit beta particles from the resultant fission products. Measurement of these beta particles in a time projection chamber with a perpendicular magnetic field provides a distinctive set of systematic considerations for comparison to the original seminal beta spectra measurements. Ancillary measurements such as gamma-ray emission and post-irradiation radiochemical analysis will further constrain the absolute normalization of beta emissions per fission. The requirements for unfolding the beta spectra measured with this method into a predicted reactor neutrino spectrum are explored.

D. M. Asner; K. Burns; L. W. Campbell; B. Greenfield; M. S. Kos; J. L. Orrell; M. Schram; B. VanDevender; 1 L. S. Wood; D. W. Wootan

2014-03-01T23:59:59.000Z

162

Production capabilities in US nuclear reactors for medical radioisotopes  

SciTech Connect (OSTI)

The availability of reactor-produced radioisotopes in the United States for use in medical research and nuclear medicine has traditionally depended on facilities which are an integral part of the US national laboratories and a few reactors at universities. One exception is the reactor in Sterling Forest, New York, originally operated as part of the Cintichem (Union Carbide) system, which is currently in the process of permanent shutdown. Since there are no industry-run reactors in the US, the national laboratories and universities thus play a critical role in providing reactor-produced radioisotopes for medical research and clinical use. The goal of this survey is to provide a comprehensive summary of these production capabilities. With the temporary shutdown of the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) in November 1986, the radioisotopes required for DOE-supported radionuclide generators were made available at the Brookhaven National Laboratory (BNL) High Flux Beam Reactor (HFBR). In March 1988, however, the HFBR was temporarily shut down which forced investigators to look at other reactors for production of the radioisotopes. During this period the Missouri University Research Reactor (MURR) played an important role in providing these services. The HFIR resumed routine operation in July 1990 at 85 MW power, and the HFBR resumed operation in June 1991, at 30 MW power. At the time of the HFBR shutdown, there was no available comprehensive overview which could provide information on status of the reactors operating in the US and their capabilities for radioisotope production. The obvious need for a useful overview was thus the impetus for preparing this survey, which would provide an up-to-date summary of those reactors available in the US at both the DOE-funded national laboratories and at US universities where service irradiations are currently or expected to be conducted.

Mirzadeh, S.; Callahan, A.P.; Knapp, F.F. Jr. (Oak Ridge National Lab., TN (United States)); Schenter, R.E. (Westinghouse Hanford Co., Richland, WA (United States))

1992-11-01T23:59:59.000Z

163

A high-speed beam of lithium droplets for collecting diverted energy and particles in ITER (International Thermonuclear Experimental Reactor)  

SciTech Connect (OSTI)

A high-speed (160m/s) beam (0.14 {times} 0.86m) of liquid-lithium droplets passing through the divertor region(s) below (and above) the main plasma has the potential to replace and out-perform conventional'' solid divertor plates in both heat and particle removal. In addition to superior heat-collection properties, the lithium beam would: remove impurities; require low power to circulate the lithium; exhibit low-recycle divertor operation compatible with lower-hybrid current drive, H-mode plasma confinement, and no flow reversal in the edge plasma; be insensitive to plasma shifts; and finally protect solid structures from the plasma thermal energy for those disruptions that deposit energy preferentially into the divertor while simultaneously being rapidly re-established after a major disruption. Scoping calculations identifying the beam configuration and the droplet dynamics, including formation, MHD effects, gravitational effects, thermal response and hydrodynamics, are presented. Limitations and uncertainties are also discussed. 20 refs., 6 figs., 3 tabs.

Werley, K.A.

1989-01-01T23:59:59.000Z

164

Advanced Neutron Source Reactor thermal analysis of fuel plate defects  

SciTech Connect (OSTI)

The Advanced Neutron Source Reactor (ANSR) is a research reactor designed to provide the highest continuous neutron beam intensity of any reactor in the world. The present technology for determining safe operations were developed for the High Flux Isotope Reactor (HFIR). These techniques are conservative and provide confidence in the safe operation of HFIR. However, the more intense requirements of ANSR necessitate the development of more accurate, but still conservative, techniques. This report details the development of a Local Analysis Technique (LAT) that provides an appropriate approach. Application of the LAT to two ANSR core designs are presented. New theories of the thermal and nuclear behavior of the U{sub 3}Si{sub 2} fuel are utilized. The implications of lower fuel enrichment and of modifying the inspection procedures are also discussed. Development of the computer codes that enable the automate execution of the LAT is included.

Giles, G.E.

1995-08-01T23:59:59.000Z

165

Molecular Beam Epitaxy, Multi-source | EMSL  

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

beam epitaxy, is examined using a combination... Self-corrected Sensors Based On Atomic Absorption Spectroscopy For Atom Flux Measurements In Molecular Beam Epitaxy. A...

166

Fracture analysis of HFIR beam tube caused by radiation embrittlement  

SciTech Connect (OSTI)

With an attempt to estimate the neutron beam tube embrittlement condition for the Oak Ridge High Flux Isotope Reactor (HFIR), fracture mechanics calculations are carried out in this paper. The analysis provides some numerical result on how the tube has been structurally weakened. In this calculation, a lateral impact force is assumed. Numerical result is obtained on how much the critical crack size should be reduced if the beam tube has been subjected to an extended period of irradiation. It is also calculated that buckling strength of the tube is increased, not decreased, with irradiation.

Chang, S.J. [Oak Ridge National Lab., TN (United States). Research Reactors Div.

1994-12-31T23:59:59.000Z

167

Beta ray flux measuring device  

DOE Patents [OSTI]

A beta ray flux measuring device in an activated member in-core instrumentation system for pressurized water reactors. The device includes collector rings positioned about an axis in the reactor's pressure boundary. Activated members such as hydroballs are positioned within respective ones of the collector rings. A response characteristic such as the current from or charge on a collector ring indicates the beta ray flux from the corresponding hydroball and is therefore a measure of the relative nuclear power level in the region of the reactor core corresponding to the specific exposed hydroball within the collector ring.

Impink, Jr., Albert J. (Murrysville, PA); Goldstein, Norman P. (Murrysville, PA)

1990-01-01T23:59:59.000Z

168

The Advanced Neutron Source (ANS) project: A world-class research reactor facility  

SciTech Connect (OSTI)

This paper provides an overview of the Advanced Neutron Source (ANS), a new research facility being designed at Oak Ridge National Laboratory. The facility is based on a 330 MW, heavy-water cooled and reflected reactor as the neutron source, with a thermal neutron flux of about 7.5{times}10{sup 19}m{sup {minus}2}{center_dot}sec{sup {minus}1}. Within the reflector region will be one hot source which will serve 2 hot neutron beam tubes, two cryogenic cold sources serving fourteen cold neutron beam tubes, two very cold beam tubes, and seven thermal neutron beam tubes. In addition there will be ten positions for materials irradiation experiments, five of them instrumented. The paper touches on the project status, safety concerns, cost estimates and scheduling, a description of the site, the reactor, and the arrangements of the facilities.

Thompson, P.B. [Martin Marietta Energy Systems, Inc., Oak Ridge, TN (US); Meek, W.E. [Gilbert/Commonwealth, Inc., Pittsburgh, PA (US)

1993-07-01T23:59:59.000Z

169

Direct atomic flux measurement of electron-beam evaporated yttrium with a diode-laser-based atomic absorption monitor at 668 nm  

E-Print Network [OSTI]

with a diode-laser-based atomic absorption AA monitor at 668 nm. Atomic number density and velocity were measured through absorption and Doppler shift measurements to provide the atomic flux. The AA previously developed diode-laser-based atomic absorption AA monitors for atomic density measurements

Fejer, Martin M.

170

Final Report Independent Verification Survey of the High Flux Beam Reactor, Building 802 Fan House Brookhaven National Laboratory Upton, New York  

SciTech Connect (OSTI)

The Separations Process Research Unit (SPRU) complex located on the Knolls Atomic Power Laboratory (KAPL) site in Niskayuna, New York, was constructed in the late 1940s to research the chemical separation of plutonium and uranium (Figure A-1). SPRU operated as a laboratory scale research facility between February 1950 and October 1953. The research activities ceased following the successful development of the reduction oxidation and plutonium/uranium extraction processes. The oxidation and extraction processes were subsequently developed for large scale use by the Hanford and Savannah River sites (aRc 2008a). Decommissioning of the SPRU facilities began in October 1953 and continued through the 1990s.

Evan Harpeneau

2011-06-24T23:59:59.000Z

171

Characterization of Dosimetry of the BMRR Horizontal Thimble Tubes and Broad Beam Facility.  

SciTech Connect (OSTI)

The Brookhaven Medical Research Reactor was a 5 mega-watt, light-water cooled and heavy-graphite moderated research facility. It has two shutter-equipped treatment rooms, three horizontally extended thimble tubes, and an ex-core broad beam facility. The three experimental thimbles, or activation ports, external to the reactor tank were designed for several uses, including the investigations on diagnostic and therapeutic methods using radioactive isotopes of very short half-life, the analysis of radiation exposure on tissue-equivalent materials using a collimated neutron beam, and the evaluation of dose effects on biological cells to improve medical treatment. At the broad beam facility where the distribution of thermal neutrons was essential uniform, a wide variety of mammalian whole-body exposures were studied using animals such as burros or mice. Also studied at the broad beam were whole-body phantom experiments, involving the use of a neutron or photon beam streaming through a screen to obtain the flux spectrum suitable for dose analysis on the sugar-urea-water mixture, a tissue-equivalent material. Calculations of the flux and the dose at beam ports based on Monte Carlo particle-transport code were performed, and measurements conducted at the same tally locations were made using bare or cadmium-covered gold foils. Analytical results, which show good agreement with measurement data, are presented in the paper.

Hu,J.P.; Reciniello, R.N.; Holden, N.E.

2008-05-05T23:59:59.000Z

172

Subcritical Fission Reactor Based on Linear Collider  

E-Print Network [OSTI]

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

I. F. Ginzburg

2005-07-29T23:59:59.000Z

173

E-Print Network 3.0 - almaty wwr-k reactor Sample Search Results  

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

High Flux Isotope Reactor Center for Nanophase Materials Sciences... International Thermonuclear Experimental Reactor Center for Computational Sciences National Security 0 0 61 1...

174

High Heat Flux Components Program  

SciTech Connect (OSTI)

Purpose is the development of the technologies necessary to design, build and operate high heat flux components such as actively cooled limiters, divertor collector plates, R.F. antennas, mirror end cells, mirror halo collectors, direct convertor collectors, and neutral beam dumps. These components require an integrated design that considers the plasma-materials interaction (PMI) issues, heat removal problems and materials issues (including possible low Z coatings and claddings). As a general definition, high heat flux components see heat fluxes ranging from 1 to 100 MW/m/sup 2/. Suitable materials include copper and copper alloys.

Whitley, J.B.

1983-01-01T23:59:59.000Z

175

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

176

Horizontal Beam Tubes - HFIR Technical Parameters | ORNL Neutron...  

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

Beam Tubes The reactor has four horizontal beam tubes that supply the neutrons to the neutron scattering instruments. Details for each beam tube and instrument can be found on...

177

Predicting Reactor Antineutrino Emissions Using New Precision Beta Spectroscopy  

SciTech Connect (OSTI)

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

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

2013-05-01T23:59:59.000Z

178

A coupled neutronics/thermalhydraulics tool for calculating fluctuations in Pressurized Water Reactors  

E-Print Network [OSTI]

Water Reactors or Heavy Water Reactors (LarssoA coupled neutronics/thermal­hydraulics tool for calculating fluctuations in Pressurized Water in neutron flux, fuel temperature, moderator den- sity and flow velocity in Pressurized Water Reactors

Demazière, Christophe

179

HFIR | High Flux Isotope Reactor | ORNL  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary)morphinanInformation Desert Southwest Region service area. TheEPSCIResearchGulf of Mexico Fact SheetHFIR User

180

akm reactor: Topics by E-print Network  

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

the flux calculation and recent progresses. Jun Cao 2012-03-08 9 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

Note: This page contains sample records for the topic "flux beam reactor" 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

akm muehleberg reactor: Topics by E-print Network  

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

the flux calculation and recent progresses. Jun Cao 2012-03-08 9 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

182

almaraz-2 reactor: Topics by E-print Network  

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

the flux calculation and recent progresses. Jun Cao 2012-03-08 5 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

183

almaraz-1 reactor: Topics by E-print Network  

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

the flux calculation and recent progresses. Jun Cao 2012-03-08 5 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

184

Boiling-Water Reactor internals aging degradation study. Phase 1  

SciTech Connect (OSTI)

This report documents the results of an aging assessment study for boiling water reactor (BWR) internals. Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primary coolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant, applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two major aging-related degradation mechanisms for BWR internals. Significant reported failures include SCC in jet-pump holddown beams, in-core neutron flux monitor dry tubes and core spray spargers. Fatigue failures were detected in feedwater spargers. The implementation of a plant Hydrogen Water Chemistry (HWC) program is considered as a promising method for controlling SCC problems in BWR. More operating data are needed to evaluate its effectiveness for internal components. Long-term fast neutron irradiation effects and high-cycle fatigue in a corrosive environment are uncertainty factors in the aging assessment process. BWR internals are examined by visual inspections and the method is access limited. The presence of a large water gap and an absence of ex-core neutron flux monitors may handicap the use of advanced inspection methods, such as neutron noise vibration measurements, for BWR.

Luk, K.H. [Oak Ridge National Lab., TN (United States)

1993-09-01T23:59:59.000Z

185

Reactor Physics  

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

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

186

Advanced Test Reactor National Scientific User Facility Partnerships  

SciTech Connect (OSTI)

In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of Wisconsin-Madison; (8) Illinois Institute of Technology (IIT) Materials Research Collaborative Access Team (MRCAT) beamline at Argonne National Laboratory's Advanced Photon Source; and (9) Nanoindenter in the University of California at Berkeley (UCB) Nuclear Engineering laboratory Materials have been analyzed for ATR NSUF users at the Advanced Photon Source at the MRCAT beam, the NIST Center for Neutron Research in Gaithersburg, MD, the Los Alamos Neutron Science Center, and the SHaRE user facility at Oak Ridge National Laboratory (ORNL). Additionally, ORNL has been accepted as a partner facility to enable ATR NSUF users to access the facilities at the High Flux Isotope Reactor and related facilities.

Frances M. Marshall; Todd R. Allen; Jeff B. Benson; James I. Cole; Mary Catherine Thelen

2012-03-01T23:59:59.000Z

187

Toroidal midplane neutral beam armor and plasma limiter  

DOE Patents [OSTI]

This invention contemplates an armor shield/plasma limiter positioned upon the inner wall of a toroidal vacuum chamber within which is magnetically confined an energetic plasma in a tokamak nuclear fusion reactor. The armor shield/plasma limiter is thus of a general semi-toroidal shape and is comprised of a plurality of adjacent graphite plates positioned immediately adjacent to each other so as to form a continuous ring upon and around the toroidal chamber's inner wall and the reactor's midplane coil. Each plate has a generally semi-circular outer circumference and a recessed inner portion and is comprised of upper and lower half sections positioned immediately adjacent to one another along the midplane of the plate. With the upper and lower half sections thus joined, a channel or duct is provided within the midplane of the plate in which a magnetic flux loop is positioned. The magnetic flux loop is thus positioned immediately adjacent to the fusing toroidal plasma and serves as a diagnostic sensor with the armor shield/plasma limiter minimizing the amount of power from the energetic plasma as well as from the neutral particle beams heating the plasma incident upon the flux loop.

Kugel, H.W.; Hand, S.W. Jr.; Ksayian, H.

1985-05-31T23:59:59.000Z

188

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

SciTech Connect (OSTI)

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

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

2012-04-04T23:59:59.000Z

189

Reactor antineutrino monitoring with a plastic scintillator array as a new safeguards method  

E-Print Network [OSTI]

We developed a segmented reactor-antineutrino detector made of plastic scintillators for application as a tool in nuclear safeguards inspection and performed mostly unmanned field operations at a commercial power plant reactor. At a position outside the reactor building, we measured the difference in reactor antineutrino flux above the ground when the reactor was active and inactive.

S. Oguri; Y. Kuroda; Y. Kato; R. Nakata; Y. Inoue; C. Ito; M. Minowa

2014-05-23T23:59:59.000Z

190

Beam current sensor  

DOE Patents [OSTI]

A current sensor for measuring the dc component of a beam of charged particles employs a superconducting pick-up loop probe, with twisted superconducting leads in combination with a Superconducting Quantum Interference Device (SQUID) detector. The pick-up probe is in the form of a single-turn loop, or a cylindrical toroid, through which the beam is directed and within which a first magnetic flux is excluded by the Meisner effect. The SQUID detector acts as a flux-to-voltage converter in providing a current to the pick-up loop so as to establish a second magnetic flux within the electrode which nulls out the first magnetic flux. A feedback voltage within the SQUID detector represents the beam current of the particles which transit the pick-up loop. Meisner effect currents prevent changes in the magnetic field within the toroidal pick-up loop and produce a current signal independent of the beam's cross-section and its position within the toroid, while the combination of superconducting elements provides current measurement sensitivities in the nano-ampere range.

Kuchnir, M.; Mills, F.E.

1984-09-28T23:59:59.000Z

191

Beam current sensor  

DOE Patents [OSTI]

A current sensor for measuring the DC component of a beam of charged particles employs a superconducting pick-up loop probe, with twisted superconducting leads in combination with a Superconducting Quantum Interference Device (SQUID) detector. The pick-up probe is in the form of a single-turn loop, or a cylindrical toroid, through which the beam is directed and within which a first magnetic flux is excluded by the Meisner effect. The SQUID detector acts as a flux-to-voltage converter in providing a current to the pick-up loop so as to establish a second magnetic flux within the electrode which nulls out the first magnetic flux. A feedback voltage within the SQUID detector represents the beam current of the particles which transit the pick-up loop. Meisner effect currents prevent changes in the magnetic field within the toroidal pick-up loop and produce a current signal independent of the beam's cross-section and its position within the toroid, while the combination of superconducting elements provides current measurement sensitivites in the nano-ampere range.

Kuchnir, Moyses (Elmhurst, IL); Mills, Frederick E. (Elburn, IL)

1987-01-01T23:59:59.000Z

192

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

193

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

194

Heat dissipating nuclear reactor  

DOE Patents [OSTI]

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

Hunsbedt, A.; Lazarus, J.D.

1985-11-21T23:59:59.000Z

195

Solid tags for identifying failed reactor components  

DOE Patents [OSTI]

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

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

1987-01-01T23:59:59.000Z

196

Nuclear reactor shield including magnesium oxide  

DOE Patents [OSTI]

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

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

1981-01-01T23:59:59.000Z

197

Pulsed ion beam source  

DOE Patents [OSTI]

An improved pulsed ion beam source having a new biasing circuit for the fast magnetic field. This circuit provides for an initial negative bias for the field created by the fast coils in the ion beam source which pre-ionize the gas in the source, ionize the gas and deliver the gas to the proper position in the accelerating gap between the anode and cathode assemblies in the ion beam source. The initial negative bias improves the interaction between the location of the nulls in the composite magnetic field in the ion beam source and the position of the gas for pre-ionization and ionization into the plasma as well as final positioning of the plasma in the accelerating gap. Improvements to the construction of the flux excluders in the anode assembly are also accomplished by fabricating them as layered structures with a high melting point, low conductivity material on the outsides with a high conductivity material in the center.

Greenly, John B. (Lansing, NY)

1997-01-01T23:59:59.000Z

198

High-flux neutron source based on a liquid-lithium target  

SciTech Connect (OSTI)

A prototype compact Liquid Lithium Target (LiLiT), able to constitute an accelerator-based intense neutron source, was built. The neutron source is intended for nuclear astrophysical research, boron neutron capture therapy (BNCT) in hospitals and material studies for fusion reactors. The LiLiT setup is presently being commissioned at Soreq Nuclear research Center (SNRC). The lithium target will produce neutrons through the {sup 7}Li(p,n){sup 7}Be reaction and it will overcome the major problem of removing the thermal power generated by a high-intensity proton beam, necessary for intense neutron flux for the above applications. The liquid-lithium loop of LiLiT is designed to generate a stable lithium jet at high velocity on a concave supporting wall with free surface toward the incident proton beam (up to 10 kW). During off-line tests, liquid lithium was flown through the loop and generated a stable jet at velocity higher than 5 m/s on the concave supporting wall. The target is now under extensive test program using a high-power electron-gun. Up to 2 kW electron beam was applied on the lithium flow at velocity of 4 m/s without any flow instabilities or excessive evaporation. High-intensity proton beam irradiation will take place at SARAF (Soreq Applied Research Accelerator Facility) superconducting linear accelerator currently in commissioning at SNRC.

Halfon, S. [Soreq NRC, Yavne, 81800 (Israel) and Racah Institute of Physics, Hebrew University, Jerusalem, 91904 (Israel); Feinberg, G. [Soreq NRC, Yavne, 81800 (Israel) and Racah Institute of Physics, Hebrew University, Jerusalem, 91904 (Israel); Paul, M. [Racah Institute of Physics, Hebrew University, Jerusalem, 91904 (Israel); Arenshtam, A.; Berkovits, D.; Kijel, D.; Nagler, A.; Eliyahu, I.; Silverman, I. [Soreq NRC, Yavne, 81800 (Israel)

2013-04-19T23:59:59.000Z

199

Fast flux locked loop  

DOE Patents [OSTI]

A flux locked loop for providing an electrical feedback signal, the flux locked loop employing radio-frequency components and technology to extend the flux modulation frequency and tracking loop bandwidth. The flux locked loop of the present invention has particularly useful application in read-out electronics for DC SQUID magnetic measurement systems, in which case the electrical signal output by the flux locked loop represents an unknown magnetic flux applied to the DC SQUID.

Ganther, Jr., Kenneth R. (Olathe, KS); Snapp, Lowell D. (Independence, MO)

2002-09-10T23:59:59.000Z

200

Optimizing Medium Baseline Reactor Neutrino Experiments  

E-Print Network [OSTI]

10 years from now medium baseline reactor neutrino experiments will attempt to determine the neutrino mass hierarchy from the observed antineutrino spectra. In this letter we present the results of more than four million detailed simulations of such experiments, studying the dependence of the probability of successfully determining the hierarchy upon the analysis method, the neutrino mass matrix parameters, reactor flux models and, in particular, combinations of baselines. We show that the strong dependence of the hierarchy determination upon mass differences and flux models found by Qian et al. results from a spurious dependence of the Fourier analysis upon the high energy tail of the reactor spectrum which can be removed by using a weighted Fourier transform. Such experiments necessarily use flux from multiple reactors at distinct baselines, smearing the oscillation signal and thus impeding the determination of the hierarchy. Using the results of our simulations, we determine the optimal baselines and corre...

Ciuffoli, Emilio; Zhang, Xinmin

2013-01-01T23:59:59.000Z

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they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
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201

Pulse flux measuring device  

DOE Patents [OSTI]

A device for measuring particle flux comprises first and second photodiode detectors for receiving flux from a source and first and second outputs for producing first and second signals representing the flux incident to the detectors. The device is capable of reducing the first output signal by a portion of the second output signal, thereby enhancing the accuracy of the device. Devices in accordance with the invention may measure distinct components of flux from a single source or fluxes from several sources.

Riggan, William C. (Albuquerque, NM)

1985-01-01T23:59:59.000Z

202

E-Print Network 3.0 - allens creek-1 reactor Sample Search Results  

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Centre de mathmatiques Collection: Mathematics 96 Proposed Program: Reliability-Centered Maintenance (RCM) for the High Flux Isotope Reactor Summary: Proposed Program:...

203

E-Print Network 3.0 - argentine reactor ra-8 Sample Search Results  

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Centre de mathmatiques Collection: Mathematics 88 Proposed Program: Reliability-Centered Maintenance (RCM) for the High Flux Isotope Reactor Summary: Proposed Program:...

204

E-Print Network 3.0 - argentine reactor ra-3 Sample Search Results  

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Centre de mathmatiques Collection: Mathematics 93 Proposed Program: Reliability-Centered Maintenance (RCM) for the High Flux Isotope Reactor Summary: Proposed Program:...

205

E-Print Network 3.0 - acpr reactor Sample Search Results  

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Centre de mathmatiques Collection: Mathematics 99 Proposed Program: Reliability-Centered Maintenance (RCM) for the High Flux Isotope Reactor Summary: Proposed Program:...

206

E-Print Network 3.0 - annular reactor system Sample Search Results  

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University Collection: Engineering ; Physics 95 Proposed Program: Reliability-Centered Maintenance (RCM) for the High Flux Isotope Reactor Summary: Proposed Program:...

207

E-Print Network 3.0 - alto lazio-2 reactor Sample Search Results  

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Centre de mathmatiques Collection: Mathematics 69 Proposed Program: Reliability-Centered Maintenance (RCM) for the High Flux Isotope Reactor Summary: Proposed Program:...

208

E-Print Network 3.0 - argonaut lemont reactor Sample Search Results  

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Centre de mathmatiques Collection: Mathematics 89 Proposed Program: Reliability-Centered Maintenance (RCM) for the High Flux Isotope Reactor Summary: Proposed Program:...

209

E-Print Network 3.0 - autonomous reactor sstar Sample Search...  

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Centre de mathmatiques Collection: Mathematics 87 Proposed Program: Reliability-Centered Maintenance (RCM) for the High Flux Isotope Reactor Summary: Proposed Program:...

210

E-Print Network 3.0 - avogadro rs-1 reactor Sample Search Results  

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Centre de mathmatiques Collection: Mathematics 88 Proposed Program: Reliability-Centered Maintenance (RCM) for the High Flux Isotope Reactor Summary: Proposed Program:...

211

E-Print Network 3.0 - argus reactor Sample Search Results  

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Centre de mathmatiques Collection: Mathematics 92 Proposed Program: Reliability-Centered Maintenance (RCM) for the High Flux Isotope Reactor Summary: Proposed Program:...

212

E-Print Network 3.0 - astr reactor Sample Search Results  

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Centre de mathmatiques Collection: Mathematics 91 Proposed Program: Reliability-Centered Maintenance (RCM) for the High Flux Isotope Reactor Summary: Proposed Program:...

213

E-Print Network 3.0 - argentine reactor ra-4 Sample Search Results  

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Centre de mathmatiques Collection: Mathematics 88 Proposed Program: Reliability-Centered Maintenance (RCM) for the High Flux Isotope Reactor Summary: Proposed Program:...

214

E-Print Network 3.0 - alto lazio-1 reactor Sample Search Results  

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Centre de mathmatiques Collection: Mathematics 68 Proposed Program: Reliability-Centered Maintenance (RCM) for the High Flux Isotope Reactor Summary: Proposed Program:...

215

E-Print Network 3.0 - allens creek-2 reactor Sample Search Results  

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Centre de mathmatiques Collection: Mathematics 92 Proposed Program: Reliability-Centered Maintenance (RCM) for the High Flux Isotope Reactor Summary: Proposed Program:...

216

E-Print Network 3.0 - argonauta rio reactor Sample Search Results  

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Centre de mathmatiques Collection: Mathematics 91 Proposed Program: Reliability-Centered Maintenance (RCM) for the High Flux Isotope Reactor Summary: Proposed Program:...

217

E-Print Network 3.0 - apra reactor Sample Search Results  

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Centre de mathmatiques Collection: Mathematics 86 Proposed Program: Reliability-Centered Maintenance (RCM) for the High Flux Isotope Reactor Summary: Proposed Program:...

218

E-Print Network 3.0 - austrian triga-mark-ii reactor Sample Search...  

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Centre de mathmatiques Collection: Mathematics 94 Proposed Program: Reliability-Centered Maintenance (RCM) for the High Flux Isotope Reactor Summary: Proposed Program:...

219

E-Print Network 3.0 - astra reactor Sample Search Results  

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Centre de mathmatiques Collection: Mathematics 100 Proposed Program: Reliability-Centered Maintenance (RCM) for the High Flux Isotope Reactor Summary: Proposed Program:...

220

E-Print Network 3.0 - aerojet-general nucleonics reactors Sample...  

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Collection: Chemistry ; Biotechnology 78 Volume 89 number 44 28 october 2008 Summary: antineutrino flux from nuclear reactors,crust,and mantle at project sites. By...

Note: This page contains sample records for the topic "flux beam reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
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221

E-Print Network 3.0 - afrri triga reactor Sample Search Results  

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MICROMEGAS DETECTOR FOR IN-CORE NUCLEAR REACTOR NEUTRON FLUX MEASUREMENTS Summary: subcritical configuration based in TRIGA-ADS project which consists of coupling a 1 MW TRIGA...

222

Photon beam position monitor  

DOE Patents [OSTI]

A photon beam position monitor is disclosed for use in the front end of a beamline of a high heat flux and high energy photon source such as a synchrotron radiation storage ring detects and measures the position and, when a pair of such monitors are used in tandem, the slope of a photon beam emanating from an insertion device such as a wiggler or an undulator inserted in the straight sections of the ring. The photon beam position monitor includes a plurality of spaced blades for precisely locating the photon beam, with each blade comprised of chemical vapor deposition (CVD) diamond with an outer metal coating of a photon sensitive metal such as tungsten, molybdenum, etc., which combination emits electrons when a high energy photon beam is incident upon the blade. Two such monitors are contemplated for use in the front end of the beamline, with the two monitors having vertically and horizontally offset detector blades to avoid blade ''shadowing''. Provision is made for aligning the detector blades with the photon beam and limiting detector blade temperature during operation. 18 figs.

Kuzay, T.M.; Shu, D.

1995-02-07T23:59:59.000Z

223

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

224

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

225

Rhodium self-powered neutron detector as a suitable on-line thermal neutron flux monitor in BNCT treatments  

SciTech Connect (OSTI)

Purpose: A rhodium self-powered neutron detector (Rh SPND) has been specifically developed by the Comision Nacional de Energia Atomica (CNEA) of Argentina to measure locally and in real time thermal neutron fluxes in patients treated with boron neutron capture therapy (BNCT). In this work, the thermal and epithermal neutron response of the Rh SPND was evaluated by studying the detector response to two different reactor spectra. In addition, during clinical trials of the BNCT Project of the CNEA, on-line neutron flux measurements using the specially designed detector were assessed. Methods: The first calibration of the detector was done with the well-thermalized neutron spectrum of the CNEA RA-3 reactor thermal column. For this purpose, the reactor spectrum was approximated by a Maxwell-Boltzmann distribution in the thermal energy range. The second calibration was done at different positions along the central axis of a water-filled cylindrical phantom, placed in the mixed thermal-epithermal neutron beam of CNEA RA-6 reactor. In this latter case, the RA-6 neutron spectrum had been well characterized by both calculation and measurement, and it presented some marked differences with the ideal spectrum considered for SPND calibrations at RA-3. In addition, the RA-6 neutron spectrum varied with depth in the water phantom and thus the percentage of the epithermal contribution to the total neutron flux changed at each measurement location. Local (one point-position) and global (several points-positions) and thermal and mixed-field thermal neutron sensitivities were determined from these measurements. Thermal neutron flux was also measured during BNCT clinical trials within the irradiation fields incident on the patients. In order to achieve this, the detector was placed on patient's skin at dosimetric reference points for each one of the fields. System stability was adequate for this kind of measurement. Results: Local mixed-field thermal neutron sensitivities and global thermal and mixed-field thermal neutron sensitivities derived from measurements performed at the RA-6 were compared and no significant differences were found. Global RA-6-based thermal neutron sensitivity showed agreement with pure thermal neutron sensitivity measurements performed in the RA-3 spectrum. Additionally, the detector response proved nearly unchanged by differences in neutron spectra from real (RA-6 BNCT beam) and ideal (considered for calibration calculations at RA-3) neutron source descriptions. The results confirm that the special design of the Rh SPND can be considered as having a pure thermal response for neutron spectra with epithermal-to-thermal flux ratios up to 12%. In addition, the linear response of the detector to thermal flux allows the use of a mixed-field thermal neutron sensitivity of 1.95 {+-} 0.05 x 10{sup -21} A n{sup -1}{center_dot}cm{sup 2}{center_dot}s. This sensitivity can be used in spectra with up to 21% epithermal-to-thermal flux ratio without significant error due to epithermal neutron and gamma induced effects. The values of the measured fluxes in clinical applications had discrepancies with calculated results that were in the range of -25% to +30%, which shows the importance of a local on-line independent measurement as part of a treatment planning quality control system. Conclusions: The usefulness of the CNEA Rh SPND for the on-line local measurement of thermal neutron flux on BNCT patients has been demonstrated based on an appropriate neutron spectra calibration and clinical applications.

Miller, Marcelo E.; Sztejnberg, Manuel L.; Gonzalez, Sara J.; Thorp, Silvia I.; Longhino, Juan M.; Estryk, Guillermo [Comision Nacional de Energia Atomica, Av. del Libertador 8250, Ciudad de Buenos Aires 1429 (Argentina); Comision Nacional de Energia Atomica, Av. del Libertador 8250, Ciudad de Buenos Aires 1429, Argentina and CONICET, Av. Rivadavia 1917, Ciudad de Buenos Aires 1033 (Argentina); Comision Nacional de Energia Atomica, Av. del Libertador 8250, Ciudad de Buenos Aires 1429 (Argentina)

2011-12-15T23:59:59.000Z

226

Plasma instrumentation for fusion power reactor control  

SciTech Connect (OSTI)

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

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

1985-07-01T23:59:59.000Z

227

REACTOR DOSIMETRY STUDY OF THE RHODE ISLAND NUCLEAR SCIENCE CENTER.  

SciTech Connect (OSTI)

The Rhode Island Nuclear Science Center (RINSC), located on the Narragansett Bay Campus of the University of Rhode Island, is a state-owned and US NRC-licensed nuclear facility constructed for educational and industrial applications. The main building of RINSC houses a two-megawatt (2 MW) thermal power critical reactor immersed in demineralized water within a shielded tank. As its original design in 1958 by the Rhode Island Atomic Energy Commission focused on the teaching and research use of the facility, only a minimum of 3.85 kg fissile uranium-235 was maintained in the fuel elements to allow the reactor to reach a critical state. In 1986 when RINSC was temporarily shutdown to start US DOE-directed core conversion project for national security reasons, all the U-Al based Highly-Enriched Uranium (HEU, 93% uranium-235 in the total uranium) fuel elements were replaced by the newly developed U{sub 3}Si{sub 2}-Al based Low Enriched Uranium (LEU, {le}20% uranium-235 in the total uranium) elements. The reactor first went critical after the core conversion was achieved in 1993, and feasibility study on the core upgrade to accommodate Boron Neutron-Captured Therapy (BNCT) was completed in 2000 [3]. The 2-MW critical reactor at RINSC which includes six beam tubes, a thermal column, a gamma-ray experimental station and two pneumatic tubes has been extensive utilized as neutron-and-photon dual source for nuclear-specific research in areas of material science, fundamental physics, biochemistry, and radiation therapy. After the core conversion along with several major system upgrade (e.g. a new 3-MW cooling tower, a large secondary piping system, a set of digitized power-level instrument), the reactor has become more compact and thus more effective to generate high beam flux in both the in-core and ex-core regions for advance research. If not limited by the manpower and operating budget in recent years, the RINSC built ''in concrete'' structure and control systems should have been systematically upgraded to a 5 Mw power facility to further enhance its experimental capability while still maintaining its safe margin as designed.

HOLDEN, N.E.,; RECINIELLO, R.N.; HU, J.-P.

2005-05-08T23:59:59.000Z

228

Photovoltaic roof heat flux  

E-Print Network [OSTI]

designs (relatively) Photovoltaic Solar P a n e l AtmosphereCALIFORNIA, SAN DIEGO Photovoltaic Roof Heat Flux A ThesisABSTRACT OF T H E THESIS Photovoltaic Roof Heat Flux by

Samady, Mezhgan Frishta

2011-01-01T23:59:59.000Z

229

Improved Design of Nuclear Reactor Control System | U.S. DOE...  

Office of Science (SC) Website

instrumentation: Improved Design of Nuclear Reactor Control System Developed at: Oak Ridge National Laboratory, Holifield Radioactive Ion Beam Facility (HRIBF) Developed...

230

Thermionic switched self-actuating reactor shutdown system  

DOE Patents [OSTI]

A self-actuating reactor shutdown system incorporating a thermionic switched electromagnetic latch arrangement which is responsive to reactor neutron flux changes and to reactor coolant temperature changes. The system is self-actuating in that the sensing thermionic device acts directly to release (scram) the control rod (absorber) without reference or signal from the main reactor plant protective and control systems. To be responsive to both temperature and neutron flux effects, two detectors are used, one responsive to reactor coolant temperatures, and the other responsive to reactor neutron flux increase. The detectors are incorporated into a thermionic diode connected electrically with an electromagnetic mechanism which under normal reactor operating conditions holds the the control rod in its ready position (exterior of the reactor core). Upon reaching either a specified temperature or neutron flux, the thermionic diode functions to short-circuit the electromagnetic mechanism causing same to lose its holding power and release the control rod, which drops into the reactor core region under gravitational force.

Barrus, Donald M. (San Jose, CA); Shires, Charles D. (San Jose, CA); Brummond, William A. (Livermore, CA)

1989-01-01T23:59:59.000Z

231

Advanced Reactors Transition Program Resource Loaded Schedule  

SciTech Connect (OSTI)

The Advanced Reactors Transition (ART) Resource Loaded Schedule (RLS) provides a cost and schedule baseline for managing the project elements within the ART Program. The Fast Flux Test Facility (FFTF) activities are delineated through the end of FY 2000, assuming continued standby. The Nuclear Energy (NE) Legacies and Plutonium Recycle Test Reactor (PRTR) activities are delineated through the end of the deactivation process. This document reflects the 1 Oct 1999 baseline.

BOWEN, W.W.

1999-11-08T23:59:59.000Z

232

Adaptive Nodal Transport Methods for Reactor Transient Analysis  

SciTech Connect (OSTI)

Develop methods for adaptively treating the angular, spatial, and time dependence of the neutron flux in reactor transient analysis. These methods were demonstrated in the DOE transport nodal code VARIANT and the US NRC spatial kinetics code, PARCS.

Thomas Downar; E. Lewis

2005-08-31T23:59:59.000Z

233

Advanced fuel fusion reactors: towards a zero-waste option  

E-Print Network [OSTI]

Low activation materials are only a partial response to the requirement of a really environmentally sound fusion reactor: another way round to tackle the problem is the reduction of the neutron flux and subsequent material ...

Zucchetti, Massimo

234

Computational neutronics analysis of TRIGA reactors during power pulsing  

E-Print Network [OSTI]

Training, Research, Isotopes, General Atomics (TRIGA) reactors have the unique capability of generating high neutron flux environments with the removal of a transient control rod, creating conditions observed in fast fission ...

Bean, Malcolm (Malcolm K.)

2011-01-01T23:59:59.000Z

235

Adaptive phase measurements for narrowband squeezed beams  

E-Print Network [OSTI]

We have previously [Phys. Rev. A 65, 043803 (2002)] analyzed adaptive measurements for estimating the continuously varying phase of a coherent beam, and a broadband squeezed beam. A real squeezed beam must have finite photon flux N and hence can be significantly squeezed only over a limited frequency range. In this paper we analyze adaptive phase measurements of this type for a realistic model of a squeezed beam. We show that, provided it is possible to suitably choose the parameters of the beam, a mean-square phase uncertainty scaling as (N/kappa)^{-5/8} is possible, where kappa is the linewidth of the beam resulting from the fluctuating phase. This is an improvement over the (N/kappa)^{-1/2} scaling found previously for coherent beams. In the experimentally realistic case where there is a limit on the maximum squeezing possible, the variance will be reduced below that for coherent beams, though the scaling is unchanged.

Dominic W. Berry; Howard M. Wiseman

2006-03-22T23:59:59.000Z

236

Nuclear reactor engineering  

SciTech Connect (OSTI)

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

Glasstone, S.; Sesonske, A.

1981-01-01T23:59:59.000Z

237

Advanced Safeguards Approaches for New Fast Reactors  

SciTech Connect (OSTI)

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

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

2007-12-15T23:59:59.000Z

238

Light Water Reactor Sustainability  

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

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

239

Light Water Reactor Sustainability  

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

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

240

An Account of Oak Ridge National Laboratory's Thirteen Research Reactors  

SciTech Connect (OSTI)

The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

Rosenthal, Murray Wilford [ORNL

2009-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "flux beam reactor" 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

Advanced Neutrino Sources (Neutrino Factories and Beta Beams)  

E-Print Network [OSTI]

Advanced Neutrino Sources (Neutrino Factories and Beta Beams) · Design · R&D Status · Remaining R Meeting February, 2008 page 1 #12;· The stored beam properties & decay kinematics are well known uncertainties on neutrino flux & spectra are small PRECISION · Initial beams are flavor "pure" (BB) or "tagged

242

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

243

Bioconversion reactor  

DOE Patents [OSTI]

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

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

1992-01-01T23:59:59.000Z

244

Operational control of boiling water reactor stability  

SciTech Connect (OSTI)

Boiling water reactor cores are susceptible to instabilities, which generate power oscillations. Specific reactor operating practices can provide a mechanism for control of the instability phenomenon. An axial separation of the core into a single-phase region and a two-phase region resolves the influence of axial flux shapes on core stability. This separation provides the means to derive a core stability control that ensures significant reactor stability margin. The control is achieved by maintaining the core average bulk coolant saturation elevation above a predetermined axial plane. The control can be reliably and efficiently implemented during reactor operations. Analysis demonstrates that variations in parameters important to stability have only secondary influences on stability margin when the control is in effect. Actual plant experience with a large commercial boiling water reactor confirms the capabilities of this stability control in an operational setting.

Mowry, C.M. [PECO Energy, Wayne, PA (United States); Nir, I. [Entergy Operations, Jackson, MS (United States); Newkirk, D.W. [GE Nuclear Energy, San Jose, CA (United States)

1995-03-01T23:59:59.000Z

245

Self-actuating reactor shutdown system  

DOE Patents [OSTI]

A control system for the automatic or self-actuated shutdown or "scram" of a nuclear reactor. The system is capable of initiating scram insertion by a signal from the plant protection system or by independent action directly sensing reactor conditions of low-flow or over-power. Self-actuation due to a loss of reactor coolant flow results from a decrease of pressure differential between the upper and lower ends of an absorber element. When the force due to this differential falls below the weight of the element, the element will fall by gravitational force to scram the reactor. Self-actuation due to high neutron flux is accomplished via a valve controlled by an electromagnet and a thermionic diode. In a reactor over-power, the diode will be heated to a change of state causing the electromagnet to be shorted thereby actuating the valve which provides the changed flow and pressure conditions required for scramming the absorber element.

Barrus, Donald M. (San Jose, CA); Brummond, Willian A (Livermore, CA); Peterson, Leslie F. (Danville, CA)

1988-01-01T23:59:59.000Z

246

Atmospheric Neutrino Fluxes  

E-Print Network [OSTI]

Starting with an historical review, I summarize the status of calculations of the flux of atmospheric neutrinos and how they compare to measurements.

Thomas K. Gaisser

2005-02-18T23:59:59.000Z

247

Pulsed ion beam source  

DOE Patents [OSTI]

An improved pulsed ion beam source is disclosed having a new biasing circuit for the fast magnetic field. This circuit provides for an initial negative bias for the field created by the fast coils in the ion beam source which pre-ionize the gas in the source, ionize the gas and deliver the gas to the proper position in the accelerating gap between the anode and cathode assemblies in the ion beam source. The initial negative bias improves the interaction between the location of the nulls in the composite magnetic field in the ion beam source and the position of the gas for pre-ionization and ionization into the plasma as well as final positioning of the plasma in the accelerating gap. Improvements to the construction of the flux excluders in the anode assembly are also accomplished by fabricating them as layered structures with a high melting point, low conductivity material on the outsides with a high conductivity material in the center. 12 figs.

Greenly, J.B.

1997-08-12T23:59:59.000Z

248

Neutral beam dump with cathodic arc titanium gettering  

SciTech Connect (OSTI)

An incomplete neutral beam capture can degrade the plasma performance in neutral beam driven plasma machines. The beam dumps mitigating the shine-through beam recycling must entrap and retain large particle loads while maintaining the beam-exposed surfaces clean of the residual impurities. The cathodic arc gettering, which provides high evaporation rate coupled with a fast time response, is a powerful and versatile technique for depositing clean getter films in vacuum. A compact neutral beam dump utilizing the titanium arc gettering was developed for a field-reversed configuration plasma sustained by 1 MW, 20-40 keV neutral hydrogen beams. The titanium evaporator features a new improved design. The beam dump is capable of handling large pulsed gas loads, has a high sorption capacity, and is robust and reliable. With the beam particle flux density of 5 x 10{sup 17} H/(cm{sup 2}s) sustained for 3-10 ms, the beam recycling coefficient, defined as twice the ratio of the hydrogen molecular flux leaving the beam dump to the incident flux of high-energy neutral atoms, is {approx}0.7. The use of the beam dump allows us to significantly reduce the recycling of the shine-through neutral beam as well as to improve the vacuum conditions in the machine.

Smirnov, A.; Korepanov, S. A.; Putvinski, S. [Tri Alpha Energy Inc., Rancho Santa Margarita, California 92688 (United States); Krivenko, A. S.; Murakhtin, S. V.; Savkin, V. Ya. [Budker Institute of Nuclear Physics, Novosibirsk 630090 (Russian Federation)

2011-03-15T23:59:59.000Z

249

PHELIX for flux compression studies  

SciTech Connect (OSTI)

PHELIX (Precision High Energy-density Liner Implosion eXperiment) is a concept for studying electromagnetic implosions using proton radiography. This approach requires a portable pulsed power and liner implosion apparatus that can be operated in conjunction with an 800 MeV proton beam at the Los Alamos Neutron Science Center. The high resolution (< 100 micron) provided by proton radiography combined with similar precision of liner implosions driven electromagnetically can permit close comparisons of multi-frame experimental data and numerical simulations within a single dynamic event. To achieve a portable implosion system for use at high energy-density in a proton laboratory area requires sub-megajoule energies applied to implosions only a few cms in radial and axial dimension. The associated inductance changes are therefore relatively modest, so a current step-up transformer arrangement is employed to avoid excessive loss to parasitic inductances that are relatively large for low-energy banks comprising only several capacitors and switches. We describe the design, construction and operation of the PHELIX system and discuss application to liner-driven, magnetic flux compression experiments. For the latter, the ability of strong magnetic fields to deflect the proton beam may offer a novel technique for measurement of field distributions near perturbed surfaces.

Turchi, Peter J [Los Alamos National Laboratory; Rousculp, Christopher L [Los Alamos National Laboratory; Reinovsky, Robert E [Los Alamos National Laboratory; Reass, William A [Los Alamos National Laboratory; Griego, Jeffrey R [Los Alamos National Laboratory; Oro, David M [Los Alamos National Laboratory; Merrill, Frank E [Los Alamos National Laboratory

2010-06-28T23:59:59.000Z

250

Photoneutron effects on pulse reactor kinetics for the Annular Core Research Reactor (ACRR).  

SciTech Connect (OSTI)

The Annular Core Research Reactor (ACRR) is a swimming-pool type pulsed reactor that maintains an epithermal neutron flux and a nine-inch diameter central dry cavity. One of its uses is neutron and gamma-ray irradiation damage studies on electronic components under transient reactor power conditions. In analyzing the experimental results, careful attention must be paid to the kinetics associated with the reactor to ensure that the transient behavior of the electronic device is understood. Since the ACRR fuel maintains a substantial amount of beryllium, copious quantities of photoneutrons are produced that can significantly alter the expected behavior of the reactor power, especially following a reactor pulse. In order to understand these photoneutron effects on the reactor kinetics, the KIFLE transient reactor-analysis code was modified to include the photoneutron groups associated with the beryllium. The time-dependent behavior of the reactor power was analyzed for small and large pulses, assuming several initial conditions including following several pulses during the day, and following a long steady-state power run. The results indicate that, for these types of initial conditions, the photoneutron contribution to the reactor pulse energy can have a few to tens of percent effect.

Parma, Edward J., Jr.

2009-06-01T23:59:59.000Z

251

The measurement of absolute thermal neutron flux using liquid scintillation counting techniques  

E-Print Network [OSTI]

was computed as the square root of the sum of the squares of the individual errors . The flux at the same location in the core and at the same reactor power level was measured by the conventional technique of gold foil 34 activation. This measurement... back to 1932 when the neutron was discovered by Chadwick. With the advent of the nuclear reactor in 1942 the problem of absolute neutron flux determination became increasingly important. Since the operating power of a thermal reactor is directly...

Walker, Jack Vernon

2012-06-07T23:59:59.000Z

252

Advanced Test Reactor National Scientific User Facility  

SciTech Connect (OSTI)

The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

2011-08-01T23:59:59.000Z

253

Solar Thermal Reactor Materials Characterization  

SciTech Connect (OSTI)

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

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

2008-03-01T23:59:59.000Z

254

Hybrid adsorptive membrane reactor  

DOE Patents [OSTI]

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

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

2011-03-01T23:59:59.000Z

255

Creation of a neutrino laboratory for search for sterile neutrino at SM-3 reactor  

E-Print Network [OSTI]

In connection with the question of possible existence of sterile neutrino the laboratory on the basis of SM-3 reactor was created to search for oscillations of reactor antineutrino. A prototype of a neutrino detector with scintillator volume of 400 l can be moved at the distance of 6-11 m from the reactor core. The measurements of background conditions have been made. It is shown that the main experimental problem is associated with cosmic radiation background. Test measurements of dependence of a reactor antineutrino flux on the distance from a reactor core have been made. The prospects of search for oscillations of reactor antineutrino at short distances are discussed.

Serebrov, A P; Samoylov, R M; Fomin, A K; Zinoviev, V G; Neustroev, P V; Golovtsov, V L; Gruzinsky, N V; Solovey, V A; Cherniy, A V; Zherebtsov, O M; Martemyanov, V P; Zinoev, V G; Tarasenkov, V G; Aleshin, V I; Petelin, A L; Pavlov, S V; Izhutov, A L; Sazontov, S A; Ryazanov, D K; Gromov, M O; Afanasiev, V V; Matrosov, L N; Matrosova, M Yu

2015-01-01T23:59:59.000Z

256

High Flux Isotope Reactor quarterly report, January-March 1986  

SciTech Connect (OSTI)

Four routine cycles of operation were completed during the first quarter. Four scheduled end-of-cycle shutdowns and two unscheduled shutdowns resulted in an on-stream time of 86.9%. An unscheduled control plate and control cylinder replacement was performed after tantalum contamination in the primary coolant system was traced to the inner cylinder.

Corbett, B.L.; Farrar, M.B.

1986-06-01T23:59:59.000Z

257

High Flux Isotope Reactor | Neutron Science at ORNL  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary)morphinanInformation InInformation InExplosion Monitoring:Home| Visitors|Upcoming EventsFriday, May 1, 2015High

258

Solar Magnetic Flux Ropes  

E-Print Network [OSTI]

The most probable initial magnetic configuration of a CME is a flux rope consisting of twisted field lines which fill the whole volume of a dark coronal cavity. The flux ropes can be in stable equilibrium in the coronal magnetic field for weeks and even months, but suddenly they loose their stability and erupt with high speed. Their transition to the unstable phase depends on the parameters of the flux rope (i.e., total electric current, twist, mass loading etc.), as well as on the properties of the ambient coronal magnetic field. One of the major governing factors is the vertical gradient of the coronal magnetic field which is estimated as decay index (n). Cold dense prominence material can be collected in the lower parts of the helical flux tubes. Filaments are therefore good tracers of the flux ropes in the corona, which become visible long before the beginning of the eruption. The perspectives of the filament eruptions and following CMEs can be estimated by the comparison of observed filament heights with...

Filippov, Boris; Srivastava, Abhishek K; Uddin, Wahab

2015-01-01T23:59:59.000Z

259

Nuclear reactor engineering  

SciTech Connect (OSTI)

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

Glasstone, S.; Sesonske, A.

1982-07-01T23:59:59.000Z

260

PHYSICS AND SAFETY ANALYSIS FOR THE NIST RESEARCH REACTOR.  

SciTech Connect (OSTI)

Detailed reactor physics and safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analyses provide an update to the Final Safety Analysis Report (FSAR) and employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron and photon transport calculations were performed with the MCNP code to determine the safety parameters for the NBSR. The core depletion and determination of the fuel compositions were performed with MONTEBURNS. MCNP calculations were performed to determine the beginning, middle, and end-of-cycle power distributions, moderator temperature coefficient, and shim safety arm, beam tube and void reactivity worths. The calculational model included a plate-by-plate description of each fuel assembly, axial mid-plane water gap, beam tubes and the tubular geometry of the shim safety arms. The time-dependent analysis of the primary loop was determined with a RELAP5 transient analysis model that includes the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. The statistical analysis used to assure protection from critical heat flux (CHF) was performed using a Monte Carlo simulation of the uncertainties contributing to the CHF calculation. The power distributions used to determine the local fuel conditions and margin to CHF were determined with MCNP. Evaluations were performed for the following accidents: (1) the control rod withdrawal startup accident, (2) the maximum reactivity insertion accident, (3) loss-of-flow resulting from loss of electrical power, (4) loss-of-flow resulting from a primary pump seizure, (5) loss-of-flow resulting from inadvertent throttling of a flow control valve, (6) loss-of-flow resulting from failure of both shutdown cooling pumps and (7) misloading of a fuel element. In both the startup and maximum reactivity insertion accidents, the core power transient is terminated by a reactor trip at 26 MW. The calculations show that both the peak reactor power and the excursion energy depend on the negative reactivity insertion from reactor trip. In one of the loss-of-flow accidents offsite electrical power is assumed lost to the three operating primary pumps. A slightly delayed reactor scram is initiated as a result of primary flow coast down. The RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur, because of the momentum of the coolant flowing through the fuel channels and the negative scram reactivity insertion. For both the primary pump seizure and inadvertent throttling of a flow control valve, the RELAP5 analyses indicate that the reduction in power following the trip is sufficient to ensure that there is adequate margin to CHF and that the fuel cladding does not fail. The analysis of the loss-of-flow accident in the extremely unlikely case where both shutdown pumps fail, shows that the cooling provided by the D{sub 2}O is sufficient to ensure the cladding does not fail. The power distributions were examined for a set of fuel misloadings in which a fresh fuel element is moved from a peripheral low-reactivity location to a central high-reactivity location. The calculations show that there is adequate margin to CHF and the cladding does not fail. An additional analysis was performed to simulate the operation at low power (500 kW) without forced flow cooling. The result indicates that natural convection cooling is adequate for operation of the NBSR at a power level of 500 kW.

CHENG,L.HANSON,A.DIAMOND,D.XU,J.CAREW,J.RORER,D.

2004-03-31T23:59:59.000Z

Note: This page contains sample records for the topic "flux beam reactor" 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

Beam History  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to someone6 M. Babzien, I. Ben-Zvi, P. StudyBeam History PrintBeam History

262

Iterative methods for solving nonlinear problems of nuclear reactor criticality  

SciTech Connect (OSTI)

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

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

2012-12-15T23:59:59.000Z

263

Control system for a small fission reactor  

DOE Patents [OSTI]

A system for controlling the reactivity of a small fission reactor includes an elongated, flexible hollow tube in the general form of a helical coiled spring axially positioned around and outside of the reactor vessel in an annular space between the reactor vessel and a surrounding cylindrical-shaped neutron reflector. A neutron absorbing material is provided within the hollow tube with the rate of the reaction controlled by the extension and compression of the hollow tube, e.g., extension of the tube increases reactivity while its compression reduces reactivity, in varying the amount of neutron absorbing material disposed between the reactor vessel and the neutron reflector. Conventional mechanical displacement means may be employed to control the coil density of the hollow tube as desired. In another embodiment, a plurality of flexible hollow tubes each containing a neutron absorber are positioned adjacent to one another in spaced relation around the periphery of the reactor vessel and inside the outer neutron reflector with reactivity controlled by the extension and compression of all or some of the coiled hollow tubes. Yet another embodiment of the invention envisions the neutron reflector in the form of an expandable coil spring positioned in an annular space between the reactor vessel and an outer neutron absorbing structure for controlling the neutron flux reflected back into the reactor vessel.

Burelbach, James P. (Glen Ellyn, IL); Kann, William J. (Park Ridge, IL); Saiveau, James G. (Hickory Hills, IL)

1986-01-01T23:59:59.000Z

264

Beams 92: Proceedings. Volume 2, Ion beams, electron beams, diagnostics  

SciTech Connect (OSTI)

This report contains papers on the following topics. Ion beam papers; electron beam papers; and these papers have been indexed separately elsewhere.

Mosher, D.; Cooperstein, G. [eds.] [Naval Research Lab., Washington, DC (United States)] [eds.; Naval Research Lab., Washington, DC (United States)

1993-12-31T23:59:59.000Z

265

Volkov solution for two laser beams and ITER  

E-Print Network [OSTI]

We find the solution of the Dirac equation for two plane waves (laser beams) and we determine the modified Compton formula for the scattering of two photons on an alectron. The practical meaning of the two laser beams is, that two laser beams impinging on a targed which is constituted from material in the form of a foam, can replace 100-200 laser beams impinging on a normal targed. It means that the nuclear fusion with two laser beams is realistic in combination with the nuclear reactor such as ITER.

Miroslav Pardy

2005-07-12T23:59:59.000Z

266

Impact of conversion to mixed-oxide fuels on reactor structural components  

SciTech Connect (OSTI)

The use of mixed-oxide (MOX) fuel to replace conventional uranium fuel in commercial light-water power reactors will result in an increase in the neutron flux. The impact of the higher flux on the structural integrity of reactor structural components must be evaluated. This report briefly reviews the effects of radiation on the mechanical properties of metals. Aging degradation studies and reactor operating experience provide a basis for determining the areas where conversion to MOX fuels has the potential to impact the structural integrity of reactor components.

Yahr, G.T.

1997-04-01T23:59:59.000Z

267

Investigation of downward facing critical heat flux with water-based nanofluids for In-Vessel Retention applications  

E-Print Network [OSTI]

In-Vessel Retention ("IVR") is a severe accident management strategy that is power limiting to the Westinghouse AP1000 due to critical heat flux ("CHF") at the outer surface of the reactor vessel. Increasing the CHF level ...

DeWitt, Gregory L

2011-01-01T23:59:59.000Z

268

Radiative Flux Analysis  

DOE Data Explorer [Office of Scientific and Technical Information (OSTI)]

The Radiative Flux Analysis is a technique for using surface broadband radiation measurements for detecting periods of clear (i.e. cloudless) skies, and using the detected clear-sky data to fit functions which are then used to produce continuous clear-sky estimates. The clear-sky estimates and measurements are then used in various ways to infer cloud macrophysical properties.

Long, Chuck [NOAA

269

Fast neutron fluxes in pressure vessels using Monte Carlo methods  

SciTech Connect (OSTI)

The objective of this project is to determine the feasibility of calculating the fast neutron flux in the pressure vessel of a pressurized water reactor by Monte Carlo methods. Neutron reactions reduce the ductility of the steel and thus limit the useful life of this important reactor component. This work was performed for Virginia Power (VEPCO). VIM is a continuous-energy Monte Carlo code which provides a versatile geometrical capability and a neutron physics data base closely representing the EDNF/B-IV data from which it was derived.

Edlund, M.C.; Thomas, J.R.

1986-01-01T23:59:59.000Z

270

Fast Flux Test Facility final safety analysis report. Amendment 73  

SciTech Connect (OSTI)

This report provides Final Safety Analysis Report (FSAR) Amendment 73 for incorporation into the Fast Flux Test Facility (FFTR) FSAR set. This page change incorporates Engineering Change Notices (ECNs) issued subsequent to Amendment 72 and approved for incorparoration before May 6, 1993. These changes include: Chapter 3, design criteria structures, equipment, and systems; chapter 5B, reactor coolant system; chapter 7, instrumentation and control systems; chapter 9, auxiliary systems; chapter 11, reactor refueling system; chapter 12, radiation protection and waste management; chapter 13, conduct of operations; chapter 17, technical specifications; chapter 20, FFTF criticality specifications; appendix C, local fuel failure events; and appendix Fl, operation at 680{degrees}F inlet temperature.

Gantt, D.A.

1993-08-01T23:59:59.000Z

271

Beam current controller for laser ion source  

DOE Patents [OSTI]

The present invention relates to the design and use of an ion source with a rapid beam current controller for experimental and medicinal purposes. More particularly, the present invention relates to the design and use of a laser ion source with a magnetic field applied to confine a plasma flux caused by laser ablation.

Okamura, Masahiro

2014-10-28T23:59:59.000Z

272

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

273

SRS Small Modular Reactors  

SciTech Connect (OSTI)

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

None

2012-04-27T23:59:59.000Z

274

SRS Small Modular Reactors  

ScienceCinema (OSTI)

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

None

2014-05-21T23:59:59.000Z

275

Boosted Fast Flux Loop Final Report  

SciTech Connect (OSTI)

The Boosted Fast Flux Loop (BFFL) project was initiated to determine basic feasibility of designing, constructing, and installing in a host irradiation facility, an experimental vehicle that can replicate with reasonable fidelity the fast-flux test environment needed for fuels and materials irradiation testing for advanced reactor concepts. Originally called the Gas Test Loop (GTL) project, the activity included (1) determination of requirements that must be met for the GTL to be responsive to potential users, (2) a survey of nuclear facilities that may successfully host the GTL, (3) conceptualizing designs for hardware that can support the needed environments for neutron flux intensity and energy spectrum, atmosphere, flow, etc. needed by the experimenters, and (4) examining other aspects of such a system, such as waste generation and disposal, environmental concerns, needs for additional infrastructure, and requirements for interfacing with the host facility. A revised project plan included requesting an interim decision, termed CD-1A, that had objectives of' establishing the site for the project at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL), deferring the CD 1 application, and authorizing a research program that would resolve the most pressing technical questions regarding GTL feasibility, including issues relating to the use of booster fuel in the ATR. Major research tasks were (1) hydraulic testing to establish flow conditions through the booster fuel, (2) mini-plate irradiation tests and post-irradiation examination to alleviate concerns over corrosion at the high heat fluxes planned, (3) development and demonstration of booster fuel fabrication techniques, and (4) a review of the impact of the GTL on the ATR safety basis. A revised cooling concept for the apparatus was conceptualized, which resulted in renaming the project to the BFFL. Before the subsequent CD-1 approval request could be made, a decision was made in April 2006 that further funding for the project would be suspended. Remaining funds have been used to prepare and irradiate mini-plates of the proposed booster fuel. The current baseline design is for a set of three test positions inside an in-pile tube with a thermal neutron absorber and heat sink made of aluminum mixed with hafnium. Operating the ATR at power levels needed to achieve the required fast flux will result in an estimated increase in ATR fuel consumption between 15 and 20% above present rates and a reduction in the time between fuel replacements. Preliminary safety analyses conducted have indicted safe operation of the ATR with the GTL under normal, abnormal, and postulated accident conditions. More comprehensive analyses are needed.

Boosted Fast Flux Loop Project Staff

2009-09-01T23:59:59.000Z

276

Optimal beam pattern to maximize inclusion residence time in an electron beam melting hearth  

SciTech Connect (OSTI)

Approximate probabilities of inclusion survival through an electron beam melting hearth are computed from nitride dissolution rates, flotation velocities, and residence times. Dissolution rates were determined by measuring shrinkage rates of pure TiN and nitrided sponge in small pools of molten titanium in an electron beam melting hearth. Flotation velocities were calculated using correlations for fluid flow around spheres, and show that particles sink or float unless their densities are extremely close to that of molten titanium. Flow field characteristics which lead to effective inclusion removal are discussed in terms of heat flux pattern required to produce them, based on the electron beam`s unique ability to impart a nearly arbitrary heat flux pattern to the melt surface.

Powell, A.; Pal, U. [Massachusetts Inst. of Tech., Cambridge, MA (United States); Avyle, J. van den [Sandia National Labs., Albuquerque, NM (United States)

1997-02-01T23:59:59.000Z

277

Characterization of X-ray generator beam profiles.  

SciTech Connect (OSTI)

T to compute the radiography properties of various materials, the flux profiles of X-ray sources must be characterized. This report describes the characterization of X-ray beam profiles from a Kimtron industrial 450 kVp radiography system with a Comet MXC-45 HP/11 bipolar oil-cooled X-ray tube. The empirical method described here uses a detector response function to derive photon flux profiles based on data collected with a small cadmium telluride detector. The flux profiles are then reduced to a simple parametric form that enables computation of beam profiles for arbitrary accelerator energies.

Mitchell, Dean J; Harding, Lee T.; Thoreson, Gregory G.; Theisen, Lisa Anne; Parmeter, John Ethan; Thompson, Kyle Richard

2013-07-01T23:59:59.000Z

278

Nuclear reactor  

DOE Patents [OSTI]

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

Thomson, Wallace B. (Severna Park, MD)

2004-03-16T23:59:59.000Z

279

Accelerator beam profile analyzer  

DOE Patents [OSTI]

A beam profile analyzer employing sector or quadrant plates each servo controlled to outline the edge of a beam.

Godel, Julius B. (Bayport, NY); Guillaume, Marcel (Grivegnee, BE); Lambrecht, Richard M. (East Quogue, NY); Withnell, Ronald (East Setauket, NY)

1976-01-01T23:59:59.000Z

280

PHYSICS AND SAFETY ANALYSIS FOR THE NIST RESEARCH REACTOR.  

SciTech Connect (OSTI)

Detailed reactor physics and safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analyses provide an update to the Final Safety Analysis Report (FSAR) and employ state-of-the-art calculational methods. Three-dimensional MCNP Monte Carlo neutron and photon transport calculations were performed to determine the safety parameters for the NBSR. The core depletion and determination of the fuel compositions were performed with MONTEBURNS. MCNP calculations were performed to determine the beginning, middle, and end-of-cycle power distributions, moderator temperature coefficient, and shim arm, beam tube and void reactivity worths. The calculational model included a plate-by-plate description of each fuel assembly, axial mid-plane water gap, beam tubes and the tubular geometry of the shim arms. The time-dependent analysis of the primary loop was determined with a RELAP5 transient analysis model including the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. The statistical analysis used to assure protection from critical heat flux (CHF) was performed using a Monte Carlo simulation of the uncertainties contributing to the CHF calculation. The power distributions used to determine the local fuel conditions and margin to CHF were determined with MCNP. Evaluations were performed for the following accidents: (1) the control rod withdrawal startup accident, (2) the maximum reactivity insertion accident, (3) loss-of-flow resulting from loss of electrical power, (4) loss-of-flow resulting from a primary pump seizure, (5) loss-of-flow resulting from inadvertent throttling of a flow control valve, (6) loss-of-flow resulting from failure of both shutdown cooling pumps and (7) misloading of a fuel element. In both the startup and maximum reactivity insertion accidents, the core power transient is terminated by a reactor trip at 30 MW. The calculations show that both the peak reactor power and the excursion energy depend on the negative reactivity insertion from reactor trip. Two cases were considered for loss of electrical power. In the first case offsite power is lost, resulting in an immediate scram caused by loss of power to the control rod system. In the second case power is lost to only the three operating primary pumps, resulting in a slightly delayed scram when loss-of-flow is detected as the pumps coast down. In both instances, RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur, because of the momentum of the coolant flowing through the fuel channels and the negative scram reactivity insertion. For both the primary pump seizure and inadvertent throttling of a flow control valve, the RELAP5 analyses indicate that the reduction in power following the trip is sufficient to ensure that there is adequate margin to CHF and the fuel cladding does not fail. The analysis of the loss-of-flow accident in the extremely unlikely case where both shutdown pumps fail shows that the cooling provided by the D{sub 2}O is sufficient to ensure the cladding does not fail. The power distributions were examined for a set of fuel misloadings in which a fresh fuel element is moved from a peripheral low-reactivity location to a central high-reactivity location. The calculations show that there is adequate margin to CHF and the cladding does not fail.

CAREW,J.CHENG,L.HANSON,AXU,J.RORER,D.DIAMOND,D.

2003-08-26T23:59:59.000Z

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


281

Magnetic Flux Dynamics in Horizontally Cooled Superconducting Cavities  

E-Print Network [OSTI]

Previous studies on magnetic flux expulsion as a function of cooling details have been performed for superconducting niobium cavities with the cavity beam axis placed parallel respect to the helium cooling flow, and findings showed that for sufficient cooling thermogradients all magnetic flux could be expelled and very low residual resistance could be achieved. In this paper we investigate the flux trapping and its impact on radio frequency surface resistance when the resonators are positioned perpendicularly to the helium cooling flow, which is representative of how superconducting radio-frequency (SRF) cavities are cooled in an accelerator. We also extend the studies to different directions of applied magnetic field surrounding the resonator. Results show that in the cavity horizontal configuration there is a different impact of the various field components on the final surface resistance, and that several parameters have to be considered to understand flux dynamics. A newly discovered phenomenon of concent...

Martinello, M; Grassellino, A; Crawford, A C; Melnychuk, O; Romanenko, A; Sergatkov, D A

2015-01-01T23:59:59.000Z

282

Optical heat flux gauge  

DOE Patents [OSTI]

A heat flux gauge is disclosed comprising first and second thermographic phosphor layers separated by a layer of a thermal insulator, wherein each thermographic layer comprises a plurality of respective thermographic sensors in a juxtaposed relationship with respect to each other. The gauge may be mounted on a surface with the first thermographic phosphor in contact with the surface. A light source is directed at the gauge, causing the phosphors to luminesce. The luminescence produced by the phosphors is collected and its spectra analyzed in order to determine the heat flux on the surface. First and second phosphor layers must be different materials to assure that the spectral lines collected will be distinguishable. 9 figures.

Noel, B.W.; Borella, H.M.; Cates, M.R.; Turley, W.D.; MacArthur, C.D.; Cala, G.C.

1991-04-09T23:59:59.000Z

283

Portable radiography system using a relativistic electron beam  

DOE Patents [OSTI]

A portable radiographic generator is provided with an explosive magnetic flux compression generator producing the high voltage necessary to generate a relativistic electron beam. The relativistic electron beam is provided with target materials which generates the desired radiographic pulse. The magnetic flux compression generator may require at least two conventional explosively driven generators in series to obtain a desired output voltage of at least 1 MV. The cathode and anode configuration of the diode are selected to provide a switching action wherein a high impedance load is presented to the magnetic flux compression generator when the high voltage is being generated, and thereafter switching to a low impedance load to generate the relativistic electron beam. Magnetic flux compression generators can be explosively driven and provided in a relatively compact, portable form for use with the relativistic x-ray equipment. 8 figs.

Hoeberling, R.F.

1987-09-22T23:59:59.000Z

284

The reactor physics of the Massachusetts Institute of Technology reactor redesign  

E-Print Network [OSTI]

An H20 cooled compact MITR-II core, reflected by D20 has been designed for the MITR to increase the reflector thermal neutron flux at tips of beam ports by a factor of 3 or better, without changing the operating power level ...

Addae, Andrews Kwasi

1970-01-01T23:59:59.000Z

285

BDDR, a new CEA technological and operating reactor database  

SciTech Connect (OSTI)

The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a unique repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)

Soldevilla, M.; Salmons, S.; Espinosa, B. [CEA-Saclay, CEA/DEN/DANS/DM2S/SERMA, 91191 Gif-sur-Yvette (France); Clanet, M.; Boudin, X. [CEA-Bruyeres-le-Chatel, 91297 Arpajon (France)

2013-07-01T23:59:59.000Z

286

Search for Neutrino Oscillations at the Palo Verde Nuclear Reactors  

E-Print Network [OSTI]

We report on the initial results from a measurement of the anti-neutrino flux and spectrum at a distance of about 800 m from the three reactors of the Palo Verde Nuclear Generating Station using a segmented gadolinium-loaded scintillation detector. We find that the anti-neutrino flux agrees with that predicted in the absence of oscillations to better than 5%, excluding at 90% CL $\\rm\\bar\

F. Boehm; J. Busenitz; B. Cook; G. Gratta; H. Henrikson; J. Kornis; D. Lawrence; K. B. Lee; K. McKinny; L. Miller; V. Novikov; A. Piepke; B. Ritchie; D. Tracy; P. Vogel; Y-F. Wang; J. Wolf

1999-12-22T23:59:59.000Z

287

Radiation effects on reactor pressure vessel supports  

SciTech Connect (OSTI)

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

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

1996-05-01T23:59:59.000Z

288

Beam History  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to someone6 M. Babzien, I. Ben-Zvi, P. StudyBeam History Print Beamline

289

Beam History  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to someone6 M. Babzien, I. Ben-Zvi, P. StudyBeam History Print

290

Beam Transport  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to someone6 M. Babzien, I. Ben-Zvi, P. StudyBeam History

291

Beam Status  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary)morphinanInformation InInformationCenterResearch HighlightsToolsBES ReportsExperimentBasicBeam Status Print mA

292

Micro -Thermonuclear AB-Reactors for Aerospace  

E-Print Network [OSTI]

The author offers several innovations that he first suggested publicly early in 1983 for the AB multi-reflex engine, space propulsion, getting energy from plasma, etc. (see: A. Bolonkin, Non-Rocket Space Launch and Flight, Elsevier, London, 2006, Chapters 12, 3A). It is the micro-thermonuclear AB-Reactors. That is new micro-thermonuclear reactor with very small fuel pellet that uses plasma confinement generated by multi-reflection of laser beam or its own magnetic field. The Lawson criterion increases by hundreds of times. The author also suggests a new method of heating the power-making fuel pellet by outer electric current as well as new direct method of transformation of ion kinetic energy into harvestable electricity. These offered innovations dramatically decrease the size, weight and cost of thermonuclear reactor, installation, propulsion system and electric generator. Non-industrial countries can produce these researches and constructions. Currently, the author is researching the efficiency of these innovations for two types of the micro-thermonuclear reactors: multi-reflection reactor (ICF) and self-magnetic reactor (MCF).

Alexander Bolonkin

2007-01-08T23:59:59.000Z

293

Divertor Heat Flux Mitigation in the National Spherical Torus Experiment  

SciTech Connect (OSTI)

Steady-state handling of divertor heat flux is a critical issue for both ITER and spherical torus-based devices with compact high power density divertors. Significant reduction of heat flux to the divertor plate has been achieved simultaneously with favorable core and pedestal confinement and stability properties in a highly-shaped lower single null configuration in the National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40, 557 2000] using high magnetic flux expansion at the divertor strike point and the radiative divertor technique. A partial detachment of the outer strike point was achieved with divertor deuterium injection leading to peak flux reduction from 4-6 MW m{sup -2} to 0.5-2 MW m{sup -2} in small-ELM 0.8-1.0 MA, 4-6 MW neutral beam injection-heated H-mode discharges. A self-consistent picture of outer strike point partial detachment was evident from divertor heat flux profiles and recombination, particle flux and neutral pressure measurements. Analytic scrape-off layer parallel transport models were used for interpretation of NSTX detachment experiments. The modeling showed that the observed peak heat flux reduction and detachment are possible with high radiated power and momentum loss fractions, achievable with divertor gas injection, and nearly impossible to achieve with main electron density, divertor neutral density or recombination increases alone.

Soukhanovskii, V A; Maingi, R; Gates, D A; Menard, J E; Paul, S F; Raman, R; Roquemore, A L; Bell, M G; Bell, R E; Boedo, J A; Bush, C E; Kaita, R; Kugel, H W; LeBlanc, B P; Mueller, D

2008-08-04T23:59:59.000Z

294

Beam geometry selection using sequential beam addition  

SciTech Connect (OSTI)

Purpose: The selection of optimal beam geometry has been of interest since the inception of conformal radiotherapy. The authors report on sequential beam addition, a simple beam geometry selection method, for intensity modulated radiation therapy. Methods: The sequential beam addition algorithm (SBA) requires definition of an objective function (score) and a set of candidate beam geometries (pool). In the first iteration, the optimal score is determined for each beam in the pool and the beam with the best score selected. In the next iteration, the optimal score is calculated for each beam remaining in the pool combined with the beam selected in the first iteration, and the best scoring beam is selected. The process is repeated until the desired number of beams is reached. The authors selected three treatment sites, breast, lung, and brain, and determined beam arrangements for up to 11 beams from a pool comprised of 25 equiangular transverse beams. For the brain, arrangements were additionally selected from a pool of 22 noncoplanar beams. Scores were determined for geometries comprised equiangular transverse beams (EQA), as well as two tangential beams for the breast case. Results: In all cases, SBA resulted in scores superior to EQA. The breast case had the strongest dependence on beam geometry, for which only the 7-beam EQA geometry had a score better than the two tangential beams, whereas all SBA geometries with more than two beams were superior. In the lung case, EQA and SBA scores monotonically improved with increasing number of beams; however, SBA required fewer beams to achieve scores equivalent to EQA. For the brain case, SBA with a coplanar pool was equivalent to EQA, while the noncoplanar pool resulted in slightly better scores; however, the dose-volume histograms demonstrated that the differences were not clinically significant. Conclusions: For situations in which beam geometry has a significant effect on the objective function, SBA can identify arrangements equivalent to equiangular geometries but using fewer beams. Furthermore, SBA provides the value of the objective function as the number of beams is increased, allowing the planner to select the minimal beam number that achieves the clinical goals. The method is simple to implement and could readily be incorporated into an existing optimization system.

Popple, Richard A., E-mail: rpopple@uabmc.edu; Brezovich, Ivan A.; Fiveash, John B. [Department of Radiation Oncology, The University of Alabama at Birmingham, 1720 2nd Avenue South, Birmingham, Alabama 35294 (United States)] [Department of Radiation Oncology, The University of Alabama at Birmingham, 1720 2nd Avenue South, Birmingham, Alabama 35294 (United States)

2014-05-15T23:59:59.000Z

295

Beam-Bem interactions  

SciTech Connect (OSTI)

In high energy storage-ring colliders, the nonlinear effect arising from beam-beam interactions is a major source that leads to the emittance growth, the reduction of beam life time, and limits the collider luminosity. In this paper, two models of beam-beam interactions are introduced, which are weak-strong and strong-strong beam-beam interactions. In addition, space-charge model is introduced.

Kim, Hyung Jin; /Fermilab

2011-12-01T23:59:59.000Z

296

FAST FLUX TEST FACILITY (FFTF) A HISTORY OF SAFETY & OPERATIONAL EXCELLENCE  

SciTech Connect (OSTI)

The Fast Flux Test Facility (FFTF) is a 400-megawatt (thermal) sodium-cooled, high temperature, fast neutron flux, loop-type test reactor. The facility was constructed to support development and testing of fuels, materials and equipment for the Liquid Metal Fast Breeder Reactor program. FFTF began operation in 1980 and over the next 10 years demonstrated its versatility to perform experiments and missions far beyond the original intent of its designers. The reactor had several distinctive features including its size, flux, core design, extensive instrumentation, and test features that enabled it to simultaneously carry out a significant array of missions while demonstrating its features that contributed to a high level of plant safety and availability. FFTF is currently being deactivated for final closure.

NIELSEN, D L

2004-02-26T23:59:59.000Z

297

Advanced Test Reactor National Scientific User Facility: Addressing advanced nuclear materials research  

SciTech Connect (OSTI)

The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials with Particles and Components Testing (IMPACT) facility and the Pacific Northwest Nuclear Laboratory (PNNL) Radiochemistry Processing Laboratory (RPL) and PIE facilities were added. The ATR NSUF annually hosts a weeklong event called User’s Week in which students and faculty from universities as well as other interested parties from regulatory agencies or industry convene in Idaho Falls, Idaho to see presentations from ATR NSUF staff as well as select researchers from the materials research field. User’s week provides an overview of current materials research topics of interest and an opportunity for young researchers to understand the process of performing work through ATR NSUF. Additionally, to increase the number of researchers engaged in LWR materials issues, a series of workshops are in progress to introduce research staff to stress corrosion cracking, zirconium alloy degradation, and uranium dioxide degradation during in-reactor use.

John Jackson; Todd Allen; Frances Marshall; Jim Cole

2013-03-01T23:59:59.000Z

298

HIGS Flux Performance Projection  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-SeriesFlickr FlickrGuided Self-Assembly of GoldHAWCHIGS flux performance table

299

Removal of volatile organic compounds from polluted air in a reverse flow reactor: An experimental study  

SciTech Connect (OSTI)

An experimental study of the reverse flow reactor for the purification of contaminated air has been carried out. An experimental reactor with an inner diameter of 0.145 m has been constructed. It almost completely reached the goal of an adiabatically operating system. The influence of several operating parameters such as gas velocity, cycle period, chemical character, and concentration of the pollutants and reactor pressure are discussed. The reactor could be operated autothermally provided that the inlet concentrations were sufficiently high. If a mixture of contaminants is fed to the reactor, it might be necessary to increase the total hydrocarbon concentration to assure an autothermal process. Increasing the reactor pressure will hardly change the axial temperature profiles, if the mass flux is kept constant. Increasing the mass flow rate will lead to a higher plateau temperature. Not only the reactor behavior at fixed operating conditions, but also the response of the reactor toward variations in inlet conditions is reported.

Beld, B. van de; Borman, R.A.; Derkx, O.R.; Woezik, B.A.A. van; Westerterp, K.R. (Univ. of Twente, Enschede (Netherlands). Dept. of Chemical Engineering)

1994-12-01T23:59:59.000Z

300

Numerical tools applied to power reactor noise analysis  

SciTech Connect (OSTI)

In this paper, the development of numerical tools allowing the determination of the neutron noise in power reactors is reported. These tools give the space-dependence of the fluctuations of the neutron flux induced by fluctuating properties of the medium in the 2-group diffusion approximation and in a 2-dimensional representation of heterogeneous systems. Some applications of these tools to power reactor noise analysis are then described. These applications include the unfolding of the noise source from the resulting neutron noise measured at a few discrete locations throughout the core, the study of the space-dependence of the Decay Ratio in Boiling Water Reactors, the noise-based estimation of the Moderator Temperature Coefficient of reactivity in Pressurized Water Reactors, the modeling of shell-mode core barrel vibrations in Pressurized Water Reactors, and the investigation of the validity of the point-kinetic approximation in subcritical systems. (authors)

Demaziere, C.; Pazsit, I. [Chalmers Univ. of Technology, Dept. of Nuclear Engineering, SE-412 96 Goeteborg (Sweden)

2006-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "flux beam reactor" 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

Undergraduate reactor control experiment  

SciTech Connect (OSTI)

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

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

1992-01-01T23:59:59.000Z

302

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

303

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

304

Reactor Sharing Program  

SciTech Connect (OSTI)

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

Vernetson, W.G.

1993-01-01T23:59:59.000Z

305

High solids fermentation reactor  

DOE Patents [OSTI]

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

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

1993-03-02T23:59:59.000Z

306

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

307

Advanced Test Reactor Tour  

SciTech Connect (OSTI)

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

Miley, Don

2011-01-01T23:59:59.000Z

308

Advanced Test Reactor Tour  

ScienceCinema (OSTI)

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

Miley, Don

2013-05-28T23:59:59.000Z

309

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

310

Hypothetical Reactor Accident Study  

E-Print Network [OSTI]

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

311

Hanging core support system for a nuclear reactor. [LMFBR  

DOE Patents [OSTI]

For holding the reactor core in the confining reactor vessel, a support is disclosed that is structurally independent of the vessel, that is dimensionally accurate and stable, and that comprises tandem tension linkages that act redundantly of one another to maintain stabilized core support even in the unlikely event of the complete failure of one of the linkages. The core support has a mounting platform for the reactor core, and unitary structure including a flange overlying the top edge of the reactor vessels, and a skirt and box beams between the flange and platform for establishing one of the linkages. A plurality of tension rods connect between the deck closing the reactor vessel and the platform for establishing the redundant linkage. Loaded Belleville springs flexibly hold the tension rods at the deck and separable bayonet-type connections hold the tension rods at the platform.

Burelbach, J.P.; Kann, W.J.; Pan, Y.C.; Saiveau, J.G.; Seidensticker, R.W.

1984-04-26T23:59:59.000Z

312

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

SciTech Connect (OSTI)

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

Not Available

1986-01-01T23:59:59.000Z

313

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

314

Neutronic analysis of a fusion hybrid reactor  

SciTech Connect (OSTI)

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

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

2012-07-01T23:59:59.000Z

315

Spinning fluids reactor  

DOE Patents [OSTI]

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

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

2012-11-20T23:59:59.000Z

316

Reactor water cleanup system  

DOE Patents [OSTI]

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

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

1994-12-20T23:59:59.000Z

317

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

318

Physics of String Flux Compactifications  

E-Print Network [OSTI]

We provide a qualitative review of flux compactifications of string theory, focusing on broad physical implications and statistical methods of analysis.

Frederik Denef; Michael R. Douglas; Shamit Kachru

2007-01-06T23:59:59.000Z

319

Fluxes, Gaugings and Gaugino Condensates  

E-Print Network [OSTI]

Based on the correspondence between the N = 1 superstring compactifications with fluxes and the N = 4 gauged supergravities, we study effective N = 1 four-dimensional supergravity potentials arising from fluxes and gaugino condensates in the framework of orbifold limits of (generalized) Calabi-Yau compactifications. We give examples in heterotic and type II orientifolds in which combined fluxes and condensates lead to vacua with small supersymmetry breaking scale. We clarify the respective roles of fluxes and condensates in supersymmetry breaking, and analyze the scaling properties of the gravitino mass.

J. -P. Derendinger; C. Kounnas; P. M. Petropoulos

2006-02-10T23:59:59.000Z

320

Development of probes for assessment of ion heat transport and sheath heat flux in the boundary of the Alcator C-Mod Tokamak  

E-Print Network [OSTI]

Progress towards a viable fusion reactor will require comprehensive understanding of boundary plasma physics. Knowledge in this area has been growing, yet there are critical gaps. Measurements of the sheath heat flux ...

Brunner, Daniel Frederic

2013-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "flux beam reactor" 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

Fusion reactor control study. Volume 4: inertial confinement reactors. Final report  

SciTech Connect (OSTI)

This study of inertial confinement fusion (ICF) reactor control investigated concepts of the type intended to be driven by laser, electron, or light-ion pulsed energy beams. The study delineates the major reactor control functions, the methods and techniques advanced so far to perform those functions, and the problems, uncertainties, and issues associated with their possible implementation. The perceived shortcomings of some proposed methods of beam/target interaction initiated a search for potentially better solutions to the guidance/pointing/tracking control problem. A preliminary study of a new scheme to accomplish this most important control function is described. The simulated performance of the concept, which involves the active control of the intensity of a laser tube through which the fuel pellet travels to the target point, is encouraging. However, it is concluded that a more detailed study including experimental verification is required to establish the practicality of the concept.

Chang, F.R.; Fisher, J.L.; Madden, P.A.

1982-03-01T23:59:59.000Z

322

Uncertainties analysis of fission fraction for reactor antineutrino experiments  

E-Print Network [OSTI]

Reactor antineutrino experiment are used to study neutrino oscillation, search for signatures of nonstandard neutrino interaction, and monitor reactor operation for safeguard application. Reactor simulation is an important source of uncertainties for a reactor neutrino experiment. Commercial code is used for reactor simulation to evaluate fission fraction in Daya Bay neutrino experiment, but the source code doesn't open to our researcher results from commercial secret. In this study, The open source code DRAGON was improved to calculate the fission rates of the four most important isotopes in fissions, $^{235}$U,$^{238}$U,$^{239}$Pu and $^{241}$Pu, and then was validated for PWRs using the Takahama-3 benchmark. The fission fraction results are consistent with those of MIT's results. Then, fission fraction of Daya Bay reactor core was calculated by using improved DRAGON code, and the fission fraction calculated by DRAGON agreed well with these calculated by SCIENCE. The average deviation less than 5\\% for all the four isotopes. The correlation coefficient matrix between $^{235}$U,$^{238}$U,$^{239}$Pu and $^{241}$Pu were also studied using DRAGON, and then the uncertainty of the antineutrino flux by the fission fraction was calculated by using the correlation coefficient matrix. The uncertainty of the antineutrino flux by the fission fraction simulation is 0.6\\% per core for Daya Bay antineutrino experiment. The uncertainties source of fission fraction calculation need further to be studied in the future.

X. B. Ma; F. Lu; L. Z. Wang; Y. X. Chen; W. L. Zhong; F. P. An

2015-03-17T23:59:59.000Z

323

Modeling the interaction of high power ion or electron beams with solid target materials  

SciTech Connect (OSTI)

Intense energy deposition on first wall materials and other components as a result of plasma disruptions in magnetic fusion devices are expected to cause melting and vaporization of these materials. The exact amount of vaporization losses and melt layer thickness are very important to fusion reactor design and lifetime. Experiments using ion or electron beams to simulate the disruption effects have different environments than the actual disruption conditions in fusion reactors. A model has been developed to accurately simulate the beam-target interactions so that the results from such experiments can be meaningful and useful to reactor design. This model includes a two dimensional solution of the heat conduction equation with moving boundaries. It is found that the vaporization and melting of the sample strongly depends on the characteristics of the beam spatial distribution, beam diameter, and on the power-time variation of the beam.

Hassanein, A.M.

1983-11-01T23:59:59.000Z

324

Optical Fiber Technique for In-Reactor Mechanical Properties Measurement  

SciTech Connect (OSTI)

In-reactor measurement of material properties is required for a better understanding of radiation effects on materials. We present an optical fiber based technique for measuring changes in elastic properties which involves exciting and measuring flexural vibrations in a thin cantilever beam. By exciting the beam and measuring the natural frequency, changes in the modulus of elasticity can be monitored. The technique is demonstrated by monitoring the elastic property changes of a beam fabricated from copper, as the copper undergoes recrystallization at elevated temperature.

Robert S. Schley; Zilong Hua; David H. Hurley; Heng Ban

2012-07-01T23:59:59.000Z

325

Comparison and validation of HEU and LEU modeling results to HEU experimental benchmark data for the Massachusetts Institute of Technology MITR reactor.  

SciTech Connect (OSTI)

The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (UMo) is expected to allow the conversion of U.S. domestic high performance reactors like the MITR-II reactor. Towards this goal, comparisons of MCNP5 Monte Carlo neutronic modeling results for HEU and LEU cores have been performed. Validation of the model has been based upon comparison to HEU experimental benchmark data for the MITR-II. The objective of this work was to demonstrate a model which could represent the experimental HEU data, and therefore could provide a basis to demonstrate LEU core performance. This report presents an overview of MITR-II model geometry and material definitions which have been verified, and updated as required during the course of validation to represent the specifications of the MITR-II reactor. Results of calculations are presented for comparisons to historical HEU start-up data from 1975-1976, and to other experimental benchmark data available for the MITR-II Reactor through 2009. This report also presents results of steady state neutronic analysis of an all-fresh LEU fueled core. Where possible, HEU and LEU calculations were performed for conditions equivalent to HEU experiments, which serves as a starting point for safety analyses for conversion of MITR-II from the use of HEU fuel to the use of UMo LEU fuel.

Newton, T. H.; Wilson, E. H; Bergeron, A.; Horelik, N.; Stevens, J. (Nuclear Engineering Division); (MIT Nuclear Reactor Lab.)

2011-03-02T23:59:59.000Z

326

Fission neutron/gamma irradiation of Bacillus thuringiensis bacteria at the Texas A&M University Nuclear Science Center Reactor  

E-Print Network [OSTI]

The objective of this research is to fully characterize the effectiveness of the Texas A&M University Nuclear Science Center Reactor (TAMU NSCR) neutrons for bacterial sterilization, and to assess the secondary gamma flux produced when neutrons...

Hearnsberger, David Wayne

2012-06-07T23:59:59.000Z

327

Computational Nuclear Forensics Analysis of Weapons-grade Plutonium Separated from Fuel Irradiated in a Thermal Reactor  

E-Print Network [OSTI]

have been irradiated to the desired burnup in the Oak Ridge National Laboratory- High Flux Isotope Reactor (ORNL-HFIR), and then separated using the PUREX process to experimentally determine the intrinsic signature of the fuel. The experimental data...

Coles, Taylor Marie

2014-04-27T23:59:59.000Z

328

Anisotropic etching of polymer films by high energy ,,100s of eV... oxygen atom neutral beams  

E-Print Network [OSTI]

Anisotropic etching of polymer films by high energy ,,Ă?100s of eV... oxygen atom neutral beams to generate an energetic 100s of eV , high flux equivalent of 10s mA/cm2 oxygen atom neutral beam. Positive of the boundary voltage which controls neutral beam energy , and was independent of substrate temperature

Economou, Demetre J.

329

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

330

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

331

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

332

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

333

Next Generation Reactors  

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

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

334

The Neutrino Mass Hierarchy from Nuclear Reactor Experiments  

E-Print Network [OSTI]

10 years from now reactor neutrino experiments will attempt to determine which neutrino mass eigenstate is the most massive. In this letter we present the results of more than seven million detailed simulations of such experiments, studying the dependence of the probability of successfully determining the mass hierarchy upon the analysis method, the neutrino mass matrix parameters, reactor flux models, geoneutrinos and, in particular, combinations of baselines. We show that a recently reported spurious dependence of the data analysis upon the high energy tail of the reactor spectrum can be removed by using a weighted Fourier transform. We determine the optimal baselines and corresponding detector locations. For most values of the CP-violating, leptonic Dirac phase delta, a degeneracy prevents NOvA and T2K from determining either delta or the hierarchy. We determine the confidence with which a reactor experiment can determine the hierarchy, breaking the degeneracy.

Emilio Ciuffoli; Jarah Evslin; Xinmin Zhang

2013-08-14T23:59:59.000Z

335

Eastern Europe Research Reactor Initiative nuclear education and training courses - Current activities and future challenges  

SciTech Connect (OSTI)

The Eastern Europe Research Reactor Initiative was established in January 2008 to enhance cooperation between the Research Reactors in Eastern Europe. It covers three areas of research reactor utilisation: irradiation of materials and fuel, radioisotope production, neutron beam experiments, education and training. In the field of education and training an EERRI training course was developed. The training programme has been elaborated with the purpose to assist IAEA Member States, which consider building a research reactor (RR) as a first step to develop nuclear competence and infrastructure in the Country. The major strength of the reactor is utilisation of three different research reactors and a lot of practical exercises. Due to high level of adaptability, the course can be tailored to specific needs of institutions with limited or no access to research reactors. (authors)

Snoj, L. [Josef Stefan Inst., Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Sklenka, L.; Rataj, J. [Dept. of Nuclear Reactor, Czech Technical Univ. in Prague, V Holesovickach 2, 180 00 Prague 8 (Czech Republic); Boeck, H. [Vienna Univ. of Technology/Atominstitut, Stadionallee 2, 1020 Vienna (Austria)

2012-07-01T23:59:59.000Z

336

Description of a research reactor control system using a programmable controller  

SciTech Connect (OSTI)

This paper describes the design features, testing methods, and operational experience of a programmable controller (PC) installed as a neutron flux controller in the Oak Ridge Research Reactor (ORR) at Oak Ridge National Laboratory (ORNL). The PC was designed to control neutron flux from 1 to 105% for three selectable ranges. The PC generates a flux setpoint under operator control, calculates the reactor heat power from flow and temperature signals, calculates a neutron flux calibration factor based on the heat power, and positions a control rod based on the flux-setpoint difference. The programmable controller was tested by controlling an analog computer model of the ORR. The equipment was installed in August 1985, and except for some startup problems, the system has performed well.

Battle, R.E.

1986-01-01T23:59:59.000Z

337

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

Office of Scientific and Technical Information (OSTI)

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

338

Brookhaven Graphite Research Reactor Workshop  

Broader source: Energy.gov [DOE]

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

339

Portfolio for fast reactor collaboration  

SciTech Connect (OSTI)

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

Rippon, S.

1981-12-01T23:59:59.000Z

340

REACTOR OPERATIONS AND CONTROL  

E-Print Network [OSTI]

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

Pázsit, Imre

Note: This page contains sample records for the topic "flux beam reactor" 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

Advanced Models of LWR Pressure Vessel Embrittlement for Low Flux-HighFluence Conditions  

SciTech Connect (OSTI)

Neutron embrittlement of reactor pressure vessels (RPVs) is an unresolved issue for light water reactor life extension, especially since transition temperature shifts (TTS) must be predicted for high 80-year fluence levels up to approximately 1,020 n/cm{sup 2}, far beyond the current surveillance database. Unfortunately, TTS may accelerate at high fluence, and may be further amplified by the formation of late blooming phases that result in severe embrittlement even in low-copper (Cu) steels. Embrittlement by this mechanism is a potentially significant degradation phenomenon that is not predicted by current regulatory models. This project will focus on accurately predicting transition temperature shifts at high fluence using advanced physically based, empirically validated and calibrated models. A major challenge is to develop models that can adjust test reactor data to account for flux effects. Since transition temperature shifts depend on synergistic combinations of many variables, flux-effects cannot be treated in isolation. The best current models systematically and significantly under-predict transition temperature at high fluence, although predominantly for irradiations at much higher flux than actual RPV service. This project will integrate surveillance, test reactor and mechanism data with advanced models to address a number of outstanding RPV embrittlement issues. The effort will include developing new databases and preliminary models of flux effects for irradiation conditions ranging from very low (e.g., boiling water reactor) to high (e.g., accelerated test reactor). The team will also develop a database and physical models to help predict the conditions for the formation of Mn-Ni-Si late blooming phases and to guide future efforts to fully resolve this issue. Researchers will carry out other tasks on a best-effort basis, including prediction of transition temperature shift attenuation through the vessel wall, remediation of embrittlement by annealing, and fracture toughness master curve issues.

Odette, G. Robert; Yamamoto, Takuya

2013-06-17T23:59:59.000Z

342

Non-Vacuum Electron Beam Welding  

SciTech Connect (OSTI)

Original objectives of CRADA number BNL-01-03 between BNL and Acceleron, Inc., were to further develop the Plasma Window concept (a BNL invention covered by US Patent number 5,578,831), mate the Plasma Window to an existing electron beam welder to perform in-air electron beam welding, and mount the novel nonvacuum electron beam welder on a robot arm. Except for the last objective, all other goals were met or exceeded. Plasma Window design and operation was enhanced during the project, and it was successfully mated to a conventional4 kW electron beam welder. Unprecedented high quality non-vacuum electron beam . welding was demonstrated. Additionally, a new invention the Plasma Shield (US Patent number 7,075,030) that chemically and thermally shields a target object was set forth. Great interest in the new technology was shown by a number of industries and three arcs were sold for experimental use. However, the welding industry requested demonstration of high speed welding, which requires 100 kW electron beam welders. The cost of such a welder involved the need for additional funding. Therefore, some of the effort was directed towards Plasma Shield development. Although relatively a small portion of the R&D effort was spent on the Plasma Shield, some very encouraging results were obtained. Inair Plasma Shield was demonstrated. With only a partial shield, enhanced vacuum separation and cleaner welds were realized. And, electron beam propagation in atmosphere improved by a factor of about 3. Benefits to industry are the introduction of two new technologies. BNL benefited from licensing fee cash, from partial payment for employee salary, and from a new patent In addition to financial benefits, a new technology for physics studies was developed. Recommendations for future work are to develop an under-water plasma shield, perform welding with high-power electron beam:s, carry out other plasma shielded electron beam and laser processes. Potential benefits from further R&D are that various processes involving electron ion and laser beams that have now restrictions can, with the Plasma Shield be performed in practically any environment. For example, electron beam and laser welding can be performed under water, as well as, in situ repair of ship and nuclear reactor components. The plasma shield results in both thermal (since the plasma is hotter than the environment) and chemical shielding. The latter feature brings about in-vacuum process purity out of vacuum, and the thermal shielding aspect results in higher production rates.

Hershcovitch, Ady

2007-01-31T23:59:59.000Z

343

The Solar Wind Energy Flux  

E-Print Network [OSTI]

The solar-wind energy flux measured near the ecliptic is known to be independent of the solar-wind speed. Using plasma data from Helios, Ulysses, and Wind covering a large range of latitudes and time, we show that the solar-wind energy flux is independent of the solar-wind speed and latitude within 10%, and that this quantity varies weakly over the solar cycle. In other words the energy flux appears as a global solar constant. We also show that the very high speed solar-wind (VSW > 700 km/s) has the same mean energy flux as the slower wind (VSW < 700 km/s), but with a different histogram. We use this result to deduce a relation between the solar-wind speed and density, which formalizes the anti-correlation between these quantities.

Chat, G Le; Meyer-Vernet, N

2012-01-01T23:59:59.000Z

344

Method of fission heat flux determination from experimental data  

DOE Patents [OSTI]

A method is provided for determining the fission heat flux of a prime specimen inserted into a specimen of a test reactor. A pair of thermocouple test specimens are positioned at the same level in the holder and a determination is made of various experimental data including the temperature of the thermocouple test specimens, the temperature of bulk water channels located in the test holder, the gamma scan count ratios for the thermocouple test specimens and the prime specimen, and the thicknesses of the outer clads, the fuel fillers, and the backclad of the thermocouple test specimen. Using this experimental data, the absolute value of the fission heat flux for the thermocouple test specimens and prime specimen can be calculated.

Paxton, Frank A. (Schenectady, NY)

1999-01-01T23:59:59.000Z

345

Fusion reactor control study. Volume 3. Tandem mirror reactors. Final report  

SciTech Connect (OSTI)

A study of the control requirements of the Tandem Mirror Reactor concept is reported. The study describes the development of a control simulator that is based upon a spatially averaged physics code of the reactor concept. The simulator portrays the evolution of the plasma through the complete reactor operating cycle; it includes models of the control and measurement system, thus allowing the exploration of various strategies for reactor control. Startup, shutdown, and control during the quasi-steady-state power producing phase were explored. Configurations are described which use a variety of control effectors including modulation of the refueling rate, beam current, and electron cyclotron resonance heating. Multivariable design techniques were used to design the control laws and compensators for the feedback controllers and presume the practical measurement of only a subset of the plasma and machine variables. Performance of the various controllers is explored using the nonlinear control simulator. Derivative control strategies using new or developed sensors and effectors appropriate to a power reactor environment are postulated, based upon the results of the control configurations tested. Research and development requirements for these controls are delineated.

Chang, F.R.; DeCanio, F.; Fisher, J.L.; Madden, P.A.

1982-03-01T23:59:59.000Z

346

Reactor & Nuclear Systems Publications | ORNL  

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

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

347

Reed Reactor Facility. Final report  

SciTech Connect (OSTI)

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

Frantz, S.G.

1994-12-31T23:59:59.000Z

348

DOE reactor-pumped laser program  

SciTech Connect (OSTI)

FALCON is a high-power, steady-state, nuclear reactor-pumped laser (RPL) concept that is being developed by the Department of Energy. The FALCON program has experimentally demonstrated reactor-pumped lasing in various mixtures of xenon, argon, neon, and helium at at wavelengths of 585, 703, 725, 1271, 1733, 1792, 2032, 2630, 2650, and 3370 nm with intrinsic efficiency as high as 2.5%. The major strengths of a reactor-pumped laser are continuous high-power operation, modular construction, self-contained power, compact size, and a variety of wavelengths (from visible to infrared). These characteristics suggest numerous applications not easily accessible to other laser types. A ground-based RPL could beam its power to space for such activities as illuminating geosynchronous communication satellites in the earth`s shadow to extend their lives, beaming power to orbital transfer vehicles, removing space debris, and providing power (from earth) to a lunar base during the long lunar night. The compact size and self-contained power also makes an RPL very suitable for ship basing so that power-beaming activities could be situated around the globe. The continuous high power of an RPL opens many potential manufacturing applications such as deep-penetration welding and cutting of thick structures, wide-area hardening of metal surfaces by heat treatment or cladding application, wide-area vapor deposition of ceramics onto metal surfaces, production of sub-micron sized particles for manufacturing of ceramics, wide-area deposition of diamond-like coatings, and 3-D ceramic lithography.

Felty, J.R. [USDOE, Germantown, MD (United States). Defense Programs; Lipinski, R.J.; McArthur, D.A.; Pickard, P.S. [Sandia National Labs., Albuquerque, NM (United States)

1993-12-31T23:59:59.000Z

349

BEAMS3D Neutral Beam Injection Model  

SciTech Connect (OSTI)

With the advent of applied 3D fi elds in Tokamaks and modern high performance stellarators, a need has arisen to address non-axisymmetric effects on neutral beam heating and fueling. We report on the development of a fully 3D neutral beam injection (NBI) model, BEAMS3D, which addresses this need by coupling 3D equilibria to a guiding center code capable of modeling neutral and charged particle trajectories across the separatrix and into the plasma core. Ionization, neutralization, charge-exchange, viscous velocity reduction, and pitch angle scattering are modeled with the ADAS atomic physics database [1]. Benchmark calculations are presented to validate the collisionless particle orbits, neutral beam injection model, frictional drag, and pitch angle scattering effects. A calculation of neutral beam heating in the NCSX device is performed, highlighting the capability of the code to handle 3D magnetic fields.

Lazerson, Samuel

2014-04-14T23:59:59.000Z

350

Novel rubidium atomic beam with an alkali dispenser source  

SciTech Connect (OSTI)

We describe a novel atomic beam apparatus with a resistively heated alkali dispenser source and a cold-pumped intermediate chamber. Using laser fluorescence spectroscopy we have measured the atomic density to be 3x10{sup 11} atoms/m{sup 3} and the total flux to be 5x10{sup 8} atoms/s in a 0.3 cm diameter beam. We have also characterized the velocity distribution of the source based on the Doppler-shifted fluorescence spectrum. The compact geometry, flexibility, and simplicity of the beam may make it useful as an optical frequency reference or for experiments on atom-cooling.

Roach, Timothy M.; Henclewood, Dwayne [College of the Holy Cross, Worcester, Massachusetts 01610 (United States)

2004-11-01T23:59:59.000Z

351

A TALE OF TWO BEAMS: GAUSSIAN BEAMS AND BESSEL BEAMS ROBERT L. NOWACK  

E-Print Network [OSTI]

A TALE OF TWO BEAMS: GAUSSIAN BEAMS AND BESSEL BEAMS ROBERT L. NOWACK Abstract. An overview is given of two types of focused beams, Gaussian beams and Bessel beams. First I describe some of the basic properties of Gaussian beams in homogeneous media which stay collimated over a certain distance range after

Nowack, Robert L.

352

Laser-fusion targets for reactors  

DOE Patents [OSTI]

A laser target comprising a thermonuclear fuel capsule composed of a centrally located quantity of fuel surrounded by at least one or more layers or shells of material for forming an atmosphere around the capsule by a low energy laser prepulse. The fuel may be formed as a solid core or hollow shell, and, under certain applications, a pusher-layer or shell is located intermediate the fuel and the atmosphere forming material. The fuel is ignited by symmetrical implosion via energy produced by a laser, or other energy sources such as an electron beam machine or ion beam machine, whereby thermonuclear burn of the fuel capsule creates energy for applications such as generation of electricity via a laser fusion reactor.

Nuckolls, John H. (Livermore, CA); Thiessen, Albert R. (Livermore, CA)

1987-01-01T23:59:59.000Z

353

A review of experiments and results from the transient reactor test (TREAT) facility.  

SciTech Connect (OSTI)

The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop.

Deitrich, L. W.

1998-07-28T23:59:59.000Z

354

Beam Dynamics for ARIA  

E-Print Network [OSTI]

Beam dynamics issues are assessed for a new linear induction electron accelerator being designed for flash radiography of large explosively driven hydrodynamic experiments. Special attention is paid to equilibrium beam transport, possible emittance growth, and beam stability. It is concluded that a radiographic quality beam will be produced possible if engineering standards and construction details are equivalent to those on the present radiography accelerators at Los Alamos.

Ekdahl, Carl

2015-01-01T23:59:59.000Z

355

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

356

Nuclear reactor control column  

SciTech Connect (OSTI)

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

Bachovchin, D.M.

1982-08-10T23:59:59.000Z

357

Reactor Safety Research Programs  

SciTech Connect (OSTI)

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

Edler, S. K.

1981-07-01T23:59:59.000Z

358

ATA beam director experiment  

SciTech Connect (OSTI)

This report describes beam director elements for an experiment at the Advanced Test Accelerator. The elements described include a vernier magnet for beam aiming, an achromat magnet, and an isolation system for the beam interface. These components are built at small scale for concept testing. (JDH)

Lee, E.P.; Younger, F.C.; Cruz, G.E.; Nolting, E.

1986-06-23T23:59:59.000Z

359

Fuel and cladding nano-technologies based solutions for long life heat-pipe based reactors  

SciTech Connect (OSTI)

A novel nuclear reactor concept, unifying the fuel pipe with fuel tube functionality has been developed. The structure is a quasi-spherical modular reactor, designed for a very long life. The reactor module unifies the fuel tube with the heat pipe and a graphite beryllium reflector. It also uses a micro-hetero-structure that allows the fission products to be removed in the heat pipe flow and deposited in a getter area in the cold zone of the heat pipe, but outside the neutron flux. The reactor operates as a breed and burn reactor - it contains the fuel pipe with a variable enrichment, starting from the hot-end of the pipe, meant to assure the initial criticality, and reactor start-up followed by area with depleted uranium or thorium that get enriched during the consumption of the first part of the enriched uranium. (authors)

Popa-Simil, L. [LAVM LLC, Los Alamos (United States)

2012-07-01T23:59:59.000Z

360

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 "flux beam reactor" 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

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

362

Fast Breeder Reactor studies  

SciTech Connect (OSTI)

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

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

1980-07-01T23:59:59.000Z

363

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

364

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

365

Numerical investigation of the convection of heat-emitting liquid reactor materials taking stratification into account  

SciTech Connect (OSTI)

Convective heat transfer to a reactor vessel following core destruction is analyzed. Fuel fragment and structural materials are assumed to melt either mix uniformly or stratify. The Navier-Stokes equations were solved numerically with a rectangular coarse difference grid, neglecting the negligibly small contribution of the laminar boundary layer. Calculations for different levels of volume heat release showed that integral heat fluxes at the lateral and top surfaces of the melt are virtually independent of the convective flow scenario. However, the maximum heat flux on the lateral surface is approximately 1.5 times higher for the homogeneous case than for the stratified case. The higher heat flux could result in larger mechanical loads on the reactor vessel, requiring more cooling of the reactor vessel outer surface. 12 refs., 4 figs.

Likhanskii, V.V.; Loboiko, A.I.; Khoruzhii, O.V.

1994-11-01T23:59:59.000Z

366

Knowledge Management at the Fast Flux Test Facility  

SciTech Connect (OSTI)

One of the goals of the Department of Energy’s Office of Nuclear Energy, initiated under the Fuel Cycle Research and Development Program (FCRD) and continued under the Advanced Reactor Concepts Program (ARC) is to preserve the knowledge that has been gained in the United States on Liquid Metal Reactors (LMRs) that could support the development of an environmentally and economically sound nuclear fuel cycle. The Fast Flux Test Facility (FFTF) is the most recent LMR to operate in the United States, from 1982 to 1992, and was designed as a fully instrumented test reactor with on-line, real time test control and performance monitoring of components and tests installed in the reactor. The 10 years of operation of the FFTF provided a very useful framework for testing the advances in LMR safety technology based on passive safety features that may be of increased importance to new designs after the events at Fukushima. Knowledge preservation at the FFTF is focused on the areas of design, construction, and startup of the reactor, as well as on preserving information obtained from 10 years of successful operating history and extensive irradiation testing of fuels and materials. In order to ensure protection of information at risk, the program to date has sequestered reports, files, tapes, and drawings to allow for secure retrieval. The FFTF knowledge management program includes a disciplined and orderly approach to respond to client’s requests for documents and data in order to minimize the search effort and ensure that future requests for this information can be readily accommodated.

Wootan, David W.; Omberg, Ronald P.

2013-06-01T23:59:59.000Z

367

Neutral beam monitoring  

DOE Patents [OSTI]

Method and apparatus for monitoring characteristics of a high energy neutral beam. A neutral beam is generated by passing accelerated ions through a walled cell containing a low energy neutral gas, such that charge exchange neutralizes the high energy ion beam. The neutral beam is monitored by detecting the current flowing through the cell wall produced by low energy ions which drift to the wall after the charge exchange. By segmenting the wall into radial and longitudinal segments various beam conditions are further identified.

Fink, Joel H. (Livermore, CA)

1981-08-18T23:59:59.000Z

368

Reactor hot spot analysis  

SciTech Connect (OSTI)

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

Vilim, R.B.

1985-08-01T23:59:59.000Z

369

P Reactor Grouting  

SciTech Connect (OSTI)

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

None

2010-01-01T23:59:59.000Z

370

The Measurement, interpretation and use of unsteady momentum fluxes in two-phase flow.  

E-Print Network [OSTI]

The steady and unsteady components of the momentum flux in a two-phase flow have been measured at the exit of a vertical pipe by means of an impulse technique using a turning tee and beam. Different electrical filters have ...

Yih, Tien Sieh

1967-01-01T23:59:59.000Z

371

Nuclear reactor control  

SciTech Connect (OSTI)

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

Ingham, R.V.

1980-01-01T23:59:59.000Z

372

Polymerization reactor control  

SciTech Connect (OSTI)

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

Ray, W.H.

1985-01-01T23:59:59.000Z

373

Molten metal reactors  

DOE Patents [OSTI]

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

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

2013-11-05T23:59:59.000Z

374

Uncertainties analysis of fission fraction for reactor antineutrino experiments using DRAGON  

E-Print Network [OSTI]

Rising interest in nuclear reactors as a source of antineutrinos for experiments motivates validated, fast, and accessible simulation to predict reactor rates. First, DRAGON was developed to calculate the fission rates of the four most important isotopes in fissions,235U,238U,239Pu and141Pu, and it was validated for PWRs using the Takahama benchmark. The fission fraction calculation function was validated through comparing our calculation results with MIT's results. we calculate the fission fraction of the Daya Bay reactor core, and compare its with those calculated by the commercial reactor simulation program SCIENCE, which is used by the Daya Bay nuclear power plant, and the results was consist with each other. The uncertainty of the antineutrino flux by the fission fraction was studied, and the uncertainty of the antineutrino flux by the fission fraction simulation is 0.6% per core for Daya Bay antineutrino experiment.

X. B. Ma; L. Z. Wang; Y. X. Chen; W. L. Zhong; F. P. An

2014-05-27T23:59:59.000Z

375

REACTOR PRESSURE VESSEL ISSUES FOR THE LIGHT-WATER REACTOR SUSTAINABILITY PROGRAM  

SciTech Connect (OSTI)

The Light Water Reactor Sustainability Program Plan is a collaborative program between the U.S. Department of Energy and the private sector directed at extending the life of the present generation of nuclear power plants to enable operation to at least 80 years. The reactor pressure vessel (RPV) is one of the primary components requiring significant research to enable such long-term operation. There are significant issues that need to be addressed to reduce the uncertainties in regulatory application, such as, 1) high neutron fluence/long irradiation times, and flux effects, 2) material variability, 3) high-nickel materials, 4)specimen size effects and the fracture toughness master curve, etc. The first issue is the highest priority to obtain the data and mechanistic understanding to enable accurate, reliable embrittlement predictions at high fluences. This paper discusses the major issues associated with long-time operation of existing RPVs and the LWRSP plans to address those issues.

Nanstad, Randy K [ORNL; Odette, George Robert [UCSB

2010-01-01T23:59:59.000Z

376

The determination of neutron energy spectrum in reactor core C1 of reactor VR-1 Sparrow  

SciTech Connect (OSTI)

This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe. Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (author)

Vins, M. [Department of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University, V Holesovickach 2, 180 00 Prague 8 (Czech Republic)], E-mail: vinsmiro@seznam.cz

2008-07-15T23:59:59.000Z

377

F Reactor Inspection  

SciTech Connect (OSTI)

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

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2014-10-29T23:59:59.000Z

378

Reactor Safety Research Programs  

SciTech Connect (OSTI)

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

Dotson, CW

1980-08-01T23:59:59.000Z

379

F Reactor Inspection  

ScienceCinema (OSTI)

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

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2014-11-24T23:59:59.000Z

380

High flux solar energy transformation  

DOE Patents [OSTI]

Disclosed are multi-stage systems for high flux transformation of solar energy allowing for uniform solar intensification by a factor of 60,000 suns or more. Preferred systems employ a focusing mirror as a primary concentrative device and a non-imaging concentrator as a secondary concentrative device with concentrative capacities of primary and secondary stages selected to provide for net solar flux intensification of greater than 2000 over 95 percent of the concentration area. Systems of the invention are readily applied as energy sources for laser pumping and in other photothermal energy utilization processes. 7 figures.

Winston, R.; Gleckman, P.L.; O'Gallagher, J.J.

1991-04-09T23:59:59.000Z

Note: This page contains sample records for the topic "flux beam reactor" 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

High flux solar energy transformation  

DOE Patents [OSTI]

Disclosed are multi-stage systems for high flux transformation of solar energy allowing for uniform solar intensification by a factor of 60,000 suns or more. Preferred systems employ a focusing mirror as a primary concentrative device and a non-imaging concentrator as a secondary concentrative device with concentrative capacities of primary and secondary stages selected to provide for net solar flux intensification of greater than 2000 over 95 percent of the concentration area. Systems of the invention are readily applied as energy sources for laser pumping and in other photothermal energy utilization processes.

Winston, Roland (Chicago, IL); Gleckman, Philip L. (Chicago, IL); O'Gallagher, Joseph J. (Flossmoor, IL)

1991-04-09T23:59:59.000Z

382

Entanglement-assisted electron microscopy based on a flux qubit  

SciTech Connect (OSTI)

A notorious problem in high-resolution biological electron microscopy is radiation damage caused by probe electrons. Hence, acquisition of data with minimal number of electrons is of critical importance. Quantum approaches may represent the only way to improve the resolution in this context, but all proposed schemes to date demand delicate control of the electron beam in highly unconventional electron optics. Here we propose a scheme that involves a flux qubit based on a radio-frequency superconducting quantum interference device, inserted in a transmission electron microscope. The scheme significantly improves the prospect of realizing a quantum-enhanced electron microscope for radiation-sensitive specimens.

Okamoto, Hiroshi, E-mail: okamoto@akita-pu.ac.jp [Department of Electronics and Information Systems, Akita Prefectural University, Yurihonjo 015-0055 (Japan); Nagatani, Yukinori [National Institute for Physiological Sciences, Okazaki 444-8787 (Japan)

2014-02-10T23:59:59.000Z

383

Risk management at the Oak Ridge National Laboratory research reactors  

SciTech Connect (OSTI)

In November of 1986, the High Flux Isotope Reactor (HFIR) was shut down by Oak Ridge National Laboratory (ORNL) due to a concern regarding embrittlement of the reactor vessel. A massive review effort was undertaken by ORNL and the Department of Energy (DOE). This review resulted in an extensive list of analyses and design modifications to be completed before restart could take place. The review also focused on the improvement of management practices including implementation of several of the Institute of Nuclear Power Operations (INPO) requirements. One of the early items identified was the need to perform a Probabilistic Risk Assessment (PRA) on the reactor. It was decided by ORNL management that this PRA would not be just an exercise to assess the ``bottom`` line in order to restart, but would be used to improve the overall safety of the reactor, especially since resources (both manpower and dollars) were severely limited. The PRA would become a basic safety tool to be used instead of a more standard deterministic approach to safety used in commercial reactor power plants. This approach was further reinforced, because the reactor was nearly 25 years old at this time, and the design standards and regulations had changed significantly since the original design, and many of the safety issues could not be addressed by compliance to codes and standards.

Flanagan, G.F.; Linn, M.A.; Proctor, L.D.; Cook, D.H.

1994-12-31T23:59:59.000Z

384

Analysis of a research reactor under anticipated transients without scram events using the RELAP5/MOD3.2 computer program  

E-Print Network [OSTI]

Simulations for two series of anticipated transients phics. without scram (ATWS) events have been carried out for a small, hypothetical, research reactor based on the High Flux Australian Reador HIFAR using the RELAPS/MOD3.Z computer program...

Hari, Sridhar

1998-01-01T23:59:59.000Z

385

Measurement of neutron spectrum in an AGN-201 reactor using a semiconductor neutron spectrometer  

E-Print Network [OSTI]

safely exceed the maximum triton range. The detectors could be refined for the specific application. A detector pair may be miniaturized 29 30 for in-core reactor spectra measurements between fuel-bearing plates and between elements. Low flux..., this point is considered to be essentially valid and free from thermal and low epi- thermal event overlap. Figure 11 presents the result of this research. The reactor core flux spectrum is believed to be valid from 0. 25 Mev to 3. 0 Mev. The measurement...

Stephenson, S. E

2012-06-07T23:59:59.000Z

386

Advanced Test Reactor Testing Experience: Past, Present and Future  

SciTech Connect (OSTI)

The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world’s premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner “lobes” to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 48" long and 5.0" diameter) provide unique testing opportunities. The current experiments in the ATR are for a variety of test sponsors -- US government, foreign governments, private researchers, and commercial companies needing neutron irradiation services. There are three basic types of test configurations in the ATR. The simplest configuration is the sealed static capsule, wherein the target material is placed in a capsule, or plate form, and the capsule is in direct contact with the primary coolant. The next level of complexity of an experiment is an instrumented lead experiment, which allows for active monitoring and control of experiment conditions during the irradiation. The highest level of complexity of experiment is the pressurized water loop experiment, in which the test sample can be subjected to the exact environment of a pressurized water reactor. For future research, some ATR modifications and enhancements are currently planned. This paper provides more details on some of the ATR capabilities, key design features, experiments, and future plans.

Frances M. Marshall

2005-04-01T23:59:59.000Z

387

LBNL-46223, CBP Note 350 BEAM-BEAM SIMULATIONS FOR SEPARATED BEAMS IN THE LHC  

E-Print Network [OSTI]

LBNL-46223, CBP Note 350 BEAM-BEAM SIMULATIONS FOR SEPARATED BEAMS IN THE LHC M. A. Furman, W. C. Turner, Center for Beam Physics, LBNL, Berkeley, CA 94720, USA Abstract We present beam-beam simulation of simulations: (a) to as- sess undesirable effects from LBNL's luminosity monitor- ing scheme for the LHC [2

Furman, Miguel

388

LBNL-45363, CBP Note 333 BEAM-BEAM SIMULATIONS FOR SEPARATED BEAMS  

E-Print Network [OSTI]

LBNL-45363, CBP Note 333 BEAM-BEAM SIMULATIONS FOR SEPARATED BEAMS Miguel A. Furman, Center for Beam Physics, LBNL, Berkeley, CA 94720 Abstract We present beam-beam simulation results from a strong undesirable effects from LBNL's sweeping lumi- nosity monitoring scheme for the LHC [1], and (b) to assess

Furman, Miguel

389

Superconducting flux flow digital circuits  

DOE Patents [OSTI]

A NOR/inverter logic gate circuit and a flip flop circuit implemented with superconducting flux flow transistors (SFFTs) are disclosed. Both circuits comprise two SFFTs with feedback lines. They have extremely low power dissipation, very high switching speeds, and the ability to interface between Josephson junction superconductor circuits and conventional microelectronics. 8 figs.

Hietala, V.M.; Martens, J.S.; Zipperian, T.E.

1995-02-14T23:59:59.000Z

390

Superconducting flux flow digital circuits  

DOE Patents [OSTI]

A NOR/inverter logic gate circuit and a flip flop circuit implemented with superconducting flux flow transistors (SFFTs). Both circuits comprise two SFFTs with feedback lines. They have extremely low power dissipation, very high switching speeds, and the ability to interface between Josephson junction superconductor circuits and conventional microelectronics.

Hietala, Vincent M. (Placitas, NM); Martens, Jon S. (Sunnyvale, CA); Zipperian, Thomas E. (Albuquerque, NM)

1995-01-01T23:59:59.000Z

391

A FLUX SCALE FOR SOUTHERN HEMISPHERE 21 cm EPOCH OF REIONIZATION EXPERIMENTS  

SciTech Connect (OSTI)

We present a catalog of spectral measurements covering a 100-200 MHz band for 32 sources, derived from observations with a 64 antenna deployment of the Donald C. Backer Precision Array for Probing the Epoch of Reionization (PAPER) in South Africa. For transit telescopes such as PAPER, calibration of the primary beam is a difficult endeavor and errors in this calibration are a major source of error in the determination of source spectra. In order to decrease our reliance on an accurate beam calibration, we focus on calibrating sources in a narrow declination range from –46° to –40°. Since sources at similar declinations follow nearly identical paths through the primary beam, this restriction greatly reduces errors associated with beam calibration, yielding a dramatic improvement in the accuracy of derived source spectra. Extrapolating from higher frequency catalogs, we derive the flux scale using a Monte Carlo fit across multiple sources that includes uncertainty from both catalog and measurement errors. Fitting spectral models to catalog data and these new PAPER measurements, we derive new flux models for Pictor A and 31 other sources at nearby declinations; 90% are found to confirm and refine a power-law model for flux density. Of particular importance is the new Pictor A flux model, which is accurate to 1.4% and shows that between 100 MHz and 2 GHz, in contrast with previous models, the spectrum of Pictor A is consistent with a single power law given by a flux at 150 MHz of 382 ± 5.4 Jy and a spectral index of –0.76 ± 0.01. This accuracy represents an order of magnitude improvement over previous measurements in this band and is limited by the uncertainty in the catalog measurements used to estimate the absolute flux scale. The simplicity and improved accuracy of Pictor A's spectrum make it an excellent calibrator in a band important for experiments seeking to measure 21 cm emission from the epoch of reionization.

Jacobs, Daniel C.; Bowman, Judd [School of Earth and Space Exploration, Arizona State University, Tempe, AZ (United States); Parsons, Aaron R.; Ali, Zaki; Pober, Jonathan C. [Astronomy Department, University of California, Berkeley, CA (United States); Aguirre, James E.; Moore, David F. [Department of Physics and Astronomy, University of Pennsylvania, Philadelphia, PA (United States); Bradley, Richard F. [Department of Electrical and Computer Engineering, University of Virginia, Charlottesville, VA (United States); Carilli, Chris L. [National Radio Astronomy Observatory, Socorro, NM (United States); DeBoer, David R.; Dexter, Matthew R.; MacMahon, Dave H. E. [Radio Astronomy Lab., University of California, Berkeley, CA (United States); Gugliucci, Nicole E.; Klima, Pat [National Radio Astronomy Observatory, Charlottesville, VA (United States); Manley, Jason R.; Walbrugh, William P. [Square Kilometer Array, South Africa Project, Cape Town (South Africa); Stefan, Irina I. [Cavendish Laboratory, Cambridge (United Kingdom)

2013-10-20T23:59:59.000Z

392

REACTOR GROUT THERMAL PROPERTIES  

SciTech Connect (OSTI)

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

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

2011-01-28T23:59:59.000Z

393

INITIAL IRRADIATION OF THE FIRST ADVANCED GAS REACTOR FUEL DEVELOPMENT AND QUALIFICATION EXPERIMENT IN THE ADVANCED TEST REACTOR  

SciTech Connect (OSTI)

The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

S. Blaine Grover; David A. Petti

2007-09-01T23:59:59.000Z

394

COMSOL-based Nuclear Reactor Kinetics Studies at the HFIR  

SciTech Connect (OSTI)

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

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

2011-01-01T23:59:59.000Z

395

Broad beam ion implanter  

DOE Patents [OSTI]

An ion implantation device for creating a large diameter, homogeneous, ion beam is described, as well as a method for creating same, wherein the device is characterized by extraction of a diverging ion beam and its conversion by ion beam optics to an essentially parallel ion beam. The device comprises a plasma or ion source, an anode and exit aperture, an extraction electrode, a divergence-limiting electrode and an acceleration electrode, as well as the means for connecting a voltage supply to the electrodes.

Leung, Ka-Ngo (Hercules, CA)

1996-01-01T23:59:59.000Z

396

Methanation assembly using multiple reactors  

DOE Patents [OSTI]

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

Jahnke, Fred C.; Parab, Sanjay C.

2007-07-24T23:59:59.000Z

397

Power Burst Facility (PBF) Reactor Reactor Decommissioning  

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

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

398

Small beam nonparaxiality arrests selffocusing of optical beams Gadi Fibich  

E-Print Network [OSTI]

Small beam nonparaxiality arrests self­focusing of optical beams Gadi Fibich Department­focusing in the presence of small beam nonparaxiality is derived. Analysis of this equation shows that nonparaxiality remains small as the beam propa­ gates. Nevertheless, nonparaxiality arrests self­focusing when the beam

Soatto, Stefano

399

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

Office of Scientific and Technical Information (OSTI)

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

400

An Analysis of Fluxes by Duality  

E-Print Network [OSTI]

M-theory on K3xK3 with non-supersymmetry-breaking G-flux is dual to M-theory on a Calabi-Yau threefold times a 2-torus without flux. This allows for a thorough analysis of the effects of flux without relying on supergravity approximations. We discuss several dual pairs showing that the usual rules of G-flux compactifications work well in detail. We discuss how a transition can convert M2-branes into G-flux. We see how new effects can arise at short distances allowing fluxes to obstruct more moduli than one expects from the supergravity analysis.

Paul S. Aspinwall

2005-04-05T23:59:59.000Z

Note: This page contains sample records for the topic "flux beam reactor" 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

Inexpensive Mini Thermonuclear Reactor  

E-Print Network [OSTI]

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

Alexander Bolonkin; Alexander Bolonkin

402

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

403

Fusion reactor control  

SciTech Connect (OSTI)

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

Plummer, D.A.

1995-12-31T23:59:59.000Z

404

Thermionic Reactor Design Studies  

SciTech Connect (OSTI)

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

Schock, Alfred

1994-08-01T23:59:59.000Z

405

Reactor for exothermic reactions  

DOE Patents [OSTI]

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

Smith, Jr., Lawrence A. (Bellaire, TX); Hearn, Dennis (Houston, TX); Jones, Jr., Edward M. (Friendswood, TX)

1993-01-01T23:59:59.000Z

406

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

407

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

SciTech Connect (OSTI)

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

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

2011-07-01T23:59:59.000Z

408

If sterile neutrinos exist, how can one determine the total solar neutrino fluxes? John N. Bahcall,1,  

E-Print Network [OSTI]

the center of the Sun. This flavor change was seen directly by the comparison of the Sudbury Neutrino measurements made with the KamLAND reactor experiment and with the SNO CC solar neutrino experiment, provided determine the total solar neutrino fluxes (8 B,7 Be, and pp) for comparison with solar model predic- tions

Bahcall, John

409

NEUTRON RADIOGRAPHY (NRAD) REACTOR 64-ELEMENT CORE UPGRADE  

SciTech Connect (OSTI)

The neutron radiography (NRAD) reactor is a 250 kW TRIGA (registered) (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The interim critical configuration developed during the core upgrade, which contains only 62 fuel elements, has been evaluated as an acceptable benchmark experiment. The final 64-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (approximately +/-1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

John D. Bess

2014-03-01T23:59:59.000Z

410

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

411

Fast quench reactor method  

DOE Patents [OSTI]

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

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

1999-08-10T23:59:59.000Z

412

Diagnostics for hybrid reactors  

SciTech Connect (OSTI)

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

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

2012-06-19T23:59:59.000Z

413

Perspectives on reactor safety  

SciTech Connect (OSTI)

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

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

1994-03-01T23:59:59.000Z

414

A method for measurement of delayed neutron parameters for liquid-metal-cooled power reactors  

SciTech Connect (OSTI)

The trend toward increased reliance on passive features for power reactor safety makes it important to obtain the characteristics of the reactor system from measurements on the system. A method is described for solving for the delayed neutron parameters in a liquid-metal power reactor by fitting an analytic solution of the point-kinetics equations to the flux die-away from a dropped rod in an initially critical core. The method includes treatment of those conditions found in a power reactor that depart from those in a critical assembly experiment. These include a comparatively long rod drop time and a detector signal that instead of providing an integrated count rate is a sampled data signal proportional to the instantaneous fission power. The delayed neutron parameter values calculated from a rod drop experiment in the Experimental Breeder Reactor II are in agreement with values calculated using first principles and knowledge of core material composition and nuclear cross sections.

Vilim, R.B. [Argonne National Lab., IL (United States); Brock, R.W. [Babcock and Wilcox, Lynchburg, VA (United States)

1996-06-01T23:59:59.000Z

415

Spatially resolved temperature and heat flux measurements for slow evaporating droplets heated by a microfabricated heater array  

E-Print Network [OSTI]

flux datum per one droplet. No spatial or temporal heat flux information was given. Klassen et al. [12] and di Marzo et al. [13] were the first to use an infrared thermography technique to attempt to measure the spatially and temporally resolved... infrared thermography. Because of the aforementioned limitation of the IR thermography, measurements were only possible outside of the droplets. Michiyoshi and Makino [15] used a dual beam synchroscope to measure the variation of the heater supply...

Paik, Sokwon

2006-08-16T23:59:59.000Z

416

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

417

Scaling of solid state lasers for satellite power beaming applications  

SciTech Connect (OSTI)

The power requirements for a satellite power beaming laser system depend upon the diameter of the beam director, the performance of the adaptive optics system, and the mission requirements. For an 8 meter beam director and overall Strehl ratio of 50%, a 30 kW laser at 850 nm can deliver an equivalent solar flux to a satellite at geostationary orbit. Advances in Diode Pumped Solid State Lasers (DPSSL) have brought these small, efficient and reliable devices to high average power and they should be considered for satellite power beaming applications. Two solid state systems are described: a diode pumped Alexandrite and diode pumped Thulium doped YAG. Both can deliver high average power at 850 nm in a single aperture.

Friedman, H.W.; Albrecht, G.F.; Beach, R.J.

1994-01-01T23:59:59.000Z

418

Reactor operation environmental information document  

SciTech Connect (OSTI)

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

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

1989-12-01T23:59:59.000Z

419

Laser beam generating apparatus  

DOE Patents [OSTI]

Laser beam generating apparatus including a septum segment disposed longitudinally within the tubular structure of the apparatus. The septum provides for radiatively dissipating heat buildup within the tubular structure and for generating relatively uniform laser beam pulses so as to minimize or eliminate radial pulse delays (the chevron effect). 11 figures.

Warner, B.E.; Duncan, D.B.

1993-12-28T23:59:59.000Z

420

Laser beam generating apparatus  

DOE Patents [OSTI]

Laser beam generating apparatus including a septum segment disposed longitudinally within the tubular structure of the apparatus is described. The septum provides for radiatively dissipating heat buildup within the tubular structure and for generating relatively uniform laser beam pulses so as to minimize or eliminate radial pulse delays (the chevron effect). 7 figures.

Warner, B.E.; Duncan, D.B.

1994-02-15T23:59:59.000Z

Note: This page contains sample records for the topic "flux beam reactor" 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

Reactor operation safety information document  

SciTech Connect (OSTI)

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

Not Available

1990-01-01T23:59:59.000Z

422

Reed Reactor Facility Annual Report  

SciTech Connect (OSTI)

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

Frantz, Stephen G.

2000-09-01T23:59:59.000Z

423

Center vortices as composites of monopole fluxes  

E-Print Network [OSTI]

We study the relation between the flux of a center vortex obtained from the center vortex model and the flux formed between monopoles obtained from the Abelian gauge fixing method. Motivated by the Monte Carlo simulations which have shown that almost all monopoles are sitting on the top of vortices, we construct the fluxes of center vortices for $SU(2)$ and $SU(3)$ gauge groups using fractional fluxes of monopoles. Then, we compute the potentials in the fundamental representation induced by center vortices and fractional fluxes of monopoles. We show that by combining the fractional fluxes of monopoles one can produce the center vortex fluxes for $SU(3)$ gauge group in a "center vortex model". Comparing the potentials, we conclude that the fractional fluxes of monopoles attract each other.

Deldar, Sedigheh

2015-01-01T23:59:59.000Z

424

Transport and Non-Invasive Position Detection of Electron Beams from Laser-Plasma Accelerators  

SciTech Connect (OSTI)

The controlled imaging and transport of ultra-relativistic electrons from laser-plasma accelerators is of crucial importance to further use of these beams, e.g. in high peak-brightness light sources. We present our plans to realize beam transport with miniature permanent quadrupole magnets from the electron source through our THUNDER undulator. Simulation results demonstrate the importance of beam imaging by investigating the generated XUV-photon flux. In addition, first experimental findings of utilizing cavity-based monitors for non-invasive beam-position measurements in a noisy electromagnetic laser-plasma environment are discussed.

Osterhoff, Jens; Sokollik, Thomas; Nakamura, Kei; Bakeman, Michael; Weingartner, R; Gonsalves, Anthony; Shiraishi, Satomi; Lin, Chen; vanTilborg, Jeroen; Geddes, Cameron; Schroeder, Carl; Esarey, Eric; Toth, Csaba; DeSantis, Stefano; Byrd, John; Gruner, F; Leemans, Wim

2011-07-20T23:59:59.000Z

425

Performance predictions for a laser intensified thermal beam for use in high resolution Focused Ion Beam instruments  

E-Print Network [OSTI]

Photo-ionization of a laser-cooled and compressed atomic beam from a high-flux thermal source can be used to create a high-brightness ion beam for use in Focus Ion Beam (FIB) instruments. Here we show using calculations and Doppler cooling simulations that an atomic rubidium beam with a brightness of $2.1 \\times 10^7 A/(m^2\\,sr\\,eV)$ at a current of 1 nA can be created using a compact 5 cm long 2D magneto-optical compressor which is more than an order of magnitude better than the current state of the art Liquid Metal Ion Source.

Wouters, S H W; Notermans, R P M J W; Debernardi, N; Mutsaers, P H A; Luiten, O J; Vredenbregt, E J D

2014-01-01T23:59:59.000Z

426

Thermal Reactor Safety  

SciTech Connect (OSTI)

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

Not Available

1980-06-01T23:59:59.000Z

427

NETL - Chemical Looping Reactor  

ScienceCinema (OSTI)

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

None

2014-06-26T23:59:59.000Z

428

Nuclear Reactors and Technology  

SciTech Connect (OSTI)

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

Cason, D.L.; Hicks, S.C. [eds.

1992-01-01T23:59:59.000Z

429

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

430

Today's Material Gauss' Law and Flux  

E-Print Network [OSTI]

by the contents of the box, the box must contain zero net electric charge. Slide 27-31 #12;Gauss' Law and Flux: · The Concept of Flux · Calculating Electric Flux · Symmetry · Gauss's Law · Using Gauss's Law · Conductors that the box must contain net positive electric charge. Slide 27-29 #12;© 2013 Pearson Education, Inc

Ashlock, Dan

431

Neutrino Factories and Beta Beams  

E-Print Network [OSTI]

a Neutrino Factory Based on Muon Beams,” Proc. 2001 ParticleMD. [19] C. Rubbia et al. , “Beam Cooling with Ionisationthe required unstable ion beams has recently been suggested

Zisman, Michael S.

2006-01-01T23:59:59.000Z

432

A Flux Scale for Southern Hemisphere 21cm EoR Experiments  

E-Print Network [OSTI]

We present a catalog of spectral measurements covering a 100-200 MHz band for 32 sources, derived from observations with a 64-antenna deployment of the Donald C. Backer Precision Array for Probing the Epoch of Reionization (PAPER) in South Africa. For transit telescopes such as PAPER, calibration of the primary beam is a difficult endeavor, and errors in this calibration are a major source of error in the determination of source spectra. In order to decrease reliance on accurate beam calibration, we focus on calibrating sources in a narrow declination range from -46d to -40d. Since sources at similar declinations follow nearly identical paths through the primary beam, this restriction greatly reduces errors associated with beam calibration, yielding a dramatic improvement in the accuracy of derived source spectra. Extrapolating from higher frequency catalogs, we derive the flux scale using a Monte-Carlo fit across multiple sources that includes uncertainty from both catalog and measurement errors. Fitting spe...

Jacobs, Daniel C; Aguirre, James E; Ali, Zaki; Bowman, Judd; Bradley, Richard F; Carilli, Christopher L; DeBoer, David R; Dexter, Matthew; Gugliucci, Nicole E; Klima, Pat; MacMahon, Dave H E; Manley, Jason R; Moore, David F; Pober, Jonathan C; Stefan, Irina I; Walbrugh, William P

2013-01-01T23:59:59.000Z

433

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

434

Simulations of beam-beam and beam-wire interactions in RHIC  

SciTech Connect (OSTI)

The beam-beam interaction is one of the dominant sources of emittance growth and luminosity lifetime deterioration. A current carrying wire has been proposed to compensate long-range beam-beam effects in the LHC and strong localized long-range beam-beam effects are experimentally investigated in the RHIC collider. Tune shift, beam transfer function, and beam loss rate are measured in dedicated experiments. In this paper, they report on simulations to study the effect of beam-wire interactions based on diffusive apertures, beam loss rates, and beam transfer function using a parallelized weak-strong beam simulation code (BBSIMC). The simulation results are compared with measurements performed in RHIC during 2007 and 2008.

Kim, Hyung J.; Sen, Tanaji; /Fermilab; Abreu, Natalia P.; Fischer, Wolfram; /Brookhaven

2009-02-01T23:59:59.000Z

435

Beta-beams  

E-Print Network [OSTI]

Beta-beams is a new concept for the production of intense and pure neutrino beams. It is at the basis of a proposed neutrino facility, whose main goal is to explore the possible existence of CP violation in the lepton sector. Here we briefly review the original scenario and the low energy beta-beam. This option would offer a unique opportunity to perform neutrino interaction studies of interest for particle physics, astrophysics and nuclear physics. Other proposed scenarios for the search of CP violation are mentioned.

C. Volpe

2008-02-22T23:59:59.000Z

436

Integration of Novel Flux Coupling Motor and Current Source Inverter...  

Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

of Novel Flux Coupling Motor and Current Source Inverter Novel Flux Coupling Machine without Permanent Magnets John Hsu, Oak Ridge National Laboratory, Flux Coupling...

437

High Heat Flux Thermoelectric Module Using Standard Bulk Material...  

Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

Heat Flux Thermoelectric Module Using Standard Bulk Material High Heat Flux Thermoelectric Module Using Standard Bulk Material Presents high heat flux thermoelectric module design...

438

Benchmark for evaluation and validation of reactor simulations (BEAVRS)  

SciTech Connect (OSTI)

Advances in parallel computing have made possible the development of high-fidelity tools for the design and analysis of nuclear reactor cores, and such tools require extensive verification and validation. This paper introduces BEAVRS, a new multi-cycle full-core Pressurized Water Reactor (PWR) depletion benchmark based on two operational cycles of a commercial nuclear power plant that provides a detailed description of fuel assemblies, burnable absorbers, in-core fission detectors, core loading patterns, and numerous in-vessel components. This benchmark enables analysts to develop extremely detailed reactor core models that can be used for testing and validation of coupled neutron transport, thermal-hydraulics, and fuel isotopic depletion. The benchmark also provides measured reactor data for Hot Zero Power (HZP) physics tests, boron letdown curves, and three-dimensional in-core flux maps from fifty-eight instrumented assemblies. Initial comparisons between calculations performed with MIT's OpenMC Monte Carlo neutron transport code and measured cycle 1 HZP test data are presented, and these results display an average deviation of approximately 100 pcm for the various critical configurations and control rod worth measurements. Computed HZP radial fission detector flux maps also agree reasonably well with the available measured data. All results indicate that this benchmark will be extremely useful in validation of coupled-physics codes and uncertainty quantification of in-core physics computational predictions. The detailed BEAVRS specification and its associated data package is hosted online at the MIT Computational Reactor Physics Group web site (http://crpg.mit.edu/), where future revisions and refinements to the benchmark specification will be made publicly available. (authors)

Horelik, N.; Herman, B.; Forget, B.; Smith, K. [Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 77 Massachusetts Avenue, Cambridge, MA 02139 (United States)

2013-07-01T23:59:59.000Z

439

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

440

Neutral particle beam intensity controller  

DOE Patents [OSTI]

The neutral beam intensity controller is based on selected magnetic defocusing of the ion beam prior to neutralization. The defocused portion of the beam is dumped onto a beam dump disposed perpendicular to the beam axis. Selective defocusing is accomplished by means of a magnetic field generator disposed about the neutralizer so that the field is transverse to the beam axis. The magnetic field intensity is varied to provide the selected partial beam defocusing of the ions prior to neutralization. The desired focused neutral beam portion passes along the beam path through a defining aperture in the beam dump, thereby controlling the desired fraction of neutral particles transmitted to a utilization device without altering the kinetic energy level of the desired neutral particle fraction. By proper selection of the magnetic field intensity, virtually zero through 100% intensity control of the neutral beam is achieved.

Dagenhart, W.K.

1984-05-29T23:59:59.000Z

Note: This page contains sample records for the topic "flux beam reactor" 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

Beam monitoring and Near detector requirements for a Neutrino factory or long baseline beams  

E-Print Network [OSTI]

Neutrino Factory is a facility for future precision studies of neutrino oscillations. A so called near detector is essential for reaching the aimed precision of neutrino oscillation analysis. Main task of the near detector is to measure the flux of the neutrino beam. Such brilliant neutrino source like Neutrino Factory provides also opportunity for precision studies of various neutrino interaction processes in the near detector. We discuss design concepts of such a detector. Results of simulations of a high resolution scintillating fiber tracker show that it is capable to measure the neutrino flux through pure leptonic interactions with an uncertainty of the order of 1%. A full set-up of the near detector consisting of high granularity vertex detector, high resolution tracker and muon catcher is also presented.

Rosen Matev; Roumen Tsenov

2011-10-10T23:59:59.000Z

442

Fusion reactor blanket-main design aspects  

SciTech Connect (OSTI)

The main function of the fusion reactor blanket is ensuring tritium breeding and radiation shield. The blanket version depends on the reactor type (experimental, DEMO, commercial) and its parameters. Blanket operation conditions are defined with the heat flux, neutron load/fluence, cyclic operation, dynamic heating/force loading, MHD effects etc. DEMO/commercial blanket design is distinguished e.g. by rather high heat load and neutron fluence - up to 100 W/cm{sup 2} and 7 MWa/m{sup 2} accordingly. This conditions impose specific requirements for the materials, structure, maintenance of the blanket and its most loaded components - FW and limiter. The liquid Li-Pb eutectic is one of the possible breeder for different kinds of blanket in view of its advantages one of which is the blanket convertibility that allow to have shielding blanket (borated water) or breeding one (Li-Pb eutectic). Using Li-Pb eutectic for both ITER and DEMO blankets have been considered. In the conceptual ITER design the solid eutectic blanket was carried out. The liquid eutectic breeder/coolant is suggested also for the advanced (high parameter) blanket.

Strebkov, Yu.; Sidorov, A.; Danilov, I. [Research and Development Inst. of Power Engineering, Moscow (Russian Federation)

1994-12-31T23:59:59.000Z

443

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

444

Nuclear divisional reactor  

SciTech Connect (OSTI)

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

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

1982-11-02T23:59:59.000Z

445

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

446

Wire Scanner Beam Profile Measurements: LANSCE Facility Beam Development  

SciTech Connect (OSTI)

The Los Alamos Neutron Science Center (LANSCE) is replacing Wire Scanner (WS) beam profile measurement systems. Three beam development tests have taken place to test the new wire scanners under beam conditions. These beam development tests have integrated the WS actuator, cable plant, electronics processors and associated software and have used H{sup -} beams of different beam energy and current conditions. In addition, the WS measurement-system beam tests verified actuator control systems for minimum profile bin repeatability and speed, checked for actuator backlash and positional stability, tested the replacement of simple broadband potentiometers with narrow band resolvers, and tested resolver use with National Instruments Compact Reconfigurable Input and Output (cRIO) Virtual Instrumentation. These beam tests also have verified how trans-impedance amplifiers react with various types of beam line background noise and how noise currents were not generated. This paper will describe these beam development tests and show some resulting data.

Gilpatrick, John D. [Los Alamos National Laboratory; Batygin, Yuri K. [Los Alamos National Laboratory; Gonzales, Fermin [Los Alamos National Laboratory; Gruchalla, Michael E. [Los Alamos National Laboratory; Kutac, Vincent G. [Los Alamos National Laboratory; Martinez, Derwin [Los Alamos National Laboratory; Sedillo, James Daniel [Los Alamos National Laboratory; Pillai, Chandra [Los Alamos National Laboratory; Rodriguez Esparza, Sergio [Los Alamos National Laboratory; Smith, Brian G. [Los Alamos National Laboratory

2012-05-15T23:59:59.000Z

447

Enhanced in-pile instrumentation at the advanced test reactor  

SciTech Connect (OSTI)

Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper reports results from this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and realtime flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted. (authors)

Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T.; Chase, B. M.; Palmer, J.; Condie, K. G.; Davis, K. L. [Idaho National Laboratory, MS 3840, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

2011-07-01T23:59:59.000Z

448

Enhanced In-Pile Instrumentation at the Advanced Test Reactor  

SciTech Connect (OSTI)

Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper provides an update on this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and real-time flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted.

Joy Rempe; Darrell Knudson; Joshua Daw; Troy Unruh; Benjamin Chase; Kurt Davis; Robert Schley; Steven Taylor

2012-08-01T23:59:59.000Z

449

Enhanced In-Pile Instrumentation at the Advanced Test Reactor  

SciTech Connect (OSTI)

Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility (NSUF) in 2007 to support the development and deployment of enhanced in-pile sensors. This paper reports results from this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and real-time flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted.

J. Rempe; D. Knudson; J. Daw; T. Unruh; B. Chase; K. Condie

2011-06-01T23:59:59.000Z

450

A high power beam-on-target test of liquid lithium target for RIA.  

SciTech Connect (OSTI)

Experiments were conducted to demonstrate the stable operation of a windowless liquid lithium target under extreme thermal loads that are equivalent to uranium beams from the proposed Rare Isotope Accelerator (RIA) driver linac. The engineering and safety issues accompanying liquid lithium systems are first discussed. The liquid metal technology knowledge base generated primarily for fast reactors, and liquid metal cooled fusion reactors, was applied to the development of these systems in a nuclear physics laboratory setting. The use of a high energy electron beam for simulating a high power uranium beam produced by the RIA driver linac is also described. Calculations were performed to obtain energy deposition profiles produced by electron beams at up to a few MeV to compare with expected uranium beam energy deposition profiles. It was concluded that an experimental simulation using a 1-MeV electron beam would be a valuable tool to assess beam-jet interaction. In the experiments, the cross section of the windowless liquid lithium target was 5 mm x 10 mm, which is a 1/3rd scale prototype target, and the velocity of the liquid lithium was varied up to 6 m/s. Thermal loads up to 20 kW within a beam spot diameter of 1mm were applied on the windowless liquid lithium target by the 1-MeV electron beam. The calculations showed that the maximum power density and total power deposited within the target, from the electron beam, was equivalent to that of a 200-kW, 400-MeV/u uranium beam. It was demonstrated that the windowless liquid lithium target flowing at velocities as low as 1.8 m/s stably operated under beam powers up to 20 kW without disruption or excessive vaporization.

Nolen, J.; Reed, C.; Novick, V.; Specht, J.; Plotkin, P.; Momozaki,Y.; Gomes, I.

2005-08-29T23:59:59.000Z

451

Single element laser beam shaper  

DOE Patents [OSTI]

A single lens laser beam shaper for converting laser beams from any spatial profile to a flat-top or uniform spatial profile. The laser beam shaper includes a lens having two aspheric surfaces. The beam shaper significantly simplifies the overall structure in comparison with conventional 2-element systems and therefore provides great ease in alignment and reduction of cost.

Zhang, Shukui (Yorktown, VA); Michelle D. Shinn (Newport News, VA)

2005-09-13T23:59:59.000Z

452

REVIEW ARTICLE Taming molecular beams  

E-Print Network [OSTI]

REVIEW ARTICLE Taming molecular beams The motion of neutral molecules in a beam can be manipulated time-varying fields can be used to decelerate or accelerate beams of molecules to any desired velocity. We review the possibilities that this molecular-beam technology offers, ranging from ultrahigh

Loss, Daniel

453

Thermionic Reactor Design Studies  

SciTech Connect (OSTI)

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

Schock, Alfred

1994-06-01T23:59:59.000Z

454

Etalon-induced Baseline Drift And Correction In Atom Flux Sensors Based On Atomic Absorption Spectroscopy  

SciTech Connect (OSTI)

Atom flux sensors based on atomic absorption (AA) spectroscopy are of significant interest in thin film growth as they can provide unobtrusive, element specific, real-time flux sensing and control. The ultimate sensitivity and performance of the sensors are strongly affected by the long-term and short term baseline drift. Here we demonstrate that an etalon effect resulting from temperature changes in optical viewport housings is a major source of signal instability which has not been previously considered or corrected by existing methods. We show that small temperature variations in the fused silica viewports can introduce intensity modulations of up to 1.5%, which in turn significantly deteriorate AA sensor performance. This undesirable effect can be at least partially eliminated by reducing the size of the beam and tilting the incident light beam off the viewport normal.

Du, Yingge; Chambers, Scott A.

2014-10-20T23:59:59.000Z

455

Reliable-linac design for accelerator-driven subcritical reactor systems.  

SciTech Connect (OSTI)

Accelerator reliability corresponding to a very low frequency of beam interrupts is an important new accelerator requirement for accelerator-driven subcritical reactor systems. In this paper we review typical accelerator-reliability requirements and discuss possible methods for meeting these goals with superconducting proton-linac technology.

Wangler, Thomas P.,

2002-01-01T23:59:59.000Z

456

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

SciTech Connect (OSTI)

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

Glasstone, S.; Sesonske, A.

1994-12-31T23:59:59.000Z

457

Light beam frequency comb generator  

DOE Patents [OSTI]

A light beam frequency comb generator uses an acousto-optic modulator to generate a plurality of light beams with frequencies which are uniformly separated and possess common noise and drift characteristics. A well collimated monochromatic input light beam is passed through this modulator to produce a set of both frequency shifted and unshifted optical beams. An optical system directs one or more frequency shifted beams along a path which is parallel to the path of the input light beam such that the frequency shifted beams are made incident on the modulator proximate to but separated from the point of incidence of the input light beam. After the beam is thus returned to and passed through the modulator repeatedly, a plurality of mutually parallel beams are generated which are frequency-shifted different numbers of times and possess common noise and drift characteristics.

Priatko, Gordon J. (Cupertino, CA); Kaskey, Jeffrey A. (Livermore, CA)

1992-01-01T23:59:59.000Z

458

Designing Reactors to Facilitate Decommissioning  

SciTech Connect (OSTI)

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

Richard H. Meservey

2006-06-01T23:59:59.000Z

459

Composite Materials under Extreme Radiation and Temperature Environments of the Next Generation Nuclear Reactors  

SciTech Connect (OSTI)

In the nuclear energy renaissance, driven by fission reactor concepts utilizing very high temperatures and fast neutron spectra, materials with enhanced performance that exceeds are expected to play a central role. With the operating temperatures of the Generation III reactors bringing the classical reactor materials close to their performance limits there is an urgent need to develop and qualify new alloys and composites. Efforts have been focused on the intricate relations and the high demands placed on materials at the anticipated extreme states within the next generation fusion and fission reactors which combine high radiation fluxes, elevated temperatures and aggressive environments. While nuclear reactors have been in operation for several decades, the structural materials associated with the next generation options need to endure much higher temperatures (1200 C), higher neutron doses (tens of displacements per atom, dpa), and extremely corrosive environments, which are beyond the experience on materials accumulated to-date. The most important consideration is the performance and reliability of structural materials for both in-core and out-of-core functions. While there exists a great body of nuclear materials research and operating experience/performance from fission reactors where epithermal and thermal neutrons interact with materials and alter their physio-mechanical properties, a process that is well understood by now, there are no operating or even experimental facilities that will facilitate the extreme conditions of flux and temperature anticipated and thus provide insights into the behaviour of these well understood materials. Materials, however, still need to be developed and their interaction and damage potential or lifetime to be quantified for the next generation nuclear energy. Based on material development advances, composites, and in particular ceramic composites, seem to inherently possess properties suitable for key functions within the operating envelope of both fission and fusion reactors. In advanced fission reactors composite materials are being designed in an effort to extend the life and improve the reliability of fuel rod cladding as well as structural materials. Composites are being considered for use as core internals in the next generation of gas-cooled reactors. Further, next-generation plasma-fusion reactors, such as the International Thermonuclear Experimental Reactor (ITER) will rely on the capabilities of advanced composites to safely withstand extremely high neutron fluxes while providing superior thermal shock resistance.

Simos, N.

2011-05-01T23:59:59.000Z

460

On the interest of carbon-coated plasma reactor for advanced gate stack etching processes  

SciTech Connect (OSTI)

In integrated circuit fabrication the most wide spread strategy to achieve acceptable wafer-to-wafer reproducibility of the gate stack etching process is to dry-clean the plasma reactor walls between each wafer processed. However, inherent exposure of the reactor walls to fluorine-based plasma leads to formation and accumulation of nonvolatile fluoride residues (such as AlF{sub x}) on reactor wall surfaces, which in turn leads to process drifts and metallic contamination of wafers. To prevent this while keeping an Al{sub 2}O{sub 3} reactor wall material, a coating strategy must be used, in which the reactor is coated by a protective layer between wafers. It was shown recently that deposition of carbon-rich coating on the reactor walls allows improvements of process reproducibility and reactor wall protection. The authors show that this strategy results in a higher ion-to-neutral flux ratio to the wafer when compared to other strategies (clean or SiOCl{sub x}-coated reactors) because the carbon walls load reactive radical densities while keeping the same ion current. As a result, the etching rates are generally smaller in a carbon-coated reactor, but a highly anisotropic etching profile can be achieved in silicon and metal gates, whose etching is strongly ion assisted. Furthermore, thanks to the low density of Cl atoms in the carbon-coated reactor, silicon etching can be achieved almost without sidewall passivation layers, allowing fine critical dimension control to be achieved. In addition, it is shown that although the O atom density is also smaller in the carbon-coated reactor, the selectivity toward ultrathin gate oxides is not reduced dramatically. Furthermore, during metal gate etching over high-k dielectric, the low level of parasitic oxygen in the carbon-coated reactor also allows one to minimize bulk silicon reoxidation through HfO{sub 2} high-k gate dielectric. It is then shown that the BCl{sub 3} etching process of the HfO{sub 2} high-k material is highly selective toward the substrate in the carbon-coated reactor, and the carbon-coating strategy thus allows minimizing the silicon recess of the active area of transistors. The authors eventually demonstrate that the carbon-coating strategy drastically reduces on-wafer metallic contamination. Finally, the consumption of carbon from the reactor during the etching process is discussed (and thus the amount of initial deposit that is required to protect the reactor walls) together with the best way of cleaning the reactor after a silicon etching process.

Ramos, R.; Cunge, G.; Joubert, O. [Freescale Semiconductor Inc., 850 Rue Jean Monnet, 38921 Crolles Cedex (France) and Laboratoire des Technologies de la Microelectronique, CNRS, 17 Rue des Martyrs (c/o CEA-LETI), 38054 Grenoble Cedex 9 (France); Laboratoire des Technologies de la Microelectronique, CNRS, 17 Rue des Martyrs (c/o CEA-LETI), 38054 Grenoble Cedex 9 (France)

2007-03-15T23:59:59.000Z

Note: This page contains sample records for the topic "flux beam reactor" from the National Library of EnergyBeta (NLEBeta).
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461

ICFA Beam Dynamics Newsletter  

SciTech Connect (OSTI)

The Collider-Accelerator Department at Brookhaven National Laboratory is building a high-brightness 500 mA capable Energy Recovery Linac (ERL) as one of its main R&D thrusts towards eRHIC, the polarized electron - hadron collider as an upgrade of the operating RHIC facility. The ERL is in final assembly stages, with injection commisioning starting in October 2012. The objective of this ERL is to serve as a platform for R&D into high current ERL, in particular issues of halo generation and control, Higher-Order Mode (HOM) issues, coherent emissions for the beam and high-brightness, high-power beam generation and preservation. The R&D ERL features a superconducting laser-photocathode RF gun with a high quantum efficiency photoccathode served with a load-lock cathode delivery system, a highly damped 5-cell accelerating cavity, a highly flexible single-pass loop and a comprehensive system of beam instrumentation. In this ICFA Beam Dynamics Newsletter article we will describe the ERL in a degree of detail that is not usually found in regular publications. We will discuss the various systems of the ERL, following the electrons from the photocathode to the beam dump, cover the control system, machine protection etc and summarize with the status of the ERL systems.

Ben-Zvi I.; Kuczewski A.; Altinbas, Z.; Beavis, D.; Belomestnykh,; Dai, J. et al

2012-07-01T23:59:59.000Z

462

Progress Update: Reactor Disassembly Grouting  

SciTech Connect (OSTI)

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

Cody, Tom

2010-01-01T23:59:59.000Z

463

Neutrino Oscillation Studies with Reactors  

E-Print Network [OSTI]

Nuclear reactors are one of the most intense, pure, controllable, cost-effective, and well-understood sources of neutrinos.