National Library of Energy BETA

Sample records for flux beam reactor

  1. HFBR handbook, 1992: High flux beam reactor

    SciTech Connect (OSTI)

    Axe, J.D.; Greenberg, R.

    1992-10-01

    Welcome to the High Flux Beam Reactor (HFBR), one of the world premier neutron research facilities. This manual is intended primarily to acquaint outside users (and new Brookhaven staff members) with (almost) everything they need to know to work at the HFBR and to help make the stay at Brookhaven pleasant as well as profitable. Safety Training Programs to comply with US Department of Energy (DOE) mandates are in progress at BNL. There are several safety training requirements which must be met before users can obtain unescorted access to the HFBR. The Reactor Division has prepared specific safety training manuals which are to be sent to experimenters well in advance of their expected arrival at BNL to conduct experiments. Please familiarize yourself with this material and carefully pay strict attention to all the safety and security procedures that are in force at the HFBR. Not only your safety, but the continued operation of the facility, depends upon compliance.

  2. Radiation dosimetry at the BNL High Flux Beam Reactor

    SciTech Connect (OSTI)

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

    1998-02-01

    The HFBR is a heavy water, D{sub 2}O, cooled and moderated reactor with twenty-eight fuel elements containing a maximum of 9.8 kilograms of {sup 235}U. The core is 53 cm high and 48 cm in diameter and has an active volume of 97 liters. The HFBR, which was designed to operate at forty mega-watts, 40 NW, was upgraded to operate at 60 NW. Since 1991, it has operated at 30 MW. In a normal 30 MW operating cycle the HFBR operates 24 hours a day for thirty days, with a six to fourteen day shutdown period for refueling and maintenance work. While most reactors attempts to minimize the escape of neutrons from the core, the HFBR`s D{sub 2}O design allows the thermal neutron flux to peak in the reflector region and maximizes the number of thermal neutrons available to nine horizontal external beams, H-1 to H-9. The HFBR neutron dosimetry effort described here compares measured and calculated energy dependent neutron and gamma ray flux densities and/or dose rates at horizontal beam lines and vertical irradiation thimbles.

  3. RADIATION DOSIMETRY AT THE BNL HIGH FLUX BEAM REACTOR AND MEDICAL RESEARCH REACTOR.

    SciTech Connect (OSTI)

    HOLDEN,N.E.

    1999-09-10

    RADIATION DOSIMETRY MEASUREMENTS HAVE BEEN PERFORMED OVER A PERIOD OF MANY YEARS AT THE HIGH FLUX BEAM REACTOR (HFBR) AND THE MEDICAL RESEARCH REACTOR (BMRR) AT BROOKHAVEN NATIONAL LABORATORY TO PROVIDE INFORMATION ON THE ENERGY DISTRIBUTION OF THE NEUTRON FLUX, NEUTRON DOSE RATES, GAMMA-RAY FLUXES AND GAMMA-RAY DOSE RATES. THE MCNP PARTICLE TRANSPORT CODE PROVIDED MONTE CARLO RESULTS TO COMPARE WITH VARIOUS DOSIMETRY MEASUREMENTS PERFORMED AT THE EXPERIMENTAL PORTS, AT THE TREATMENT ROOMS AND IN THE THIMBLES AT BOTH HFBR AND BMRR.

  4. Updated flux information for neutron scattering and irradiation facilities at the BNL High Flux Beam Reactor

    SciTech Connect (OSTI)

    Holden, N.E.; Hu, J.P.; Reciniello, R.N.; Sengupta, S.; Greenwood, L.R.; Farrell, K.

    1997-08-01

    The HFBR is a heavy water, D{sub 2}O, cooled and moderated reactor with twenty-eight fuel elements containing a maximum of 9.8 kilograms of {sup 235}U. While most reactors attempt to minimize the escape of neutrons from the core, the HFBR`s D{sub 2}O design allows the thermal neutron flux to peak in the reflector region and maximizes the number of thermal neutrons available to nine horizontal external beams, H-1 to H-9, used for neutron scattering and capture reactions, supporting physics, chemistry and biology experiments. All horizontal beam tubes were built tangential to the direction of the emerging neutrons, except for the H-2 beam tube, which looks directly at the core and has been used for neutron cross section measurements utilizing fast neutrons and for the TRISTAN fission product studies. In recent years, there have been some beam modifications and new instrumentation introduced at the HFBR. A high resolution neutron powder diffractometer instrument is now operating with a resolution of 5 {times} 10{sup {minus}4} at horizontal beam line H-1. To study scattering from liquid surfaces, a neutron reflection spectrometer was introduced on the CNF beam line at H-9. In the past year, a fourth beam line has been added to the CNF line at H-9. The existing beam plug at the H-6 beam line has recently been removed and a new plug, which will feature super mirrored surfaces, is now being installed. Last year, the vertical beam thimble, V-13, a fixed port filled with thirty year old samples used for HFBR material surveillance studies was replaced by a new thimble and charging station at the core edge creating an irradiation facility to substitute for the original V-13. A neutron dosimetry program has begun to measure and calculate the energy dependent neutron and gamma ray flux densities and/or dose rates at horizontal beam lines and vertical irradiation thimbles.

  5. Level 1 Tornado PRA for the High Flux Beam Reactor

    SciTech Connect (OSTI)

    Bozoki, G.E.; Conrad, C.S.

    1994-05-01

    This report describes a risk analysis primarily directed at providing an estimate for the frequency of tornado induced damage to the core of the High Flux Beam Reactor (HFBR), and thus it constitutes a Level 1 Probabilistic Risk Assessment (PRA) covering tornado induced accident sequences. The basic methodology of the risk analysis was to develop a ``tornado specific`` plant logic model that integrates the internal random hardware failures with failures caused externally by the tornado strike and includes operator errors worsened by the tornado modified environment. The tornado hazard frequency, as well as earlier prepared structural and equipment fragility data, were used as input data to the model. To keep modeling/calculational complexity as simple as reasonable a ``bounding`` type, slightly conservative, approach was applied. By a thorough screening process a single dominant initiating event was selected as a representative initiator, defined as: ``Tornado Induced Loss of Offsite Power.`` The frequency of this initiator was determined to be 6.37E-5/year. The safety response of the HFBR facility resulted in a total Conditional Core Damage Probability of .621. Thus, the point estimate of the HFBR`s Tornado Induced Core Damage Frequency (CDF) was found to be: (CDF){sub Tornado} = 3.96E-5/year. This value represents only 7.8% of the internal CDF and thus is considered to be a small contribution to the overall facility risk expressed in terms of total Core Damage Frequency. In addition to providing the estimate of (CDF){sub Tornado}, the report documents, the relative importance of various tornado induced system, component, and operator failures that contribute most to (CDF){sub Tornado}.

  6. 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).

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

    SciTech Connect (OSTI)

    P.C. Weaver

    2010-10-22

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

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

    SciTech Connect (OSTI)

    P.C. Weaver

    2010-12-15

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

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

    SciTech Connect (OSTI)

    P.C. Weaver

    2010-07-09

    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

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

    DOE Patents [OSTI]

    Lasche, G.P.

    1983-09-29

    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.

  11. 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)

    E.M. Harpenau

    2010-12-15

    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

  12. 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)

    P.C. Weaver

    2010-11-03

    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

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

    DOE Patents [OSTI]

    Lasche, G.P.

    1987-02-20

    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.

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

    DOE Patents [OSTI]

    Lasche, George P.

    1988-01-01

    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.

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

    SciTech Connect (OSTI)

    Lasche, G.P.

    1988-04-05

    A method for recovering energy in an inertial confinement fusion reactor having a reactor chamber and a sphere forming means positioned above an opening in the reactor chamber is described, comprising: embedding a fusion target fuel capsule having a predetermined yield in the center of a hollow solid lithium tube and subsequently embedding the hollow solid lithium tube in a liquid lithium medium; using the sphere forming means for forming the liquid lithium into a spherical shaped liquid lithium mass having a diameter smaller than the length of the hollow solid lithium tube with the hollow solid lithium tube being positioned along a diameter of the spherical shaped mass, providing the spherical shaped liquid lithium mass with the fusion fuel target capsule and hollow solid lithium tube therein as a freestanding liquid lithium shaped spherical shaped mass without any external means for maintaining the spherical shape by dropping the liquid lithium spherical shaped mass from the sphere forming means into the reactor chamber; producing a magnetic field in the reactor chamber; imploding the target capsule in the reactor chamber to produce fusion energy; absorbing fusion energy in the liquid lithium spherical shaped mass to convert substantially all the fusion energy to shock induced kinetic energy of the liquid lithium spherical shaped mass which expands the liquid lithium spherical shaped mass; and compressing the magnetic field by expansion of the liquid lithium spherical shaped mass and recovering useful energy.

  16. High flux reactor

    DOE Patents [OSTI]

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

    1988-01-01

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

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

    SciTech Connect (OSTI)

    Harpeneau, Evan M.

    2011-06-24

    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.

  18. Type A verification report for the high flux beam reactor stack and grounds, Brookhaven National Laboratory, Upton, New York

    SciTech Connect (OSTI)

    Harpenau, Evan M.

    2012-01-13

    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). The HFBR Stack and Grounds surveys began in June 2011 and were completed in September 2011. Survey activities by BSA included gamma walkover scans and sampling of the as-left soils in accordance with the BSA Work Procedure (BNL 2010a). The Field Sampling Plan - Stack and Remaining HFBR Outside Areas (FSP) stated that gamma walk-over surveys would be conducted with a bare sodium iodide (NaI) detector, and a collimated detector would be used to check areas with elevated count rates to locate the source of the high readings (BNL 2010b). BSA used the Mult- Agency Radiation Survey and Site Investigation Manual (MARSSIM) principles for determining the classifications of each survey unit. Therefore, SUs 6 and 7 were identified as Class 1 and SU 8 was deemed Class 2 (BNL 2010b). Gamma walkover surveys of SUs 6, 7, and 8 were completed using a 2?2 NaI detector coupled to a data-logger with a global positioning system (GPS). The 100% scan surveys conducted prior to the final status survey (FSS) sampling identified two general soil areas and two isolated soil locations with elevated radioactivity. The general areas of elevated activity identified

  19. 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)

    Evan Harpenau

    2011-11-30

    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).

  20. OSTIblog Articles in the High Flux Isotope Reactor Topic | OSTI...

    Office of Scientific and Technical Information (OSTI)

    High Flux Isotope Reactor Topic The NXS Class of 2014 by Kathy Chambers 19 Nov, 2014 in ... National Laboratory, High Flux Isotope Reactor, National School on Neutron and X-ray ...

  1. Revision of HFIR (High Flux Isotope Reactor) operating procedures

    SciTech Connect (OSTI)

    McGinty, D.M.

    1987-01-23

    This report documents modifications to the facility and changes in some operating procedures for the High Flux Isotope Reactor (HFIR). The topics covered include: Reactor Operation, Reactor Start-up, Reactor Safety Systems, Reactor Control Systems, Reporting Requirements, and Administrative Procedures. (FI)

  2. RELAP5 model of the high flux isotope reactor with low enriched fuel thermal flux profiles

    SciTech Connect (OSTI)

    Banfield, J.; Mervin, B.; Hart, S.; Ritchie, J.; Walker, S.; Ruggles, A.; Maldonado, G. I.

    2012-07-01

    The High Flux Isotope Reactor (HFIR) currently uses highly enriched uranium (HEU) fabricated into involute-shaped fuel plates. It is desired that HFIR be able to use low enriched uranium (LEU) fuel while preserving the current performance capability for its diverse missions in material irradiation studies, isotope production, and the use of neutron beam lines for basic research. Preliminary neutronics and depletion simulations of HFIR with LEU fuel have arrived to feasible fuel loadings that maintain the neutronics performance of the reactor. This article illustrates preliminary models developed for the analysis of the thermal-hydraulic characteristics of the LEU core to ensure safe operation of the reactor. The beginning of life (BOL) LEU thermal flux profile has been modeled in RELAP5 to facilitate steady state simulation of the core cooling, and of anticipated and unanticipated transients. Steady state results are presented to validate the new thermal power profile inputs. A power ramp, slow depressurization at the outlet, and flow coast down transients are also evaluated. (authors)

  3. High Flux Isotope Reactor cold neutron source reference design concept

    SciTech Connect (OSTI)

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

    1998-05-01

    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.

  4. Spheromak reactor with poloidal flux-amplifying transformer

    DOE Patents [OSTI]

    Furth, Harold P.; Janos, Alan C.; Uyama, Tadao; Yamada, Masaaki

    1987-01-01

    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.

  5. Bayesian calibration of reactor neutron flux spectrum using activation detectors measurements: Application to CALIBAN reactor

    SciTech Connect (OSTI)

    Cartier, J.; Casoli, P.; Chappert, F.

    2013-07-01

    In this paper, we present calibration methods in order to estimate reactor neutron flux spectrum and its uncertainties by using integral activation measurements. These techniques are performed using Bayesian and MCMC framework. These methods are applied to integral activation experiments in the cavity of the CALIBAN reactor. We estimate the neutron flux and its related uncertainties. The originality of this work is that these uncertainties take into account measurements uncertainties, cross-sections uncertainties and model error. In particular, our results give a very good approximation of the total flux and indicate that neutron flux from MCNP simulation for energies above about 5 MeV seems to overestimate the 'real flux'. (authors)

  6. Neutron flux profile monitor for use in a fission reactor

    DOE Patents [OSTI]

    Kopp, Manfred K.; Valentine, Kenneth H.

    1983-01-01

    A neutron flux monitor is provided which consists of a plurality of fission counters arranged as spaced-apart point detectors along a delay line. As a fission event occurs in any one of the counters, two delayed current pulses are generated at the output of the delay line. The time separation of the pulses identifies the counter in which the particular fission event occured. Neutron flux profiles of reactor cores can be more accurately measured as a result.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. High Flux Metallic Membranes for Hydrogen Recovery and Membrane Reactors

    SciTech Connect (OSTI)

    Buxbaum, Robert

    2010-06-30

    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.

  12. Performance and safety parameters for the high flux isotope reactor

    SciTech Connect (OSTI)

    Ilas, G. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831-6172 (United States); Primm III, T. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831-6172 (United States); Primm Consulting, LLC, 945 Laurel Hill Road, Knoxville, TN 37923 (United States)

    2012-07-01

    A Monte Carlo depletion model for the High Flux Isotope Reactor (HFIR) Cycle 400 and its use in calculating parameters of relevance to the reactor performance and safety during the reactor cycle are presented in this paper. This depletion model was developed to serve as a reference for the design of a low-enriched uranium (LEU) fuel for an ongoing study to convert HFIR from high-enriched uranium (HEU) to LEU fuel; both HEU and LEU depletion models use the same methodology and ENDF/B-VII nuclear data as discussed in this paper. The calculated HFIR Cycle 400 parameters, which are compared with measurement data from critical experiments performed at HFIR, data included in the HFIR Safety Analysis Report (SAR), or data reported by previous calculations, provide a basis for verification or updating of the corresponding SAR data. (authors)

  13. Performance and Safety Parameters for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Ilas, Germina [ORNL; Primm, Trent [Primm Consulting, LLC

    2012-01-01

    A Monte Carlo depletion model for the High Flux Isotope Reactor (HFIR) Cycle 400 and its use in calculating parameters of relevance to the reactor performance and safety during the reactor cycle are presented in this paper. This depletion model was developed to serve as a reference for the design of a low-enriched uranium (LEU) fuel for an ongoing study to convert HFIR from high-enriched uranium (HEU) to LEU fuel; both HEU and LEU depletion models use the same methodology and ENDV/B-VII nuclear data as discussed in this paper. The calculated HFIR Cycle 400 parameters, which are compared when available with measurement data from critical experiments performed at HFIR, data included in the HFIR Safety Analysis Report (SAR), or data reported by previous calculations, provide a basis for verification or updating of the corresponding SAR data.

  14. Integral window/photon beam position monitor and beam flux detectors for x-ray beams

    DOE Patents [OSTI]

    Shu, Deming; Kuzay, Tuncer M.

    1995-01-01

    A monitor/detector assembly in a synchrotron for either monitoring the position of a photon beam or detecting beam flux may additionally function as a vacuum barrier between the front end and downstream segment of the beamline in the synchrotron. A base flange of the monitor/detector assembly is formed of oxygen free copper with a central opening covered by a window foil that is fused thereon. The window foil is made of man-made materials, such as chemical vapor deposition diamond or cubic boron nitrate and in certain configurations includes a central opening through which the beams are transmitted. Sensors of low atomic number materials, such as aluminum or beryllium, are laid on the window foil. The configuration of the sensors on the window foil may be varied depending on the function to be performed. A contact plate of insulating material, such as aluminum oxide, is secured to the base flange and is thereby clamped against the sensor on the window foil. The sensor is coupled to external electronic signal processing devices via a gold or silver lead printed onto the contact plate and a copper post screw or alternatively via a copper screw and a copper spring that can be inserted through the contact plate and coupled to the sensors. In an alternate embodiment of the monitor/detector assembly, the sensors are sandwiched between the window foil of chemical vapor deposition diamond or cubic boron nitrate and a front foil made of similar material.

  15. Temperature measurements during high flux ion beam irradiations

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Crespillo, Miguel L.; Graham, Joseph T.; Zhang, Yanwen; Weber, William J.

    2016-02-16

    A systematic study of the ion beam heating effect was performed in a temperature range of –170 to 900 °C using a 10 MeV Au3+ ion beam and a Yttria stabilized Zirconia (YSZ) sample at a flux of 5.5 × 1012 cm–2 s–1. Different geometric configurations of beam, sample, thermocouple positioning, and sample holder were compared to understand the heat/charge transport mechanisms responsible for the observed temperature increase. The beam heating exhibited a strong dependence on the background (initial) sample temperature with the largest temperature increases occurring at cryogenic temperatures and decreasing with increasing temperature. Comparison with numerical calculations suggestsmore » that the observed heating effect is, in reality, a predominantly electronic effect and the true temperature rise is small. Furthermore, a simple model was developed to explain this electronic effect in terms of an electrostatic potential that forms during ion irradiation. Such an artificial beam heating effect is potentially problematic in thermostated ion irradiation and ion beamanalysis apparatus, as the operation of temperature feedback systems can be significantly distorted by this effect.« less

  16. Fabrication of control rods for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Sease, J.D.

    1998-03-01

    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.

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

    Office of Science (SC) Website

    Facilities » High Flux Isotope Reactor (HFIR) Scientific User Facilities (SUF) Division SUF Home About User Facilities X-Ray Light Sources Neutron Scattering Facilities Spallation Neutron Source (SNS) High Flux Isotope Reactor (HFIR) Nanoscale Science Research Centers (NSRCs) Projects Accelerator & Detector Research Science Highlights Principal Investigators' Meetings BES Home Neutron Scattering Facilities High Flux Isotope Reactor (HFIR) Print Text Size: A A A FeedbackShare Page Quick

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

    SciTech Connect (OSTI)

    Ilas, Germina [ORNL; Primm, Trent [ORNL

    2011-05-01

    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.

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

    SciTech Connect (OSTI)

    Pierce, B.F.

    1986-07-01

    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.

  20. High Flux Isotope Reactor system RELAP5 input model

    SciTech Connect (OSTI)

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

    1993-01-01

    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.

  1. Progress on the realization of a new GEM based neutron diagnostic concept for high flux neutron beams

    SciTech Connect (OSTI)

    Croci, G.; Tardocchi, M.; Rebai, M.; Cippo, E. Perelli; Gorini, G.; Cazzaniga, C.; Palma, M. Dalla; Pasqualotto, R.; Tollin, M.; Grosso, G.; Muraro, A.; Murtas, F.; Claps, G.; Cavenago, M.

    2014-08-21

    Fusion reactors will need high flux neutron detectors to diagnose the deuterium-deuterium and deuterium-tritium. A candidate detection technique is the Gas Electron Multiplier (GEM). New GEM based detectors are being developed for application to a neutral deuterium beam test facility. The proposed detection system is called Close-contact Neutron Emission Surface Mapping (CNESM). The diagnostic aims at providing the map of the neutron emission due to interaction of the deuterium beam with the deuterons implanted in the beam dump surface. This is done by placing a detector in close contact, right behind the dump. CNESM uses nGEM detectors, i.e. GEM detectors equipped with a cathode that also serves as neutron-proton converter foil. After the realization and test of several small area prototypes, a full size prototype has been realized and tested with laboratory sources. Test on neutron beams are foreseen for the next months.

  2. Fuel Grading Study on a Low-Enriched Uranium Fuel Design for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Ilas, Germina; Primm, Trent

    2009-11-01

    An engineering design study that would enable the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium to low-enriched uranium fuel is ongoing at Oak Ridge National Laboratory. The computational models used to search for a low-enriched uranium (LEU) fuel design that would meet the requirements for the conversion study, and the recent results obtained with these models during FY 2009, 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 high-enriched uranium fuel core. These studies 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.

  3. Fast Flux Test Facility Reactor Vessel Removal Study

    SciTech Connect (OSTI)

    BOWMAN, B.R.

    2002-10-23

    This study assesses the feasibility of removing the FFTF reactor vessel from its current location in the reactor cavity inside the Containment vessel to a transporter for relocation to a burial pit in the 200 Area.

  4. Emulation of reactor irradiation damage using ion beams

    SciTech Connect (OSTI)

    Was, G. S.; Jiao, Z.; Getto, E.; Sun, K.; Monterrosa, A. M.; Maloy, S. A.; Anderoglu, O.; Sencer, B. H.; Hackett, M.

    2014-06-14

    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 irradiation 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 irradiation establishes the capability of tailoring ion irradiation to emulate the reactor-irradiated microstructure.

  5. Emulation of reactor irradiation damage using ion beams

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Was, G. S.; Jiao, Z.; Getto, E.; Sun, K.; Monterrosa, A. M.; Maloy, S. A.; Anderoglu, O.; Sencer, B. H.; Hackett, M.

    2014-06-14

    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,more » 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 irradiation 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 irradiation establishes the capability of tailoring ion irradiation to emulate the reactor-irradiated microstructure.« less

  6. Emulation of reactor irradiation damage using ion beams

    SciTech Connect (OSTI)

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

    2014-10-01

    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 443C 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 460C 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.

  7. Study of a multi-beam accelerator driven thorium reactor

    SciTech Connect (OSTI)

    Ludewig, H.; Aronson, A.

    2011-03-01

    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

  8. Fast flux test reactor fuel canister. (Journal Article) | SciTech...

    Office of Scientific and Technical Information (OSTI)

    Fast flux test reactor fuel canister. Citation Details ... 952779 Report Number(s): SAND2004-2604J TRN: US0902577 DOE Contract Number: AC04-94AL85000 Resource Type: Journal ...

  9. Decomposing VOCs with an electron-beam plasma reactor

    SciTech Connect (OSTI)

    Vitale, S.A.; Hadidi, K.; Cohn, D.R.; Bromberg, L.; Falkos, P.

    1996-04-01

    Several emerging technologies are being tested for decomposing VOCs. Among these are bioreactors, catalytic oxidation, photoinduced decomposition, thermal plasma processes, and nonequilibrium plasma processes. For a new technology to be successfully offered for commercial use, it must be reliable, economically competitive, and ready for use on an industrial scale. The authors have been working on an electron-beam-generated plasma reactor that now meets these prerequisites. The reactor can decompose halogenated organic compounds in dilute concentrations 1--3,000 ppm in airstreams at atmospheric pressure. The technology is more energy efficient than thermal technologies and thus represents lower electricity costs for the overall process. The reactor can easily be scaled to industrial needs and was tested successfully in Hanford, WA, to treat the offgas from the remediation of soils contaminated with CCl{sub 4}.

  10. Plasma focus ion beam fluence and fluxFor various gases

    SciTech Connect (OSTI)

    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)

    2013-06-15

    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.

  11. Homogeneous fast-flux isotope-production reactor

    DOE Patents [OSTI]

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

    1982-08-19

    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.

  12. High heat flux testing of HIP bonded DS-Cu/316SS first wall panel for fusion experimental reactors

    SciTech Connect (OSTI)

    Hatano, Toshihisa; Sato, Kazuyoshi; Dairaku, Masayuki

    1996-12-31

    A shielding blanket design in a fusion reactor such as ITER has been proposed to be a modulator structure integrated with the first wall. In terms of the fabrication, HIP (Hot Isostatic Pressing) method has been proposed for the joining of dispersion strengthened copper (DS-Cu) and type 316L stainless steel (SS316L) at FW. High heat flux tests of HIP bonded DS-Cu/SS316L first wall panel were performed at particle Beam Engineering Facility in JAERI to investigate its thermo-mechanical performance. After four campaigns of high heat flux testing, the FW panel was cut to observe the HIP bonded interface and heated surface of DS-Cu. Though melting of DS-Cu surface was observed, there were no cracks at the HIP bonded interface. 2 refs., 11 figs., 1 tab.

  13. Neutron-flux profile monitor for use in a fission reactor

    DOE Patents [OSTI]

    Kopp, M.K.; Valentine, K.H.

    1981-09-15

    A neutron flux monitor is provided which consists of a plurality of fission counters arranged as spaced-apart point detectors along a delay line. As a fission event occurs in any one of the counters, two delayed current pulses are generated at the output of the delay line. The time separation of the pulses identifies the counter in which the particular fission event occurred. Neutron flux profiles of reactor cores can be more accurately measured as a result.

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

    DOE Patents [OSTI]

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

    2003-06-24

    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.

  15. Control Means for Reactor

    DOE Patents [OSTI]

    Manley, J. H.

    1961-06-27

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

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

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

    Office of Energy Efficiency and Renewable Energy (EERE)

    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.

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

    Office of Energy Efficiency and Renewable Energy (EERE)

    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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

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

    SciTech Connect (OSTI)

    Cook, David Howard

    2009-01-01

    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.

  15. The ORNL High Flux Isotope Reactor and New Advanced Fuel Testing Capabilities

    SciTech Connect (OSTI)

    Ott, Larry J; McDuffee, Joel Lee

    2011-01-01

    The U.S. Department of Energy s High Flux Isotope Reactor (HFIR), located at the Oak Ridge National Laboratory (ORNL), was originally designed (in the 1960s) primarily as a part of the overall program to produce transuranic isotopes for use in the heavy-element research program of the United States. Today, the reactor is a highly versatile machine, producing medical and transuranic isotopes and performing materials test experimental irradiations and neutron-scattering experiments. The ability to test advanced fuels and cladding materials in a thermal neutron spectrum in the United States is limited, and a fast-spectrum irradiation facility does not currently exist in this country. The HFIR has a distinct advantage for consideration as a fuel/cladding irradiation facility because of the extremely high neutron fluxes that this reactor provides over the full thermal- to fast-neutron energy range. New test capabilities have been developed that will allow testing of advanced nuclear fuels and cladding materials in the HFIR under prototypic light-water reactor (LWR) and fast-reactor (FR) operating conditions.

  16. Radiation dosimetry at the BNL reactor facilities

    SciTech Connect (OSTI)

    Holden, N.E.; Hu, J.P.; Reciniello, R.N.; Greenberg, D.D.; Sengupta, S.; Farrell, K.; Greenwood, L.R.

    1999-07-01

    Neutron and gamma-ray dosimetry measurements have been performed at various facilities in the High Flux Beam Reactor (HFBR) and in the Brookhaven National Laboratory Medical Research Reactor (BMRR). These experimental results are discussed.

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

    SciTech Connect (OSTI)

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

    2010-08-19

    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

  18. OSTIblog Articles in the High Flux Isotope Reactor Topic | OSTI, US Dept of

    Office of Scientific and Technical Information (OSTI)

    Energy Office of Scientific and Technical Information High Flux Isotope Reactor Topic The NXS Class of 2014 by Kathy Chambers 19 Nov, 2014 in Every summer for the past 16 years, the Department of Energy has invited the best and brightest graduates from across the country to attend the National School on Neutron and X-ray Scattering (NXS). This year, 65 graduate students attending North American universities, and studying physics, chemistry, materials science, or related fields, participated

  19. Measurement of the reactor antineutrino flux and spectrum at Daya Bay

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    D. E. Jaffe; Bishai, M; Diwan, M.; Gill, R.; Hackenburg, R. W.; Hans, S.; Hu, L. M.; Jaffe, D. E.; Kettell, S. H.; Tang, W.; et al

    2016-02-12

    This Letter reports a measurement of the flux and energy spectrum of electron antineutrinos from six 2.9~GWth nuclear reactors with six detectors deployed in two near (effective baselines 512~m and 561~m) and one far (1,579 m) underground experimental halls in the Daya Bay experiment. Using 217 days of data, 296,721 and 41,589 inverse beta decay (IBD) candidates were detected in the near and far halls, respectively. The measured IBD yield is (1.55 ± 0.04) × 10–18 cm2/GW/day or (5.92 ± 0.14) × 10–43 cm2/fission. This flux measurement is consistent with previous short-baseline reactor antineutrino experiments and is 0.946 ± 0.022more » (0.991 ± 0.023) relative to the flux predicted with the Huber+Mueller (ILL+Vogel) fissile antineutrino model. The measured IBD positron energy spectrum deviates from both spectral predictions by more than 2σ over the full energy range with a local significance of up to ~4σ between 4-6 MeV. Furthermore, a reactor antineutrino spectrum of IBD reactions is extracted from the measured positron energy spectrum for model-independent predictions.« less

  20. Advanced Multiphysics Thermal-Hydraulics Models for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Jain, Prashant K; Freels, James D

    2015-01-01

    Engineering design studies to determine the feasibility of converting the High Flux Isotope Reactor (HFIR) from using highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL). This work is part of an effort sponsored by the US Department of Energy (DOE) Reactor Conversion Program. HFIR is a very high flux pressurized light-water-cooled and moderated flux-trap type research reactor. HFIR s current missions are to support neutron scattering experiments, isotope production, and materials irradiation, including neutron activation analysis. Advanced three-dimensional multiphysics models of HFIR fuel were developed in COMSOL software for safety basis (worst case) operating conditions. Several types of physics including multilayer heat conduction, conjugate heat transfer, turbulent flows (RANS model) and structural mechanics were combined and solved for HFIR s inner and outer fuel elements. Alternate design features of the new LEU fuel were evaluated using these multiphysics models. This work led to a new, preliminary reference LEU design that combines a permanent absorber in the lower unfueled region of all of the fuel plates, a burnable absorber in the inner element side plates, and a relocated and reshaped (but still radially contoured) fuel zone. Preliminary results of estimated thermal safety margins are presented. Fuel design studies and model enhancement continue.

  1. The wave energy flux of high frequency diffracting beams in complex geometrical optics

    SciTech Connect (OSTI)

    Maj, Omar; Poli, Emanuele; Mariani, Alberto; Farina, Daniela

    2013-04-15

    We consider the construction of asymptotic solutions of Maxwell's equations for a diffracting wave beam in the high frequency limit and address the description of the wave energy flux transported by the beam. With this aim, the complex eikonal method is applied. That is a generalization of the standard geometrical optics method in which the phase function is assumed to be complex valued, with the non-negative imaginary part accounting for the finite width of the beam cross section. In this framework, we propose an argument which simplifies significantly the analysis of the transport equation for the wave field amplitude and allows us to derive the wave energy flux. The theoretical analysis is illustrated numerically for the case of electron cyclotron beams in tokamak plasmas by using the GRAY code [D. Farina, Fusion Sci. Technol. 52, 154 (2007)], which is based upon the complex eikonal theory. The results are compared to those of the paraxial beam tracing code TORBEAM [E. Poli et al., Comput. Phys. Commun. 136, 90 (2001)], which provides an independent calculation of the energy flow.

  2. PREPARING THE HIGH FLUX ISOTOPE REACTOR FOR CONVERSION TO LOW ENRICHED URANIUM FUEL ? RETURN TO 100 MW

    SciTech Connect (OSTI)

    Smith, Kevin Arthur [ORNL; Primm, Trent [ORNL

    2009-01-01

    The feasibility of low-enriched uranium (LEU) fuel as a replacement for the current, high enriched uranium (HEU) fuel for the High Flux Isotope Reactor (HFIR) has been under study since 2006. Reactor performance studies have been completed for conceptual plate designs and show that maintaining reactor performance while converting to LEU fuel requires returning the reactor power to 100 MW from 85 MW. The analyses required to up-rate the reactor power and the methods to perform these analyses are discussed. Comments regarding the regulatory approval process are provided along with a conceptual schedule.

  3. Optimization of Depletion Modeling and Simulation for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Betzler, Benjamin R; Ade, Brian J; Chandler, David; Ilas, Germina; Sunny, Eva E

    2015-01-01

    Monte Carlo based depletion tools used for the high-fidelity modeling and simulation of the High Flux Isotope Reactor (HFIR) come at a great computational cost; finding sufficient approximations is necessary to make the use of these tools feasible. The optimization of the neutronics and depletion model for the HFIR is based on two factors: (i) the explicit representation of the involute fuel plates with sets of polyhedra and (ii) the treatment of depletion mixtures and control element position during depletion calculations. A very fine representation (i.e., more polyhedra in the involute plate approximation) does not significantly improve simulation accuracy. The recommended representation closely represents the physical plates and ensures sufficient fidelity in regions with high flux gradients. Including the fissile targets in the central flux trap of the reactor as depletion mixtures has the greatest effect on the calculated cycle length, while localized effects (e.g., the burnup of specific isotopes or the power distribution evolution over the cycle) are more noticeable consequences of including a critical control element search or depleting burnable absorbers outside the fuel region.

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

    SciTech Connect (OSTI)

    McAdams, R.

    2014-02-15

    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.

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

    SciTech Connect (OSTI)

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

    2010-10-01

    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.

  6. Development of a Scale Model for High Flux Isotope Reactor Cycle 400

    SciTech Connect (OSTI)

    Ilas, Dan

    2012-03-01

    The development of a comprehensive SCALE computational model for the High Flux Isotope Reactor (HFIR) is documented and discussed in this report. The SCALE model has equivalent features and functionality as the reference MCNP model for Cycle 400 that has been used extensively for HFIR safety analyses and for HFIR experiment design and analyses. Numerical comparisons of the SCALE and MCNP models for the multiplication constant, power density distribution in the fuel, and neutron fluxes at several locations in HFIR indicate excellent agreement between the results predicted with the two models. The SCALE HFIR model is presented in sufficient detail to provide the users of the model with a tool that can be easily customized for various safety analysis or experiment design requirements.

  7. Neutron scattering at the high flux isotope reactor at Oak Ridge National Laboratory

    SciTech Connect (OSTI)

    Yethiraj, M.; Fernandez-Baca, J.A.

    1995-03-01

    Since its beginnings in Oak Ridge and Argonne in the late 1940`s, neutron scattering has been established as the premier tool to study matter in its various states. Since the thermal neutron wavelength is of the same order of magnitude as typical atomic spacings and because they have comparable energies to those of atomic excitations in solids, both structure and dynamics of matter can be studied via neutron scattering. The High Flux Isotope Reactor (HFIR) provides an intense source of neutrons with which to carry out these measurements. This paper summarizes the available neutron scattering facilities at the HFIR.

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

    SciTech Connect (OSTI)

    Primm, R.T., III

    2003-11-01

    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.

  9. Decomposition of chlorinated ethylenes and ethanes in an electron beam generated plasma reactor

    SciTech Connect (OSTI)

    Vitale, S.A.

    1996-02-01

    An electron beam generated plasma reactor (EBGPR) is used to determine the plasma chemistry kinetics, energetics and decomposition pathways of six chlorinated ethylenes and ethanes: 1,1,1-trichloroethane, 1,1-dichloroethane, ethyl chloride, trichloroethylene, 1,1-dichloroethylene, and vinyl chloride. A traditional chemical kinetic and chemical engineering analysis of the data from the EBGPR is performed, and the following hypothesis was verified: The specific energy required for chlorinated VOC decomposition in the electron beam generated plasma reactor is determined by the electron attachment coefficient of the VOC and the susceptibility of the molecule to radical attack. The technology was demonstrated at the Hanford Reservation to remove VOCs from soils.

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

    SciTech Connect (OSTI)

    Hawari, Ayman I.; Moxom, Jeremy; Hathaway, Alfred G.; Brown, Benjamin; Xu, Jun

    2009-03-10

    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.

  11. High Flux Isotope Reactor Core Analysis-Challenges and Recent Enhancements in Modeling and Simulation

    SciTech Connect (OSTI)

    Ilas, Germina

    2016-01-01

    A concerted effort over the past few years has focused on enhancing the core depletion models for the High Flux Isotope Reactor (HFIR) as part of a comprehensive study for designing a HFIR core that would use low-enriched uranium (LEU) fuel. A HFIR core depletion model that is based on current state-of-the-art methods and nuclear data was needed for use as a reference for the design of an LEU fuel for HFIR and to improve the basis for analyses that support HFIR s current operation with high-enriched uranium (HEU) fuel. This paper summarizes the recent improvements in modeling and simulation for HFIR core analyses, with a focus on core depletion models.

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

    SciTech Connect (OSTI)

    Lastres, Oscar; Chandler, David; Jarrell, Joshua J; Maldonado, G. Ivan

    2011-01-01

    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

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

    SciTech Connect (OSTI)

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

    2014-09-01

    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.

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

    SciTech Connect (OSTI)

    Sozer, M.C.

    1992-04-01

    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.

  15. High active nitrogen flux growth of GaN by plasma assisted molecular beam epitaxy

    SciTech Connect (OSTI)

    McSkimming, Brian M. Speck, James S.; Chaix, Catherine

    2015-09-15

    In the present study, the authors report on a modified Riber radio frequency (RF) nitrogen plasma source that provides active nitrogen fluxes more than 30 times higher than those commonly used for plasma assisted molecular beam epitaxy (PAMBE) growth of gallium nitride (GaN) and thus a significantly higher growth rate than has been previously reported. GaN films were grown using N{sub 2} gas flow rates between 5 and 25 sccm while varying the plasma source's RF forward power from 200 to 600 W. The highest growth rate, and therefore the highest active nitrogen flux, achieved was ∼7.6 μm/h. For optimized growth conditions, the surfaces displayed a clear step-terrace structure with an average RMS roughness (3 × 3 μm) on the order of 1 nm. Secondary ion mass spectroscopy impurity analysis demonstrates oxygen and hydrogen incorporation of 1 × 10{sup 16} and ∼5 × 10{sup 17}, respectively. In addition, the authors have achieved PAMBE growth of GaN at a substrate temperature more than 150 °C greater than our standard Ga rich GaN growth regime and ∼100 °C greater than any previously reported PAMBE growth of GaN. This growth temperature corresponds to GaN decomposition in vacuum of more than 20 nm/min; a regime previously unattainable with conventional nitrogen plasma sources. Arrhenius analysis of the decomposition rate shows that samples with a flux ratio below stoichiometry have an activation energy greater than decomposition of GaN in vacuum while samples grown at or above stoichiometry have decreased activation energy. The activation energy of decomposition for GaN in vacuum was previously determined to be ∼3.1 eV. For a Ga/N flux ratio of ∼1.5, this activation energy was found to be ∼2.8 eV, while for a Ga/N flux ratio of ∼0.5, it was found to be ∼7.9 eV.

  16. Ion species control in high flux deuterium plasma beams produced by a linear plasma generator

    SciTech Connect (OSTI)

    Luo, G.-N.; Shu, W.M.; Nakamura, H.; O'Hira, S.; Nishi, M.

    2004-11-01

    The ion species ratios in low energy high flux deuterium plasma beams formed in a linear plasma generator were measured by a quadrupole mass spectrometer. And the species control in the plasma generator was evaluated by changing the operational parameters like neutral pressure, arc current, and axial magnetic confinement to the plasma column. The measurements reveal that the lower pressures prefer to form more D{sup +} ions, and the medium magnetic confinement at the higher pressures results in production of more D{sub 2}{sup +}, while the stronger confinement and/or larger arc current are helpful to D{sub 2}{sup +} conversion into D{sub 3}{sup +}. Therefore, the ion species can be controlled by adjusting the operational parameters of the plasma generator. With suitable adjustment, we can achieve plasma beams highly enriched with a single species of D{sup +}, D{sub 2}{sup +}, or D{sub 3}{sup +}, to a ratio over 80%. It has been found that the axial magnetic configuration played a significant role in the formation of D{sub 3}{sup +} within the experimental pressure range.

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

    SciTech Connect (OSTI)

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

    2009-12-01

    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.

  18. Simulating High Flux Isotope Reactor Core Thermal-Hydraulics via Interdimensional Model Coupling

    SciTech Connect (OSTI)

    Travis, Adam R

    2014-05-01

    A coupled interdimensional model is presented for the simulation of the thermal-hydraulic characteristics of the High Flux Isotope Reactor core at Oak Ridge National Laboratory. The model consists of two domains a solid involute fuel plate and the surrounding liquid coolant channel. The fuel plate is modeled explicitly in three-dimensions. The coolant channel is approximated as a twodimensional slice oriented perpendicular to the fuel plate s surface. The two dimensionally-inconsistent domains are linked to one another via interdimensional model coupling mechanisms. The coupled model is presented as a simplified alternative to a fully explicit, fully three-dimensional model. Involute geometries were constructed in SolidWorks. Derivations of the involute construction equations are presented. Geometries were then imported into COMSOL Multiphysics for simulation and modeling. Both models are described in detail so as to highlight their respective attributes in the 3D model, the pursuit of an accurate, reliable, and complete solution; in the coupled model, the intent to simplify the modeling domain as much as possible without affecting significant alterations to the solution. The coupled model was created with the goal of permitting larger portions of the reactor core to be modeled at once without a significant sacrifice to solution integrity. As such, particular care is given to validating incorporated model simplifications. To the greatest extent possible, the decrease in solution time as well as computational cost are quantified versus the effects such gains have on the solution quality. A variant of the coupled model which sufficiently balances these three solution characteristics is presented alongside the more comprehensive 3D model for comparison and validation.

  19. Modeling and Simulations for the High Flux Isotope Reactor Cycle 400

    SciTech Connect (OSTI)

    Ilas, Germina; Chandler, David; Ade, Brian J; Sunny, Eva E; Betzler, Benjamin R; Pinkston, Daniel

    2015-03-01

    A concerted effort over the past few years has been focused on enhancing the core model for the High Flux Isotope Reactor (HFIR), as part of a comprehensive study for HFIR conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel. At this time, the core model used to perform analyses in support of HFIR operation is an MCNP model for the beginning of Cycle 400, which was documented in detail in a 2005 technical report. A HFIR core depletion model that is based on current state-of-the-art methods and nuclear data was needed to serve as reference for the design of an LEU fuel for HFIR. The recent enhancements in modeling and simulations for HFIR that are discussed in the present report include: (1) revision of the 2005 MCNP model for the beginning of Cycle 400 to improve the modeling data and assumptions as necessary based on appropriate primary reference sources HFIR drawings and reports; (2) improvement of the fuel region model, including an explicit representation for the involute fuel plate geometry that is characteristic to HFIR fuel; and (3) revision of the Monte Carlo-based depletion model for HFIR in use since 2009 but never documented in detail, with the development of a new depletion model for the HFIR explicit fuel plate representation. The new HFIR models for Cycle 400 are used to determine various metrics of relevance to reactor performance and safety assessments. The calculated metrics are compared, where possible, with measurement data from preconstruction critical experiments at HFIR, data included in the current HFIR safety analysis report, and/or data from previous calculations performed with different methods or codes. The results of the analyses show that the models presented in this report provide a robust and reliable basis for HFIR analyses.

  20. Utilization of the High Flux Isotope Reactor at Oak Ridge National Laboratory

    SciTech Connect (OSTI)

    Selby, Douglas L; Bilheux, Hassina Z; Meilleur, Flora; Jones, Amy; Bailey, William Barton; Vandergriff, David H

    2015-01-01

    This paper addresses several aspects of the scientific utilization of the Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR). Topics to be covered will include: 1) HFIR neutron scattering instruments and the formal instrument user program; 2) Recent upgrades to the neutron scattering instrument stations at the reactor, and 3) eMod a new tool for addressing instrument modifications and providing configuration control and design process for scientific instruments at HFIR and the Spallation Neutron Source (SNS). There are 15 operating neutron instrument stations at HFIR with 12 of them organized into a formal user program. Since the last presentation on HFIR instruments at IGORR we have installed a Single Crystal Quasi-Laue Diffractometer instrument called IMAGINE; and we have made significant upgrades to HFIR neutron scattering instruments including the Cold Triple Axis Instrument, the Wide Angle Neutron Diffractometer, the Powder Diffractometer, and the Neutron Imaging station. In addition, we have initiated upgrades to the Thermal Triple Axis Instrument and the Bio-SANS cold neutron instrument detector system. All of these upgrades are tied to a continuous effort to maintain a high level neutron scattering user program at the HFIR. For the purpose of tracking modifications such as those mentioned and configuration control we have been developing an electronic system for entering instrument modification requests that follows a modification or instrument project through concept development, design, fabrication, installation, and commissioning. This system, which we call eMod, electronically leads the task leader through a series of questions and checklists that then identifies such things as ES&H and radiological issues and then automatically designates specific individuals for the activity review process. The system has been in use for less than a year and we are still working out some of the inefficiencies, but we believe that this will become a very

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

    SciTech Connect (OSTI)

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

    2013-11-01

    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 -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.

  2. Delivery of completed irradiation vehicles and the quality assurance document to the High Flux Isotope Reactor for irradiation

    SciTech Connect (OSTI)

    Petrie, Christian M.; McDuffee, Joel Lee; Katoh, Yutai; Terrani, Kurt A.

    2015-10-01

    This report details the initial fabrication and delivery of two Fuel Cycle Research and Development (FCRD) irradiation capsules (ATFSC01 and ATFSC02), with associated quality assurance documentation, to the High Flux Isotope Reactor (HFIR). The capsules and documentation were delivered by September 30, 2015, thus meeting the deadline for milestone M3FT-15OR0202268. These irradiation experiments are testing silicon carbide composite tubes in order to obtain experimental validation of thermo-mechanical models of stress states in SiC cladding irradiated under a prototypic high heat flux. This document contains a copy of the completed capsule fabrication request sheets, which detail all constituent components, pertinent drawings, etc., along with a detailed summary of the capsule assembly process performed by the Thermal Hydraulics and Irradiation Engineering Group (THIEG) in the Reactor and Nuclear Systems Division (RNSD). A complete fabrication package record is maintained by the THIEG and is available upon request.

  3. Total absorption spectroscopy study of ?Rb decay: A major contributor to reactor antineutrino spectrum shape [Total absorption spectroscopy study of ?Rb: A major contributor to reactor antineutrino flux

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Sonzogni, A.; Zakari-Issoufou, A. -A.; Fallot, M.; Porta, A.; Algora, A.; Tain, J. L.; Valencia, E.; Rice, S.; Bui, V. M.; Cormon, S.; et al

    2015-03-09

    The accurate determination of the emitted reactor antineutrino flux is still a major challenge for actual and future neutrino experiments at reactors, especially after the evidence of a disagreement between the measured antineutrino energy spectrum by Double Chooz, Daya Bay, and Reno and calculated antineutrino spectra obtained from the conversion of the unique integral beta spectra measured at the ILL reactor. Using nuclear data to compute reactor antineutrino spectra may help understanding this bias, with the study of the underlying nuclear physics. Summation calculations allow identifying a list of nuclei that contribute importantly to the antineutrino energy spectra emitted aftermorethe fission of ?,?Pu and ?,?U, and whose beta decay properties might deserve new measurements. Among these nuclei, ?Rb exhausts by itself about 16% of of the antineutrino energy spectrum emitted by Pressurized Water Reactors in the 5 to 8 MeV range. In this Letter, we report new Total Absorption Spectroscopy (TAS) results for this important contributor. The obtained beta feeding from ?Rb shows beta intensity unobserved before in the 4.5 to 5.5 MeV energy region and gives a ground state to ground state branch of 87.5 % 3%. These new data induce a dramatic change in recent summation calculations where a 51% GS to GS branch was considered for ?Rb, increasing the summation antineutrino spectrum in the region nearby the observed bias.The new data still have an important impact on other summation calculations in which more recent data were consideredless

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

    SciTech Connect (OSTI)

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

    2010-02-01

    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.

  5. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2008

    SciTech Connect (OSTI)

    Primm, Trent [ORNL; Chandler, David [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Sease, John D [ORNL; Jolly, Brian C [ORNL

    2009-03-01

    This report documents progress made during FY 2008 in studies of converting the High Flux Isotope Reactor (HFIR) from highly 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. Scoping experiments with various manufacturing methods for forming the LEU alloy profile are presented.

  6. High-flux low-divergence positron beam generation from ultra-intense laser irradiated a tapered hollow target

    SciTech Connect (OSTI)

    Liu, Jian-Xun; Ma, Yan-Yun; Zhao, Jun; Yu, Tong-Pu Yang, Xiao-Hu; Gan, Long-Fei; Zhang, Guo-Bo; Yan, Jian-Feng; Zhuo, Hong-Bin; Liu, Jin-Jin; Zhao, Yuan; Kawata, Shigeo

    2015-10-15

    By using two-dimensional particle-in-cell simulations, we demonstrate high-flux dense positrons generation by irradiating an ultra-intense laser pulse onto a tapered hollow target. By using a laser with an intensity of 4 × 10{sup 23 }W/cm{sup 2}, it is shown that the Breit-Wheeler process dominates the positron production during the laser-target interaction and a positron beam with a total number >10{sup 15} is obtained, which is increased by five orders of magnitude than in the previous work at the same laser intensity. Due to the focusing effect of the transverse electric fields formed in the hollow cone wall, the divergence angle of the positron beam effectively decreases to ∼15° with an effective temperature of ∼674 MeV. When the laser intensity is doubled, both the positron flux (>10{sup 16}) and temperature (963 MeV) increase, while the divergence angle gets smaller (∼13°). The obtained high-flux low-divergence positron beam may have diverse applications in science, medicine, and engineering.

  7. Analyses of the reflector tank, cold source, and beam tube cooling for ANS reactor

    SciTech Connect (OSTI)

    Marland, S.

    1992-07-01

    This report describes my work as an intern with Martin Marietta Energy Systems, Inc., in the summer of 1991. I was assigned to the Reactor Technology Engineering Department, working on the Advanced Neutron Source (ANS). My first project was to select and analyze sealing systems for the top of the diverter/reflector tank. This involved investigating various metal seals and calculating the forces necessary to maintain an adequate seal. The force calculations led to an analysis of several bolt patterns and lockring concepts that could be used to maintain a seal on the vessel. Another project involved some pressure vessel stress calculations and the calculation of the center of gravity for the cold source assembly. I also completed some sketches of possible cooling channel patterns for the inner vessel of the cold source. In addition, I worked on some thermal design analyses for the reflector tank and beam tubes, including heat transfer calculations and assisting in Patran and Pthermal analyses. To supplement the ANS work, I worked on other projects. I completed some stress/deflection analyses on several different beams. These analyses were done with the aid of CAASE, a beam-analysis software package. An additional project involved bending analysis on a carbon removal system. This study was done to find the deflection of a complex-shaped beam when loaded with a full waste can.

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

    SciTech Connect (OSTI)

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

    2014-04-24

    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.

  9. Self-corrected sensors based on atomic absorption spectroscopy for atom flux measurements in molecular beam epitaxy

    SciTech Connect (OSTI)

    Du, Y. E-mail: scott.chambers@pnnl.gov; Liyu, A. V.; Droubay, T. C.; Chambers, S. A. E-mail: scott.chambers@pnnl.gov; Li, G.

    2014-04-21

    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 detector in a double-beam configuration, we employ either a non-resonant line or a resonant line with low cross section 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.

  10. Total absorption spectroscopy study of ?Rb decay: A major contributor to reactor antineutrino spectrum shape [Total absorption spectroscopy study of ?Rb: A major contributor to reactor antineutrino flux

    SciTech Connect (OSTI)

    Sonzogni, A.; Zakari-Issoufou, A. -A.; Fallot, M.; Porta, A.; Algora, A.; Tain, J. L.; Valencia, E.; Rice, S.; Bui, V. M.; Cormon, S.; Estienne, M.; Agramunt, J.; Aysto, J.; Bowry, M.; Briz Monago, J. A.; Caballero-Folch, R.; Cano-Ott, D.; Cucoanes, A.; Eloma, V.; Estvez, E.; Farrelly, G. F.; Garcia, A.; Gelletly, W.; Gomez-Hornillos, M. B.; Gorlychev, V.; Hakala, J.; Jokinen, A.; Jordan, M. D.; Kankainen, A.; Kondev, F. G.; Martinez, T.; Mendoza, E.; Molina, F.; Moore, I.; Perez, A.; Podolyak, Zs.; Penttil, H.; Regan, P. H.; Shiba, T.; Rissanen, J.; Rubio, B.; Weber, C.

    2015-03-09

    The accurate determination of the emitted reactor antineutrino flux is still a major challenge for actual and future neutrino experiments at reactors, especially after the evidence of a disagreement between the measured antineutrino energy spectrum by Double Chooz, Daya Bay, and Reno and calculated antineutrino spectra obtained from the conversion of the unique integral beta spectra measured at the ILL reactor. Using nuclear data to compute reactor antineutrino spectra may help understanding this bias, with the study of the underlying nuclear physics. Summation calculations allow identifying a list of nuclei that contribute importantly to the antineutrino energy spectra emitted after the fission of ?,?Pu and ?,?U, and whose beta decay properties might deserve new measurements. Among these nuclei, ?Rb exhausts by itself about 16% of of the antineutrino energy spectrum emitted by Pressurized Water Reactors in the 5 to 8 MeV range. In this Letter, we report new Total Absorption Spectroscopy (TAS) results for this important contributor. The obtained beta feeding from ?Rb shows beta intensity unobserved before in the 4.5 to 5.5 MeV energy region and gives a ground state to ground state branch of 87.5 % 3%. These new data induce a dramatic change in recent summation calculations where a 51% GS to GS branch was considered for ?Rb, increasing the summation antineutrino spectrum in the region nearby the observed bias.The new data still have an important impact on other summation calculations in which more recent data were considered

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

    SciTech Connect (OSTI)

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

    2011-02-01

    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.

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

    SciTech Connect (OSTI)

    Primm, Trent; Guida, Tracey

    2010-02-01

    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.

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

    SciTech Connect (OSTI)

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

    2007-11-01

    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

  14. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    SciTech Connect (OSTI)

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

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

    SciTech Connect (OSTI)

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

    2012-03-01

    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.

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

    SciTech Connect (OSTI)

    Shibata, Y. Manabe, T.; Ohno, N.; Takagi, M.; Kajita, S.; Tsuchiya, H.; Morisaki, T.

    2014-09-15

    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.

  17. Preliminary Assessment of the Impact on Reactor Vessel dpa Rates Due to Installation of a Proposed Low Enriched Uranium (LEU) Core in the High Flux Isotope Reactor (HFIR)

    SciTech Connect (OSTI)

    Daily, Charles R.

    2015-10-01

    An assessment of the impact on the High Flux Isotope Reactor (HFIR) reactor vessel (RV) displacements-per-atom (dpa) rates due to operations with the proposed low enriched uranium (LEU) core described by Ilas and Primm has been performed and is presented herein. The analyses documented herein support the conclusion that conversion of HFIR to low-enriched uranium (LEU) core operations using the LEU core design of Ilas and Primm will have no negative impact on HFIR RV dpa rates. Since its inception, HFIR has been operated with highly enriched uranium (HEU) cores. As part of an effort sponsored by the National Nuclear Security Administration (NNSA), conversion to LEU cores is being considered for future HFIR operations. The HFIR LEU configurations analyzed are consistent with the LEU core models used by Ilas and Primm and the HEU balance-of-plant models used by Risner and Blakeman in the latest analyses performed to support the HFIR materials surveillance program. The Risner and Blakeman analyses, as well as the studies documented herein, are the first to apply the hybrid transport methods available in the Automated Variance reduction Generator (ADVANTG) code to HFIR RV dpa rate calculations. These calculations have been performed on the Oak Ridge National Laboratory (ORNL) Institutional Cluster (OIC) with version 1.60 of the Monte Carlo N-Particle 5 (MCNP5) computer code.

  18. (Reactor dosimetry)

    SciTech Connect (OSTI)

    West, C.D.

    1990-09-13

    The lead in most aspects of research reactor design and use passed from the USA about 15 years ago, soon after the construction of the HFIR and HFBR. The Europeans have consistently upgraded and improved their existing facilities and have built new ones including the HFR at Grenoble and ORPHEE at Saclay. They studied ultra-high flux concepts ({approximately}10{sup 20}/m{sup {minus}2}{center dot}s{sup {minus}1}) about 10 years ago, and are in the design phase of a new, highly efficient medium flux reactor to be built at Garching, near Munich in Germany. A visit was made to Interatom, the firm -- the equivalent of the Architect/Engineer for the ANS project -- responsible, under contract to the Technical University of Munich, for the new Munich reactor design. There are many similarities to the ANS design, and we reviewed and discussed technical and safety aspects of the two reactors. A request was made for some new, hitherto proprietary, experimental data on reactor thermal hydraulics and cooling that will be very valuable to the ANS project. I presented a seminar on the ANS project. A visit was made to Kernforschungszentrum Karlsruhe and knowledge was gained from Dr. Kuchle, a true pioneer of ultra-high flux reactor concepts, of their work. Dr. Kuchle kindly reviewed the ANS reference core and cooling system design (with favorable conclusions). I then talked with researchers working on materials irradiation damage and activation of structural materials by neutron irradiation, both key issues for the ANS. I was shown some new techniques they have developed for testing materials irradiation effects at high fluences, in a short time, using accelerated particle beams.

  19. An experimental study of external reactor vessel cooling strategy on the critical heat flux using the graphene oxide nano-fluid

    SciTech Connect (OSTI)

    Park, S. D.; Lee, S. W.; Kang, S.; Kim, S. M.; Seo, H.; Bang, I. C.

    2012-07-01

    External reactor vessel cooling (ERVC) for in-vessel retention (IVR) of corium as a key severe accident management strategy can be achieved by flooding the reactor cavity during a severe accident. In this accident mitigation strategy, the decay heat removal capability depends on whether the imposed heat flux exceeds critical heat flux (CHF). To provide sufficient cooling for high-power reactors such as APR1400, there have been some R and D efforts to use the reactor vessel with micro-porous coating and nano-fluids boiling-induced coating. The dispersion stability of graphene-oxide nano-fluid in the chemical conditions of flooding water that includes boric acid, lithium hydroxide (LiOH) and tri-sodium phosphate (TSP) was checked in terms of surface charge or zeta potential before the CHF experiments. Results showed that graphene-oxide nano-fluids were very stable under ERVC environment. The critical heat flux (CHF) on the reactor vessel external wall was measured using the small scale two-dimensional slide test section. The radius of the curvature is 0.1 m. The dimension of each part in the facility simulated the APR-1400. The heater was designed to produce the different heat flux. The magnitude of heat flux follows the one of the APR-1400 when the severe accident occurred. All tests were conducted under inlet subcooling 10 K. Graphene-oxide nano-fluids (concentration: 10 -4 V%) enhanced CHF limits up to about 20% at mass flux 50 kg/m{sup 2}s and 100 kg/m{sup 2}s in comparison with the results of the distilled water at same test condition. (authors)

  20. 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)

    Not Available

    1992-07-01

    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.

  1. Neutronics Conversion Analyses of the Laue-Langevin Institute (ILL) High Flux Reactor (RHF)

    SciTech Connect (OSTI)

    Bergeron, A.; Dionne, B.; Calzavara, Y.

    2014-09-30

    The following report describes the neutronics results obtained with the MCNP model of the RHF U7Mo LEU reference design that has been established in 2010 during the feasibility analysis. This work constitutes a complete and detailed neutronics analysis of that LEU design using models that have been significantly improved since 2010 and the release of the feasibility report. When possible, the credibility of the neutronics model is tested by comparing the HEU model results with experimental data or other codes calculations results. The results obtained with the LEU model are systematically compared to the HEU model. The changes applied to the neutronics model lead to better comparisons with experimental data or improved the calculation efficiency but do not challenge the conclusion of the feasibility analysis. If the U7Mo fuel is commercially available, not cost prohibitive, a back-end solution is established and if it is possible to manufacture the proposed element, neutronics analyses show that the performance of the reactor would not be challenged by the conversion to LEU fuel.

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

    SciTech Connect (OSTI)

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

    2011-09-02

    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.

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

    SciTech Connect (OSTI)

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

    2010-01-01

    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.

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

    SciTech Connect (OSTI)

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

    2006-11-01

    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.

  5. Molecular beam mass spectrometer equipped with a catalytic wall reactor for in situ studies in high temperature catalysis research

    SciTech Connect (OSTI)

    Horn, R.; Ihmann, K.; Ihmann, J.; Jentoft, F.C.; Geske, M.; Taha, A.; Pelzer, K.; Schloegl, R.

    2006-05-15

    A newly developed apparatus combining a molecular beam mass spectrometer and a catalytic wall reactor is described. The setup has been developed for in situ studies of high temperature catalytic reactions (>1000 deg. C), which involve besides surface reactions also gas phase reactions in their mechanism. The goal is to identify gas phase radicals by threshold ionization. A tubular reactor, made from the catalytic material, is positioned in a vacuum chamber. Expansion of the gas through a 100 {mu}m sampling orifice in the reactor wall into differentially pumped nozzle, skimmer, and collimator chambers leads to the formation of a molecular beam. A quadrupole mass spectrometer with electron impact ion source designed for molecular beam inlet and threshold ionization measurements is used as the analyzer. The sampling time from nozzle to detector is estimated to be less than 10 ms. A detection time resolution of up to 20 ms can be reached. The temperature of the reactor is measured by pyrometry. Besides a detailed description of the setup components and the physical background of the method, this article presents measurements showing the performance of the apparatus. After deriving the shape and width of the energy spread of the ionizing electrons from measurements on N{sub 2} and He we estimated the detection limit in threshold ionization measurements using binary mixtures of CO in N{sub 2} to be in the range of several hundreds of ppm. Mass spectra and threshold ionization measurements recorded during catalytic partial oxidation of methane at 1250 deg. C on a Pt catalyst are presented. The detection of CH{sub 3}{center_dot} radicals is successfully demonstrated.

  6. Production of Medical Radioisotopes in the ORNL High Flux Isotope Reactor (HFIR) for Cancer Treatment and Arterial Restenosis Therapy after PTCA

    DOE R&D Accomplishments [OSTI]

    Knapp, F. F. Jr.; Beets, A. L.; Mirzadeh, S.; Alexander, C. W.; Hobbs, R. L.

    1998-06-01

    The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) represents an important resource for the production of a wide variety of medical radioisotopes. In addition to serving as a key production site for californium-252 and other transuranic elements, important examples of therapeutic radioisotopes which are currently routinely produced in the HFIR for distribution include dysprosium-166 (parent of holmium-166), rhenium-186, tin-117m and tungsten-188 (parent of rhenium-188). The nine hydraulic tube (HT) positions in the central high flux region permit the insertion and removal of targets at any time during the operating cycle and have traditionally represented a major site for production of medical radioisotopes. To increase the irradiation capabilities of the HFIR, special target holders have recently been designed and fabricated which will be installed in the six Peripheral Target Positions (PTP), which are also located in the high flux region. These positions are only accessible during reactor refueling and will be used for long-term irradiations, such as required for the production of tin-117m and tungsten-188. Each of the PTP tubes will be capable of housing a maximum of eight HT targets, thus increasing the total maximum number of HT targets from the current nine, to a total of 57. In this paper the therapeutic use of reactor-produced radioisotopes for bone pain palliation and vascular brachytherapy and the therapeutic medical radioisotope production capabilities of the ORNL HFIR are briefly discussed.

  7. REACTOR

    DOE Patents [OSTI]

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  8. REACTOR

    DOE Patents [OSTI]

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  9. Analyses of the reflector tank, cold source, and beam tube cooling for ANS reactor. [Advanced Neutron Source (ANS)

    SciTech Connect (OSTI)

    Marland, S. )

    1992-07-01

    This report describes my work as an intern with Martin Marietta Energy Systems, Inc., in the summer of 1991. I was assigned to the Reactor Technology Engineering Department, working on the Advanced Neutron Source (ANS). My first project was to select and analyze sealing systems for the top of the diverter/reflector tank. This involved investigating various metal seals and calculating the forces necessary to maintain an adequate seal. The force calculations led to an analysis of several bolt patterns and lockring concepts that could be used to maintain a seal on the vessel. Another project involved some pressure vessel stress calculations and the calculation of the center of gravity for the cold source assembly. I also completed some sketches of possible cooling channel patterns for the inner vessel of the cold source. In addition, I worked on some thermal design analyses for the reflector tank and beam tubes, including heat transfer calculations and assisting in Patran and Pthermal analyses. To supplement the ANS work, I worked on other projects. I completed some stress/deflection analyses on several different beams. These analyses were done with the aid of CAASE, a beam-analysis software package. An additional project involved bending analysis on a carbon removal system. This study was done to find the deflection of a complex-shaped beam when loaded with a full waste can.

  10. REACTOR

    DOE Patents [OSTI]

    Roman, W.G.

    1961-06-27

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

  11. REACTORS

    DOE Patents [OSTI]

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  12. Two-dimensional DORT discrete ordinates X-Y geometry neutron flux calculations for the Halden Heavy Boiling Water Reactor core configurations

    SciTech Connect (OSTI)

    Slater, C.O.

    1990-07-01

    Results are reported for two-dimensional discrete ordinates, X-Y geometry calculations performed for seven Halden Heavy Boiling Water Reactor core configurations. The calculations were performed in support of an effort to reassess the neutron fluence received by the reactor vessel. Nickel foil measurement data indicated considerable underprediction of fluences by the previously used multigroup removal- diffusion method. Therefore, calculations by a more accurate method were deemed appropriate. For each core configuration, data are presented for (1) integral fluxes in the core and near the vessel wall, (2) neutron spectra at selected locations, (3) isoflux contours superimposed on the geometry models, (4) plots of the geometry models, and (5) input for the calculations. The initial calculations were performed with several mesh sizes. Comparisons of the results from these calculations indicated that the uncertainty in the calculated fluxes should be less than 10%. However, three-dimensional effects (such as axial asymmetry in the fuel loading) could contribute to much greater uncertainty in the calculated neutron fluxes. 7 refs., 22 figs., 11 tabs.

  13. A preliminary assessment of the effects of heat flux distribution and penetration on the creep rupture of a reactor vessel lower head

    SciTech Connect (OSTI)

    Chu, T.Y.; Bentz, J.; Simpson, R.; Witt, R.

    1997-02-01

    The objective of the Lower Head Failure (LHF) Experiment Program is to experimentally investigate and characterize the failure of the reactor vessel lower head due to thermal and pressure loads under severe accident conditions. The experiment is performed using 1/5-scale models of a typical PWR pressure vessel. Experiments are performed for various internal pressure and imposed heat flux distributions with and without instrumentation guide tube penetrations. The experimental program is complemented by a modest modeling program based on the application of vessel creep rupture codes developed in the TMI Vessel Investigation Project. The first three experiments under the LHF program investigated the creep rupture of simulated reactor pressure vessels without penetrations. The heat flux distributions for the three experiments are uniform (LHF-1), center-peaked (LHF-2), and side-peaked (LHF-3), respectively. For all the experiments, appreciable vessel deformation was observed to initiate at vessel wall temperatures above 900K and the vessel typically failed at approximately 1000K. The size of failure was always observed to be smaller than the heated region. For experiments with non-uniform heat flux distributions, failure typically occurs in the region of peak temperature. A brief discussion of the effect of penetration is also presented.

  14. Experimental Plan and Irradiation Target Design for FeCrAl Embrittlement Screening Tests Conducted Using the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Field, Kevin G.; Howard, Richard H.; Yamamoto, Yukinori

    2015-06-26

    The objective of the FeCrAl embrittlement screening tests being conducted through the use of Oak Ridge National Laboratories (ORNL) High Flux Isotope Reactor is to provide data on the radiation-induced changes in the mechanical properties including radiation-induced hardening and embrittlement through systematic testing and analysis. Data developed on the mechanical properties will be supported by extensive microstructural evaluations to assist in the development of structure-property relationships and provide a sound, fundamental understanding of the performance of FeCrAl alloys in intense neutron radiation fields. Data and analysis developed as part of this effort will be used to assist in the determination of FeCrAl alloys as a viable material for commercial light water reactor (LWR) applications with a primary focus as an accident tolerant cladding.

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

    SciTech Connect (OSTI)

    Bucholz, J.A.

    2000-07-01

    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.

  16. Design and Nuclear-Safety Related Simulations of Bare-Pellet Test Irradiations for the Production of Pu-238 in the High Flux Isotope Reactor using COMSOL

    SciTech Connect (OSTI)

    Freels, James D; Jain, Prashant K; Hobbs, Randy W

    2012-01-01

    The Oak Ridge National Laboratory (ORNL)is developing technology to produce plutonium-238 for the National Aeronautics and Space Administration (NASA) as a power source material for powering vehicles while in deep-space[1]. The High Flux Isotope Reactor (HFIR) of ORNL has been utilized to perform test irradiations of incapsulated neptunium oxide (NpO2) and aluminum powder bare pellets for purposes of understanding the performance of the pellets during irradiation[2]. Post irradiation examinations (PIE) are currently underway to assess the effect of temperature, thermal expansion, swelling due to gas production, fission products, and other phenomena

  17. Potential role of the Fast Flux Test Facility and the advanced test reactor in the U.S. tritium production system

    SciTech Connect (OSTI)

    Dautel, W.A.

    1996-10-01

    The Deparunent of Energy is currently engaged in a dual-track strategy to develop an accelerator and a conunercial light water reactor (CLWR) as potential sources of tritium supply. New analysis of the production capabilities of the Fast Flux Test Facility (FFTF) at the Hanford Site argues for considering its inclusion in the tritium supply,system. The use of the FFTF (alone or together with the Advanced Test Reactor [ATR] at the Idaho National Engineering Laboratory) as an integral part of,a tritium production system would help (1) ensure supply by 2005, (2) provide additional time to resolve institutional and technical issues associated with the- dual-track strategy, and (3) reduce discounted total life-cycle`costs and near-tenn annual expenditures for accelerator-based systems. The FFRF would also provide a way to get an early start.on dispositioning surplus weapons-usable plutonium as well as provide a source of medical isotopes. Challenges Associated With the Dual-Track Strategy The Departinent`s purchase of either a commercial reactor or reactor irradiation services faces challenging institutional issues associated with converting civilian reactors to defense uses. In addition, while the technical capabilities of the individual components of the accelerator have been proven, the entire system needs to be demonstrated and scaled upward to ensure that the components work toge ther 1548 as a complete production system. These challenges create uncertainty over the ability of the du2a-track strategy to provide an assured tritium supply source by 2005. Because the earliest the accelerator could come on line is 2007, it would have to operate at maximum capacity for the first few years to regenerate the reserves lost through radioactive decay aftei 2005.

  18. Decomposition of 1,1-Dichloroethane and 1,1-Dichloroethene in an electron beam generated plasma reactor

    SciTech Connect (OSTI)

    Vitale, S.A.; Hadidi, K.; Cohn, D.R.; Bromberg, L.

    1997-03-01

    An electron beam generated plasma reactor is used to decompose low concentrations (100{endash}3000 ppm) of 1,1-dichloroethane and 1,1-dichloroethene in atmospheric pressure air streams. The energy requirements for 90{percent} and 99{percent} decomposition of each compound are reported as a function of inlet concentration. Dichloroethene decomposition is enhanced by a chlorine radical propagated chain reaction. The chain length of the dichloroethene reaction is estimated to increase with dichloroethene concentration from 10 at 100 ppm initial dichloroethene concentration to 30 at 3000 ppm. Both the dichloroethane and dichloroethene reactions seem to be inhibited by electron scavenging decomposition products. A simple analytic expression is proposed for fitting decomposition data where inhibition effects are important and simple first order kinetics are not observed. {copyright} {ital 1997 American Institute of Physics.}

  19. Neutronics Simulations of 237Np Targets to Support Safety-Basis and 238Pu Production Assessment Efforts at the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Chandler, David; Ellis, Ronald James

    2015-01-01

    Fueled by two highly enriched uranium-bearing fuel elements surrounded by a large concentric ring of beryllium reflector, the High Flux Isotope Reactor (HFIR) provides one of the highest neutron fluxes in the world and is used to produce unique isotopes like plutonium-238. The National Aeronautics and Space Administration use radioisotope thermoelectric generators powered by 238Pu for deep-space missions. As part of the US Department of Energy s task to reestablish the domestic production of 238Pu, a technology demonstration sub-project has been initiated to establish a new 238Pu supply chain. HFIR safety-basis neutronics calculations are being performed to ensure the target irradiations have no adverse impacts on reactor performance and to calculate data required as input to follow-on thermal-structural, thermal-hydraulic and radionuclide/dose analyses. Plutonium-238 production assessments are being performed to estimate the amount of 238Pu that can be produced in HFIR s permanent beryllium reflector. It is estimated that a total of 0.96 1.12 kg 238Pu (~1.28 1.49 kg PuO2 at 85% 238Pu/Pu purity) could be produced per year in HFIR s permanent beryllium reflector irradiation facilities if they are all utilized.

  20. Reactor

    DOE Patents [OSTI]

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

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

    SciTech Connect (OSTI)

    Not Available

    1994-03-01

    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.

  2. 235U Determination using In-Beam Delayed Neutron Counting Technique at the NRU Reactor

    SciTech Connect (OSTI)

    Andrews, M. T.; Bentoumi, G.; Corcoran, E. C.; Dimayuga, I.; Kelly, D. G.; Li, L.; Sur, B.; Rogge, R. B.

    2015-11-17

    This paper describes a collaborative effort that saw the Royal Military College of Canada (RMC)’s delayed neutron and gamma counting apparatus transported to Canadian Nuclear Laboratories (CNL) for use in the neutron beamline at the National Research Universal (NRU) reactor. Samples containing mg quantities of fissile material were re-interrogated, and their delayed neutron emissions measured. This collaboration offers significant advantages to previous delayed neutron research at both CNL and RMC. This paper details the determination of 235U content in enriched uranium via the assay of in-beam delayed neutron magnitudes and temporal behavior. 235U mass was determined with an average absolute error of ± 2.7 %. This error is lower than that obtained at RMCC for the assay of 235U content in aqueous solutions (3.6 %) using delayed neutron counting. Delayed neutron counting has been demonstrated to be a rapid, accurate, and precise method for special nuclear material detection and identification.

  3. The role of the neutral beam fueling profile in the performance of the Tokamak Fusion Test Reactor and other tokamak plasmas

    SciTech Connect (OSTI)

    Park, H.K.; Batha, S.; Sabbagh, S.A. |

    1997-02-01

    Scalings for the stored energy and neutron yield, determined from experimental data are applied to both deuterium-only and deuterium-tritium plasmas in different neutral beam heated operational domains in Tokamak Fusion Test Reactor. The domain of the data considered includes the Supershot, High poloidal beta, Low-mode, and limiter High-mode operational regimes, as well as discharges with a reversed magnetic shear configuration. The new important parameter in the present scaling is the peakedness of the heating beam fueling profile shape. Ion energy confinement and neutron production are relatively insensitive to other plasma parameters compared to the beam fueling peakedness parameter and the heating beam power when considering plasmas that are stable to magnetohydrodynamic modes. However, the stored energy of the electrons is independent of the beam fueling peakedness. The implication of the scalings based on this parameter is related to theoretical transport models such as radial electric field shear and Ion Temperature Gradient marginality models. Similar physics interpretation is provided for beam heated discharges on other major tokamaks.

  4. Design of a High Resolution and High Flux Beam line for VUV Angle-Resolved Photoemission at UVSOR-II

    SciTech Connect (OSTI)

    Kimura, Shin-ichi; Ito, Takahiro; Nakamura, Eiken; Hosaka, Masahito; Katoh, Masahiro

    2007-01-19

    A high-energy-resolution angle-resolved photoemission beamline in the vacuum-ultraviolet (VUV) region has been designed for a 750 MeV synchrotron light source UVSOR-II. The beamline equips an APPLE-II-type undulator with the horizontally/vertically linear and right/left circular polarizations, a modified Wadsworth-type monochromator and a high-resolution photoelectron analyzer. The monochromator covers the photon energy range of 6 - 40 eV. The energy resolution (hv/{delta}hv) and the photon flux on samples are expected to be 2 x 104 and 1012 photons/sec at 10 eV, 4 x 104 and 5 x 1011 photons/sec at 20 eV, and 6 x 104 and 1011 photons/sec at 40 eV, respectively. The beamline provides the high-resolution angle-resolved photoemission spectroscopy less than 1 meV in the whole VUV energy range.

  5. Study of the Potential Impact of Gamma-Induced Radiolytic Gases on Loading of Cesium Onto Crystalline Silicotitanate Sorbent at ORNL's High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Mattus, A.J.

    2001-02-12

    The use of an engineered form of crystalline silicotitanate as a potential sorbent for the removal and concentration of cesium from the high-level waste at the Savannah River Site was investigated. Results conclusively showed this sorbent to be unaffected by gamma-induced radiolytic gas formation during column loading. Closely controlled column-loading experiments were performed at the Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR) in a gamma field with a conservative dose rate expected to exceed that in a full-scale column by a factor of nearly 16. Operation of column loading under expected nominal full-scale field conditions in the HFIR pool showed that radiolytic gases were formed at a previously calculated generation rate of 0.4 mL per liter of feed solution. When the resulting cesium-loading curve in the gamma field was compared with that of a control experiment in the absence of a gamma field, no discernable difference in the curves (within analytical error) was detected. Both curves were in good agreement with the VERSE computer-generated curve. Results conclusively indicate that the production of radiolytic gases within a full-scale column is not expected to result in reduced capacity or associated gas generation problems during operation at the Savannah River Site.

  6. LCLS Spectral Flux Viewer

    Energy Science and Technology Software Center (OSTI)

    2005-10-25

    This application (FluxViewer) is a tool for displaying spectral flux data for the Linac Coherent Light Source (LCLS). This tool allows the user to view sliced spatial and energy distributions of the photons selected for specific energies and positions transverse to the beam axis.

  7. Neutral Beam Excitation

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

    ... the presence of a small fast ion population are one area ... This research is highly relevant to fusion reactors as beam ... 100 A beam of < 200 keV sodium or potassium ions which ...

  8. 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)

    Leplat, N.; Rossi, M. J.

    2013-11-15

    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 (300630 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.

  9. NEUTRON FLUX INTENSITY DETECTION

    DOE Patents [OSTI]

    Russell, J.T.

    1964-04-21

    A method of measuring the instantaneous intensity of neutron flux in the core of a nuclear reactor is described. A target gas capable of being transmuted by neutron bombardment to a product having a resonance absorption line nt a particular microwave frequency is passed through the core of the reactor. Frequency-modulated microwave energy is passed through the target gas and the attenuation of the energy due to the formation of the transmuted product is measured. (AEC)

  10. Neutronic reactor

    DOE Patents [OSTI]

    Wende, Charles W. J.; Babcock, Dale F.; Menegus, Robert L.

    1983-01-01

    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.

  11. Note: Ion-induced secondary electron emission from oxidized metal surfaces measured in a particle beam reactor

    SciTech Connect (OSTI)

    Marcak, Adrian; Corbella, Carles Keudell, Achim von; Arcos, Teresa de los

    2015-10-15

    The secondary electron emission of metals induced by slow ions is characterized in a beam chamber by means of two coaxial semi-cylindrical electrodes with different apertures. The voltages of the outer electrode (screening), inner electrode (collector), and sample holder (target) were set independently in order to measure the effective yield of potential and kinetic electron emissions during ion bombardment. Aluminum samples were exposed to quantified beams of argon ions up to 2000 eV and to oxygen atoms and molecules in order to mimic the plasma-surface interactions on metallic targets during reactive sputtering. The variation of electron emission yield was correlated to the ion energy and to the oxidation state of Al surfaces. This system provides reliable measurements of the electron yields in real time and is of great utility to explore the fundamental surface processes during target poisoning occurring in reactive magnetron sputtering applications.

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

    SciTech Connect (OSTI)

    Werley, K.A.

    1989-01-01

    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.

  13. Method of fission product beta spectra measurements for predicting reactor anti-neutrino emission

    SciTech Connect (OSTI)

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

    2015-03-01

    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 antineutrinos. These reactor neutrinos have served neutrino physics research from the initial discovery of the neutrino to today's 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 reconsiderations 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.

  14. Radiation dosimetry at the BNL Medical Research Reactor

    SciTech Connect (OSTI)

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

    1998-11-01

    The Medical Research Reactor, BMRR, at the Brookhaven National Laboratory, BNL, is a three megawatt, 3 MW, heterogeneous, tank-type, light water cooled and moderated, graphite reflected reactor, which was designed for biomedical studies, and became operational in 1959. It provides thermal and epithermal neutron beams suitable for research studies such as radiation therapy of various types of tumors. At the present time, the major program at BMRR is Boron Neutron Capture Therapy, BNCT. Modifications have been made to the BMRR to significantly increase the available epithermal neutron flux density to a patient in clinical trials of BNCT. The data indicate that the flux density and dose rate are concentrated in the center of the beam, the patient absorbs neutrons rather than gamma radiation and as noted previously even with the increasing flux values, gamma-ray dose received by the attending personnel has remained minimal. Flux densities in the center of the thermal port and epithermal port beams have been characterized with an agreement between the measurements and the calculations.

  15. Beta ray flux measuring device

    DOE Patents [OSTI]

    Impink, Jr., Albert J.; Goldstein, Norman P.

    1990-01-01

    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.

  16. REACTOR MONITORING

    DOE Patents [OSTI]

    Bugbee, S.J.; Hanson, V.F.; Babcock, D.F.

    1959-02-01

    A neutron density inonitoring means for reactors is described. According to this invention a tunnel is provided beneath and spaced from the active portion of the reactor and extends beyond the opposite faces of the activc portion. Neutron beam holes are provided between the active portion and the tunnel and open into the tunnel near the middle thereof. A carriage operates back and forth in the tunnel and is adapted to convey a neutron detector, such as an ion chamber, and position it beneath one of the neutron beam holes. This arrangement affords convenient access of neutron density measuring instruments to a location wherein direct measurement of neutron density within the piles can be made and at the same time affords ample protection to operating personnel.

  17. B Reactor - Hanford Site

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

    Reactor About Us About Hanford Cleanup Hanford History Hanford Site Wide Programs Contact Us 100 Area 118-K-1 Burial Ground 200 Area 222-S Laboratory 242-A Evaporator 300 Area 324 Building 325 Building 400 Area/Fast Flux Test Facility 618-10 and 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and Interim Storage Area Canyon Facilities Cold Test Facility D and DR Reactors Effluent Treatment Facility Environmental Restoration Disposal Facility F Reactor H

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

    SciTech Connect (OSTI)

    Evan Harpeneau

    2011-06-24

    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.

  19. THERMAL NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Spinrad, B.I.

    1960-01-12

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

  20. NUCLEAR REACTOR FUEL ELEMENT

    DOE Patents [OSTI]

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

    1963-06-11

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

  1. Fission reactors and materials

    SciTech Connect (OSTI)

    Frost, B.R.T.

    1981-12-01

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

  2. D and DR Reactors - Hanford Site

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

    300 Area 324 Building 325 Building 400 AreaFast Flux Test Facility 618-10 and 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and Interim ...

  3. Beam Status

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

    Beam Status Beam Status Print Loading... You can also view the Operations Group's Beam History archives.

  4. Beam Characterization at the Neutron Radiography Facility

    SciTech Connect (OSTI)

    Sarah Morgan; Jeffrey King

    2013-01-01

    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.

  5. Nuclear reactor shutdown system

    DOE Patents [OSTI]

    Bhate, Suresh K.; Cooper, Martin H.; Riffe, Delmar R.; Kinney, Calvin L.

    1981-01-01

    An inherent shutdown system for a nuclear reactor having neutron absorbing rods affixed to an armature which is held in an upper position by a magnetic flux flowing through a Curie temperature material. The Curie temperature material is fixedly positioned about the exterior of an inner duct in an annular region through which reactor coolant flows. Elongated fuel rods extending from within the core upwardly toward the Curie temperature material are preferably disposed within the annular region. Upon abnormal conditions which result in high neutron flux and coolant temperature, the Curie material loses its magnetic permeability, breaking the magnetic flux path and allowing the armature and absorber rods to drop into the core, thus shutting down the fissioning reaction. The armature and absorber rods are retrieved by lowering the housing for the electromagnet forming coils which create a magnetic flux path which includes the inner duct wall. The coil housing then is raised, resetting the armature.

  6. Fusion reactor pumped laser

    DOE Patents [OSTI]

    Jassby, Daniel L.

    1988-01-01

    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.

  7. Reactor production of Thoruim-229

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Boll, Rose Ann; Murphy, Karen E.; Denton, David L.; Tamara J. Haverlock; Garland, Marc A.; Mirzadeh, Saed; Hogle, Susan; Owens, Allison

    2016-05-03

    Limited availability of 229Th for clinical applications of 213Bi necessitates investigation of alternative production routes. In reactor production, 229Th is produced from neutron transmutation of 226Ra, 228Ra, 227Ac and 228Th. Here, we evaluate irradiations of 226Ra, 228Ra, and 227Ac targets at the ORNL High Flux Isotope Reactor.

  8. Toroidal midplane neutral beam armor and plasma limiter

    DOE Patents [OSTI]

    Kugel, Henry W.; Hand Jr, Samuel W.; Ksayian, Haig

    1986-02-04

    For use in a tokamak fusion reactor having a midplane magnetic coil on the inner wall of an evacuated toriodal chamber within which a neutral beam heated, fusing plasma is magnetically confined, a neutral beam armor shield and plasma limiter is provided on the inner wall of the toroidal chamber to shield the midplane coil from neutral beam shine-thru and plasma deposition. The armor shield/plasma limiter forms a semicircular enclosure around the midplane coil with the outer surface of the armor shield/plasma limiter shaped to match, as closely as practical, the inner limiting magnetic flux surface of the toroidally confined, indented, bean-shaped plasma. The armor shield/plasma limiter includes a plurality of semicircular graphite plates each having a pair of coupled upper and lower sections with each plate positioned in intimate contact with an adjacent plate on each side thereof so as to form a closed, planar structure around the entire outer periphery of the circular midplane coil. The upper and lower plate sections are adapted for coupling to heat sensing thermocouples and to a circulating water conduit system for cooling the armor shield/plasma limiter.The inner center portion of each graphite plate is adapted to receive and enclose a section of a circular diagnostic magnetic flux loop so as to minimize the power from the plasma confinement chamber incident upon the flux loop.

  9. Toroidal midplane neutral beam armor and plasma limiter

    DOE Patents [OSTI]

    Kugel, Henry W.; Hand, Jr, Samuel W.; Ksayian, Haig

    1986-01-01

    For use in a tokamak fusion reactor having a midplane magnetic coil on the inner wall of an evacuated toriodal chamber within which a neutral beam heated, fusing plasma is magnetically confined, a neutral beam armor shield and plasma limiter is provided on the inner wall of the toroidal chamber to shield the midplane coil from neutral beam shine-thru and plasma deposition. The armor shield/plasma limiter forms a semicircular enclosure around the midplane coil with the outer surface of the armor shield/plasma limiter shaped to match, as closely as practical, the inner limiting magnetic flux surface of the toroidally confined, indented, bean-shaped plasma. The armor shield/plasma limiter includes a plurality of semicircular graphite plates each having a pair of coupled upper and lower sections with each plate positioned in intimate contact with an adjacent plate on each side thereof so as to form a closed, planar structure around the entire outer periphery of the circular midplane coil. The upper and lower plate sections are adapted for coupling to heat sensing thermocouples and to a circulating water conduit system for cooling the armor shield/plasma limiter.The inner center portion of each graphite plate is adapted to receive and enclose a section of a circular diagnostic magnetic flux loop so as to minimize the power from the plasma confinement chamber incident upon the flux loop.

  10. Pulsed ion beam source

    DOE Patents [OSTI]

    Greenly, John B.

    1996-01-01

    An improved magnetically-confined anode plasma pulsed ion beam source. Beam rotation effects and power efficiency are improved by a magnetic design which places the separatrix between the fast field flux structure and the slow field structure near the anode of the ion beam source, by a gas port design which localizes the gas delivery into the gap between the fast coil and the anode, by a pre-ionizer ringing circuit connected to the fast coil, and by a bias field means which optimally adjusts the plasma formation position in the ion beam source.

  11. Internal attachment of laser beam welded stainless steel sheathed thermocouples into stainless steel upper end caps in nuclear fuel rods for the LOFT Reactor

    SciTech Connect (OSTI)

    Welty, R.K.; Reid, R.D.

    1980-01-01

    The Exxon Nuclear Company, Inc., acting as a subcontractor to EG and G Idaho Inc., Idaho National Engineering Laboratory, Idaho Falls, Idaho, conducted a laser beam welding study to attach internal stainless steel thermocouples into stainless steel upper end caps in nuclear fuel rods. The objective of this study was to determine the feasibility of laser welding a single 0.063 inch diameter stainless steel (304) sheathed thermocouple into a stainless steel (316) upper end cap for nuclear fuel rods. A laser beam was selected because of the extremely high energy input in unit volume that can be achieved allowing local fusion of a small area irrespective of the difference in material thickness to be joined. A special weld fixture was designed and fabricated to hold the end cap and the thermocouple with angular and rotational adjustment under the laser beam. A commercial pulsed laser and energy control system was used to make the welds.

  12. Beam Status

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

    Beam Status Print Loading... You can also view the Operations Group's Beam History archives

  13. Fueling of tandem mirror reactors

    SciTech Connect (OSTI)

    Gorker, G.E.; Logan, B.G.

    1985-01-01

    This paper summarizes the fueling requirements for experimental and demonstration tandem mirror reactors (TMRs), reviews the status of conventional pellet injectors, and identifies some candidate accelerators that may be needed for fueling tandem mirror reactors. Characteristics and limitations of three types of accelerators are described; neutral beam injectors, electromagnetic rail guns, and laser beam drivers. Based on these characteristics and limitations, a computer module was developed for the Tandem Mirror Reactor Systems Code (TMRSC) to select the pellet injector/accelerator combination which most nearly satisfies the fueling requirements for a given machine design.

  14. Boiling-Water Reactor internals aging degradation study. Phase 1

    SciTech Connect (OSTI)

    Luk, K.H.

    1993-09-01

    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.

  15. K-East and K-West Reactors - Hanford Site

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

    300 Area 324 Building 325 Building 400 AreaFast Flux Test Facility 618-10 and 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and Interim ...

  16. Toroidal midplane neutral beam armor and plasma limiter

    DOE Patents [OSTI]

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

    1985-05-31

    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.

  17. Beam current sensor

    DOE Patents [OSTI]

    Kuchnir, M.; Mills, F.E.

    1984-09-28

    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.

  18. Beam current sensor

    DOE Patents [OSTI]

    Kuchnir, Moyses; Mills, Frederick E.

    1987-01-01

    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.

  19. Plug Flow Reactor Simulator

    Energy Science and Technology Software Center (OSTI)

    1996-07-30

    PLUG is a computer program that solves the coupled steady state continuity, momentum, energy, and species balance equations for a plug flow reactor. Both homogeneous (gas-phase) and heterogenous (surface) reactions can be accommodated. The reactor may be either isothermal or adiabatic or may have a specified axial temperature or heat flux profile; alternatively, an ambient temperature and an overall heat-transfer coefficient can be specified. The crosssectional area and surface area may vary with axial position,more » and viscous drag is included. Ideal gas behavior and surface site conservation are assumed.« less

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

    SciTech Connect (OSTI)

    Park, R. J.; Lee, J. R.; Kim, S. B.; Jin, Y.; Kim, H. Y.

    2012-07-01

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

  1. Beam-beam simulations for separated beams

    SciTech Connect (OSTI)

    Furman, Miguel A.

    2000-04-10

    We present beam-beam simulation results from a strong-strong gaussian code for separated beams for the LHC and RHIC. The frequency spectrum produced by the beam-beam collisions is readily obtained and offers a good opportunity for experimental comparisons. Although our results for the emittance blowup are preliminary, we conclude that, for nominal parameter values, there is no significant difference between separated beams and center-on-center collisions.

  2. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fermi, E.; Zinn, W.H.; Anderson, H.L.

    1958-09-16

    Means are presenied for increasing the reproduction ratio of a gaphite- moderated neutronic reactor by diminishing the neutron loss due to absorption or capture by gaseous impurities within the reactor. This means comprised of a fluid-tight casing or envelope completely enclosing the reactor and provided with a valve through which the casing, and thereby the reactor, may be evacuated of atmospheric air.

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

    SciTech Connect (OSTI)

    Wilson, E.H.; Horelik, N.E.; Dunn, F.E.; Newton, T.H., Jr.; Hu, L.; Stevens, J.G.

    2012-04-04

    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.

  4. 400 Area/Fast Flux Test Facility - Hanford Site

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

    325 Building 400 AreaFast Flux Test Facility 618-10 ... Test Facility D and DR Reactors Effluent Treatment ... (thermal) liquid-metal (sodium)-cooled nuclear research ...

  5. Co: clqrt. Beam

    Office of Legacy Management (LM)

    Co: clqrt. Beam*/:

  6. APPARATUS FOR CONTROLLING NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Dietrich, J.R.; Harrer, J.M.

    1958-09-16

    A device is described for rapidly cortrolling the reactivity of an active portion of a reactor. The inveniion consists of coaxially disposed members each having circumferenital sections of material having dlfferent neutron absorbing characteristics and means fur moving the members rotatably and translatably relative to each other within the active portion to vary the neutron flux therein. The angular and translational movements of any member change the neutron flux shadowing effect of that member upon the other member.

  7. Fusion reactor pumped laser

    DOE Patents [OSTI]

    Jassby, D.L.

    1987-09-04

    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.

  8. NUCLEAR REACTOR CONTROL SYSTEM

    DOE Patents [OSTI]

    Howard, D.F.; Motta, E.E.

    1961-06-27

    A method for controlling the excess reactivity in a nuclear reactor throughout the core life while maintaining the neutron flux distribution at the desired level is described. The control unit embodies a container having two electrodes of different surface area immersed in an electrolytic solution of a good neutron sbsorbing metal ion such as boron, gadolinium, or cadmium. Initially, the neutron absorber is plated on the larger electrode to control the greater neutron flux of a freshly refueled core. As the fuel burns up, the excess reactivity decreases and the neutron absorber is then plated onto the smaller electrode so that the number of neutrons absorbed also decreases. The excess reactivity in the core may thus be maintained without the introduction of serious perturbations in the neutron flux distributibn.

  9. Fast flux locked loop

    DOE Patents [OSTI]

    Ganther, Jr., Kenneth R.; Snapp, Lowell D.

    2002-09-10

    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.

  10. Heat dissipating nuclear reactor

    DOE Patents [OSTI]

    Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

    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.

  11. Heat dissipating nuclear reactor

    DOE Patents [OSTI]

    Hunsbedt, Anstein; Lazarus, Jonathan D.

    1987-01-01

    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.

  12. Knowledge Preservation at the Fast Flux Test Facility

    SciTech Connect (OSTI)

    Wootan, David W.; Omberg, Ronald P.; Makenas, Bruce J.; Nielsen, Deborah L.; Nelson, Joseph V.; Polzin, David L.

    2012-01-30

    The Fast Flux Test Facility (FFTF) is the most recent Liquid Metal Reactor (LMR) to operate in the United States, from 1982 to 1992. The technologies employed in designing and constructing this reactor, along with information obtained from tests conducted during its operation, are currently being secured and archived by the Department of Energy's Office of Nuclear Energy. This report is one in a series documenting the overall project efforts to retrieve and preserve critical information related to advanced reactors

  13. Knowledge Preservation at the Fast Flux Test Facility

    SciTech Connect (OSTI)

    Wootan, David W.; Omberg, Ronald P.; Makenas, Bruce J.; Nielsen, Deborah L.; Nelson, Joseph V.; Polzin, David L.

    2011-11-30

    The Fast Flux Test Facility (FFTF) is the most recent Liquid Metal Reactor (LMR) to operate in the United States, from 1982 to 1992. The technologies employed in designing and constructing this reactor, along with information obtained from tests conducted during its operation, are currently being secured and archived by the Department of Energy's Office of Nuclear Energy. This report provides a status update documenting the overall project efforts to retrieve and preserve critical information related to advanced reactors.

  14. MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR

    DOE Patents [OSTI]

    Balent, R.

    1963-03-12

    This patent describes a method of obtaining a flatter flux and more uniform power generation across the core of a nuclear reactor. The method comprises using moderator elements having differing moderating strength. The elements have an increasing amount of the better moderating material as a function of radial and/or axial distance from the reactor core center. (AEC)

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

    SciTech Connect (OSTI)

    Halfon, S.; Feinberg, G.; Paul, M.; Arenshtam, A.; Berkovits, D.; Kijel, D.; Nagler, A.; Eliyahu, I.; Silverman, I.

    2013-04-19

    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.

  16. Reactor-Produced Medical Radionuclides

    SciTech Connect (OSTI)

    Mirzadeh, Saed; Mausner, Leonard; Garland, Marc A

    2011-01-01

    The therapeutic use of radionuclides in nuclear medicine, oncology and cardiology is the most rapidly growing use of medical radionuclides. Since most therapeutic radionuclides are neutron rich and decay by beta emission, they are reactor-produced. This chapter deals mainly with production approaches with neutrons. Neutron interactions with matter, neutron transmission and activation rates, and neutron spectra of nuclear reactors are discussed in some detail. Further, a short discussion of the neutron-energy dependence of cross sections, reaction rates in thermal reactors, cross section measurements and flux monitoring, and general equations governing the reactor production of radionuclides are presented. Finally, the chapter is concluded by providing a number of examples encompassing the various possible reaction routes for production of a number of medical radionuclides in a reactor.

  17. CONTROL SYSTEM FOR NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Crever, F.E.

    1962-05-01

    BS>A slow-acting shim rod for control of major variations in reactor neutron flux and a fast-acting control rod to correct minor flux variations are employed to provide a sensitive, accurate control system. The fast-acting rod is responsive to an error signal which is produced by changes in the neutron flux from a predetermined optimum level. When the fast rod is thus actuated in a given direction, means is provided to actuate the slow-moving rod in that direction to return the fast rod to a position near the midpoint of its control range. (AEC)

  18. BOILING REACTORS

    DOE Patents [OSTI]

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  19. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Daniels, F.

    1959-10-27

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

  20. Solid tags for identifying failed reactor components

    DOE Patents [OSTI]

    Bunch, Wilbur L.; Schenter, Robert E.

    1987-01-01

    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.

  1. Nuclear reactor shield including magnesium oxide

    DOE Patents [OSTI]

    Rouse, Carl A.; Simnad, Massoud T.

    1981-01-01

    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.

  2. Energy deposition in STARFIRE reactor components

    SciTech Connect (OSTI)

    Gohar, Y.; Brooks, J.N.

    1985-04-01

    The energy deposition in the STARFIRE commercial tokamak reactor was calculated based on detailed models for the different reactor components. The heat deposition and the 14 MeV neutron flux poloidal distributions in the first wall were obtained. The poloidal surface heat load distribution in the first wall was calculated from the plasma radiation. The Monte Carlo method was used for the calculation to allow an accurate modeling for the reactor geometry.

  3. CONVECTION REACTOR

    DOE Patents [OSTI]

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

    1960-03-22

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

  4. Pulse flux measuring device

    DOE Patents [OSTI]

    Riggan, William C.

    1985-01-01

    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.

  5. NuMI Low Energy Flux Prediction Release

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

    NuMI Low Energy Flux Prediction Release Neutrino Flux Predictions for the NuMI Beam hep-ex/1607.00704 Data Ancillary data files for this result are available on arXiv at http://arxiv.org/src/1607.00704/anc.< /li> Among the available data files are: pdf file describing format of all the available files root file of all the available fluxes python code to read and process MINERvA's flux predictions Text Files of the flux, uncertainties, and covariance matrix, with units of neutrinos/m^2/POT,

  6. Toroidal reactor

    DOE Patents [OSTI]

    Dawson, John M.; Furth, Harold P.; Tenney, Fred H.

    1988-12-06

    Method for producing fusion power wherein a neutral beam is injected into a toroidal bulk plasma to produce fusion reactions during the time permitted by the slowing down of the particles from the injected beam in the bulk plasma.

  7. NEUTRONIC REACTOR

    DOE Patents [OSTI]

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

    1961-11-21

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

  8. Photon beam position monitor

    DOE Patents [OSTI]

    Kuzay, T.M.; Shu, D.

    1995-02-07

    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.

  9. Photon beam position monitor

    DOE Patents [OSTI]

    Kuzay, Tuncer M.; Shu, Deming

    1995-01-01

    A photon beam position monitor 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.

  10. REACTOR COOLING

    DOE Patents [OSTI]

    Quackenbush, C.F.

    1959-09-29

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

  11. ENGINEERING TEST REACTOR

    DOE Patents [OSTI]

    De Boisblanc, D.R.; Thomas, M.E.; Jones, R.M.; Hanson, G.H.

    1958-10-21

    Heterogeneous reactors of the type which is both cooled and moderated by the same fluid, preferably water, and employs highly enriched fuel are reported. In this design, an inner pressure vessel is located within a main outer pressure vessel. The reactor core and its surrounding reflector are disposed in the inner pressure vessel which in turn is surrounded by a thermal shield, Coolant fluid enters the main pressure vessel, fiows downward into the inner vessel where it passes through the core containing tbe fissionable fuel assemblies and control rods, through the reflector, thence out through the bottom of the inner vessel and up past the thermal shield to the discharge port in the main vessel. The fuel assemblles are arranged in the core in the form of a cross having an opening extending therethrough to serve as a high fast flux test facility.

  12. WATER BOILER REACTOR

    DOE Patents [OSTI]

    King, L.D.P.

    1960-11-22

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

  13. Neutrino-4 experiment on the search for a sterile neutrino at the SM-3 reactor

    SciTech Connect (OSTI)

    Serebrov, A. P. Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Chernyi, A. V.; Zherebtsov, O. M.; Martemyanov, V. P.; Tsinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K.; and others

    2015-10-15

    In view of the possibility of the existence of a sterile neutrino, test measurements of the dependence of the reactor antineutrino flux on the distance from the reactor core has been performed on SM-2 reactor with the Neutrino-2 detector model in the range of 6–11 m. Prospects of the search for reactor antineutrinos at short distances have been discussed.

  14. Fast Reactor Technology Preservation

    SciTech Connect (OSTI)

    Wootan, David W.; Omberg, Ronald P.

    2008-01-11

    There is renewed worldwide interest in developing and implementing a new generation of advanced fast reactors. International cooperative efforts are underway such as the Global Nuclear Energy Partnership (GNEP). Advanced computer modeling and simulation efforts are a key part of these programs. A recognized and validated set of Benchmark Cases are an essential component of such modeling efforts. Testing documentation developed during the operation of the Fast Flux Test Facility (FFTF) provide the information necessary to develop a very useful set of Benchmark Cases.

  15. Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

    SciTech Connect (OSTI)

    McCulloch, R.W.; MacPherson, R.E.

    1983-03-01

    The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370/sup 0/C.

  16. Beam Transport

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

    Beam Transport A simplified drawing of the beam transport system from the linac to Target-1 (Lujan Center), Target-2 (Blue Room) and Target-4 is shown below. In usual operation ...

  17. Nuclear reactor building

    DOE Patents [OSTI]

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

    1994-04-05

    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.

  18. Nuclear reactor building

    DOE Patents [OSTI]

    Gou, Perng-Fei; Townsend, Harold E.; Barbanti, Giancarlo

    1994-01-01

    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.

  19. Five years operating experience at the Fast Flux Test Facility

    SciTech Connect (OSTI)

    Baumhardt, R. J.; Bechtold, R. A.

    1987-04-01

    The Fast Flux Test Facility (FFTF) is a 400 Mw(t), loop-type, sodium-cooled, fast neutron reactor. It is operated by the Westinghouse Hanford Company for the United States Department of Energy at Richland, Washington. The FFTF is a multipurpose test reactor used to irradiate fuels and materials for programs such as Liquid Metal Reactor (LMR) research, fusion research, space power systems, isotope production and international research. FFTF is also used for testing concepts to be used in Advanced Reactors which will be designed to maximize passive safety features and not require complex shutdown systems to assure safe shutdown and heat removal. The FFTF also provides experience in the operation and maintenance of a reactor having prototypic components and systems typical of large LMR (LMFBR) power plants. The 5 year operational performance of the FFTF reactor is discussed in this report. 6 refs., 10 figs., 2 tabs.

  20. NUCLEAR REACTOR

    DOE Patents [OSTI]

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

    1958-12-01

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

  1. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Hurwitz, H. Jr.; Brooks, H.; Mannal, C.; Payne, J.H.; Luebke, E.A.

    1959-03-24

    A reactor of the heterogeneous, liquid cooled type is described. This reactor is comprised of a central region of a plurality of vertically disposed elongated tubes surrounded by a region of moderator material. The central region is comprised of a central core surrounded by a reflector region which is surrounded by a fast neutron absorber region, which in turn is surrounded by a slow neutron absorber region. Liquid sodium is used as the primary coolant and circulates through the core which contains the fuel elements. Control of the reactor is accomplished by varying the ability of the reflector region to reflect neutrons back into the core of the reactor. For this purpose the reflector is comprised of moderator and control elements having varying effects on reactivity, the control elements being arranged and actuated by groups to give regulation, shim, and safety control.

  2. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Wigner, E.P.

    1960-11-22

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

  3. REACTOR SHIELD

    DOE Patents [OSTI]

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

    1959-02-17

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

  4. Reactor apparatus

    DOE Patents [OSTI]

    Echtler, J. Paul

    1981-01-01

    A reactor apparatus for hydrocracking a polynuclear aromatic hydrocarbonaceous feedstock to produce lighter hydrocarbon fuels by contacting the hydrocarbonaceous feedstock with hydrogen in the presence of a molten metal halide catalyst.

  5. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Anderson, H.L.

    1960-09-20

    A nuclear reactor is described comprising fissionable material dispersed in graphite blocks, helium filling the voids of the blocks and the spaces therebetween, and means other than the helium in thermal conductive contact with the graphite for removing heat.

  6. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fermi, E.

    1960-04-01

    A nuclear reactor is described consisting of blocks of graphite arranged in layers, natural uranium bodies disposed in holes in alternate layers of graphite blocks, and coolant tubes disposed in the layers of graphite blocks which do not contain uranium.

  7. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  8. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Vernon, H.C.

    1959-01-13

    A neutronic reactor of the heterogeneous, fluid cooled tvpe is described. The reactor is comprised of a pressure vessel containing the moderator and a plurality of vertically disposed channels extending in spaced relationship through the moderator. Fissionable fuel material is placed within the channels in spaced relationship thereto to permit circulation of the coolant fluid. Separate means are provided for cooling the moderator and for circulating a fluid coolant thru the channel elements to cool the fuel material.

  9. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Anderson, C.R.

    1962-07-24

    A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)

  10. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Breden, C.R.; Dietrich, J.R.

    1961-06-20

    A water-soluble non-volatile poison may be introduced into a reactor to nullify excess reactivity. The poison is removed by passing a side stream of the water containing the soluble poison to an evaporation chamber. The vapor phase is returned to the reactor to decrease the concentration of soluble poison and the liquid phase is returned to increase the concentration of soluble poison.

  11. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Grebe, J.J.

    1959-07-14

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

  12. ARM - Measurement - Methane flux

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

    flux ARM Data Discovery Browse Data Comments? We would love to hear from you! Send us a note below or call us at 1-888-ARM-DATA. Send Measurement : Methane flux Vertical flux of methane near the surface due to turbulent transport. Categories Surface Properties, Atmospheric Carbon Instruments The above measurement is considered scientifically relevant for the following instruments. Refer to the datastream (netcdf) file headers of each instrument for a list of all available measurements, including

  13. Startup Testing of the Fast Flux Test Facility

    SciTech Connect (OSTI)

    Wootan, David W.; Butner, R. Scott; Omberg, Ronald P.; Makenas, Bruce J.; Nielsen, Deborah L.; Polzin, David L.

    2010-06-30

    This paper is one in a series documenting the current effort to retrieve, secure, and preserve critical information related to advanced reactors. . Information from this testing is being retrieved under the Fuel Cycle Research and Development (FCRD) program conducted by the Office of Nuclear Energy (NE) of the DOE. The Fast Flux Test Facility (FFTF) is the most recent Liquid Metal Reactor (LMR) to be designed, constructed, and operated by the U.S. Department of Energy (DOE).

  14. Recovery of Information from the Fast Flux Test Facility for the Advanced Fuel Cycle Initiative

    SciTech Connect (OSTI)

    Nielsen, Deborah L.; Makenas, Bruce J.; Wootan, David W.; Butner, R. Scott; Omberg, Ronald P.

    2009-09-30

    The Fast Flux Test Facility is the most recent Liquid Metal Reactor to operate in the United States. Information from the design, construction, and operation of this reactor was at risk as the facilities associated with the reactor are being shut down. The Advanced Fuel Cycle Initiative is a program managed by the Office of Nuclear Energy of the U.S. Department of Energy with a mission to develop new fuel cycle technologies to support both current and advanced reactors. Securing and preserving the knowledge gained from operation and testing in the Fast Flux Test Facility is an important part of the Knowledge Preservation activity in this program.

  15. Thermionic switched self-actuating reactor shutdown system

    DOE Patents [OSTI]

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

    1989-01-01

    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.

  16. Advanced Reactors Transition Program Resource Loaded Schedule

    SciTech Connect (OSTI)

    BOWEN, W.W.

    1999-11-08

    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.

  17. Advanced Reactors Transition Program Resource Loaded Schedule

    SciTech Connect (OSTI)

    GANTT, D.A.

    2000-01-12

    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 (FETF) 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 revision reflects the 19 Oct 1999 baseline.

  18. Adaptive Nodal Transport Methods for Reactor Transient Analysis

    SciTech Connect (OSTI)

    Thomas Downar; E. Lewis

    2005-08-31

    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.

  19. BEAM PROPAGATOR

    Energy Science and Technology Software Center (OSTI)

    003691MLTPL00 Beam Propagator for Weather Radars, Modules 1 and 2 http://www.exelisvis.com/ProductsServices/IDL.aspx

  20. Beam History

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

    Beam Status Beam History Print Beamline History Request Form To request a beam current histograph from the ALS storage ring beam histograph database, select the year, month, and day, then click on "Submit Request". Histographs are available as far back as February 2, 1994. Year 2016 2015 2014 2013 2012 2011 2010 2009 2008 2007 2006 2005 2004 2003 2002 2001 2000 1999 1998 1997 1996 1995 1994 Month Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec Day 01 02 03 04 05 06 07 08 09 10 11 12 13

  1. Advanced Safeguards Approaches for New Fast Reactors

    SciTech Connect (OSTI)

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

    2007-12-15

    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.

  2. Ion source development for a photoneutralization based NBI system for fusion reactors

    SciTech Connect (OSTI)

    Simonin, A.; Esch, H. P. L. de; Garibaldi, P.; Grand, C.; Bechu, S.; Bs, A.; Lacoste, A.

    2015-04-08

    The next step after ITER is to demonstrate the viability and generation of electricity by a future fusion reactor (DEMO). The specifications required to operate an NBI system on DEMO are very demanding. The system has to provide a very high level of power and energy, ~100MW of D beam at 1MeV, including high wall-plug efficiency (? > 60%). For this purpose, a new injector concept, called Siphore, is under investigation between CEA and French universities. Siphore is based on the stripping of the accelerated negative ions by photo-detachment provided by several Fabry-Perot cavities (3.5MW of light power per cavity) implemented along the D{sup ?} beam. The beamline is designed to be tall and narrow in order that the photon flux overlaps the entire negative ion beam. The paper will describe the present R and D at CEA which addresses the development of an ion source and pre-accelerator prototypes for Siphore, the main goal being to produce an intense negative ion beam sheet. The negative ion source Cybele is based on a magnetized plasma column where hot electrons are emitted from the source center. Parametric studies of the source are performed using Langmuir probes in order to characterize the plasma and to compare with numerical models being developed in French universities.

  3. On fast reactor kinetics studies

    SciTech Connect (OSTI)

    Seleznev, E. F.; Belov, A. A.; Matveenko, I. P.; Zhukov, A. M.; Raskach, K. F.

    2012-07-01

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

  4. NEUTRONIC REACTOR

    DOE Patents [OSTI]

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

    1958-09-01

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

  5. Pulsed ion beam source

    DOE Patents [OSTI]

    Greenly, J.B.

    1997-08-12

    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.

  6. HIGS Flux Performance Projection

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

    HIGS flux performance table for high-flux, quasi-CW operation, DFELL/TUNL, Nov. 9, 2010 (Version 2.3). HIGS Flux Performance Projection (2010 - 2011) Total Flux [g/s] CW Operation Two-Bunch (*) Collimated Flux (∆E γ /E γ = 5% FWHM) (#), (@) FEL λ [nm] Comment No-loss Mode : < 20 MeV Linear Pol. with OK-4 Circular Pol with OK-5 E γ = 1 - 2 MeV (E e = 237 - 336 MeV) 1 x 10 8 - 4 x 10 8 6 x 10 6 - 2.4 x 10 7 1064 Linear and Circular (a), (b) E γ = 2 - 2.9 MeV (E e = 336 - 405 MeV) 4 x 10

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

    SciTech Connect (OSTI)

    Rosenthal, Murray Wilford

    2009-08-01

    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.

  8. Catalytic reactor

    DOE Patents [OSTI]

    Aaron, Timothy Mark; Shah, Minish Mahendra; Jibb, Richard John

    2009-03-10

    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.

  9. POWER REACTOR

    DOE Patents [OSTI]

    Zinn, W.H.

    1958-07-01

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

  10. REACTOR CONTROL

    DOE Patents [OSTI]

    Fortescue, P.; Nicoll, D.

    1962-04-24

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

  11. Bioconversion reactor

    DOE Patents [OSTI]

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    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.

  12. NEUTRONIC REACTORS

    DOE Patents [OSTI]

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

    1958-10-14

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

  13. PHELIX for flux compression studies

    SciTech Connect (OSTI)

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

    2010-06-28

    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.

  14. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fermi, E.; Szilard, L.

    1957-09-24

    Reactors of the type employing plates of natural uranium in a moderator are discussed wherein the plates are um-formly disposed in parallel relationship to each other thereby separating the moderator material into distinct and individual layers. Each plate has an uninterrupted sunface area substantially equal to the cross-sectional area of the active portion of the reactor, the particular size of the plates and the volume ratio of moderator to uranium required to sustain a chain reaction being determinable from the known purity of these materials and other characteristics such as the predictable neutron losses due to the formation of radioactive elements of extremely high neutron capture cross section.

  15. NEUTRONIC REACTOR

    DOE Patents [OSTI]

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

    1958-04-15

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

  16. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Young, G.

    1963-01-01

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

  17. A brief History of Neutron Scattering at the Oak Ridge High Flux...

    Office of Scientific and Technical Information (OSTI)

    A brief History of Neutron Scattering at the Oak Ridge High Flux Isotope Reactor Citation Details In-Document Search Title: A brief History of Neutron Scattering at the Oak Ridge ...

  18. Measurement of neutrino flux from neutrino-electron elastic scattering

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Park, J.; Aliaga, L.; Altinok, O.; Bellantoni, L.; Bercellie, A.; Betancourt, M.; Bodek, A.; Bravar, A.; Budd, H.; Cai, T.; et al

    2016-06-10

    In muon-neutrino elastic scattering on electrons is an observable neutrino process whose cross section is precisely known. Consequently a measurement of this process in an accelerator-based νμ beam can improve the knowledge of the absolute neutrino flux impinging upon the detector; typically this knowledge is limited to ~10% due to uncertainties in hadron production and focusing. We also isolated a sample of 135±17 neutrino-electron elastic scattering candidates in the segmented scintillator detector of MINERvA, after subtracting backgrounds and correcting for efficiency. We show how this sample can be used to reduce the total uncertainty on the NuMI νμ flux frommore » 9% to 6%. Finally, our measurement provides a flux constraint that is useful to other experiments using the NuMI beam, and this technique is applicable to future neutrino beams operating at multi-GeV energies.« less

  19. Neutral beam dump with cathodic arc titanium gettering

    SciTech Connect (OSTI)

    Smirnov, A.; Korepanov, S. A.; Putvinski, S.; Krivenko, A. S.; Murakhtin, S. V.; Savkin, V. Ya.

    2011-03-15

    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.

  20. Homopolar Gun for Pulsed Spheromak Fusion Reactors II

    SciTech Connect (OSTI)

    Fowler, T

    2004-06-14

    A homopolar gun is discussed that could produce the high currents required for pulsed spheromak fusion reactors even with unit current amplification and open field lines during injection, possible because close coupling between the gun and flux conserver reduces gun losses to acceptable levels. Example parameters are given for a gun compatible with low cost pulsed reactors and for experiments to develop the concept.

  1. Self-actuating reactor shutdown system

    DOE Patents [OSTI]

    Barrus, Donald M.; Brummond, Willian A; Peterson, Leslie F.

    1988-01-01

    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.

  2. Beam tuning

    SciTech Connect (OSTI)

    Pardo, R.C.; Zinkann, G.P.

    1995-08-01

    A program for configuring the linac, based on previously run configurations for any desired beam was used during the past year. This program uses only a small number of empirical tunes to scale resonator fields to properly accelerate a beam with a different charge-to-mass (q/A) ratio from the original tune configuration. The program worked very well for the PII linac section where we can easily match a new beam`s arrival phase and velocity to the tuned value. It was also fairly successful for the Booster and ATLAS sections of the linac, but not as successful as for the PII linac. Most of the problems are associated with setting the beam arrival time correctly for each major linac section. This problem is being addressed with the development of the capacitive pickup beam phase monitor discussed above. During the next year we expect to improve our ability to quickly configure the linac for new beams and reduce the time required for linac tuning. Already the time required for linac tuning as a percentage of research hours has decreased from 22% in FY 1993 to 15% in the first quarter of FY 1995.

  3. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Starr, C.

    1963-01-01

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

  4. NEUTRONIC REACTOR

    DOE Patents [OSTI]

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

    1959-10-27

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

  5. Neutronic reactor

    DOE Patents [OSTI]

    Wende, Charles W. J.

    1976-08-17

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

  6. Neutronic reactor

    DOE Patents [OSTI]

    Carleton, John T.

    1977-01-25

    A graphite-moderated nuclear reactor includes channels between blocks of graphite and also includes spacer blocks between adjacent channeled blocks with an axis of extension normal to that of the axis of elongation of the channeled blocks to minimize changes in the physical properties of the graphite as a result of prolonged neutron bombardment.

  7. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Anderson, H.L.

    1958-10-01

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

  8. Beam History

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

    Beam History Print Beamline History Request Form To request a beam current histograph from the ALS storage ring beam histograph database, select the year, month, and day, then click on "Submit Request". Histographs are available as far back as February 2, 1994. Year 2016 2015 2014 2013 2012 2011 2010 2009 2008 2007 2006 2005 2004 2003 2002 2001 2000 1999 1998 1997 1996 1995 1994 Month Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec Day 01 02 03 04 05 06 07 08 09 10 11 12 13 14 15 16 17

  9. Beam History

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

    Beam History Print Beamline History Request Form To request a beam current histograph from the ALS storage ring beam histograph database, select the year, month, and day, then click on "Submit Request". Histographs are available as far back as February 2, 1994. Year 2016 2015 2014 2013 2012 2011 2010 2009 2008 2007 2006 2005 2004 2003 2002 2001 2000 1999 1998 1997 1996 1995 1994 Month Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec Day 01 02 03 04 05 06 07 08 09 10 11 12 13 14 15 16 17

  10. ARM - Measurement - Actinic flux

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

    would love to hear from you Send us a note below or call us at 1-888-ARM-DATA. Send Measurement : Actinic flux The quantity of light in the atmosphere available to molecules at a...

  11. ``Sleeping reactor`` irradiations: Shutdown reactor determination of short-lived activation products

    SciTech Connect (OSTI)

    Jerde, E.A.; Glasgow, D.C.

    1998-09-01

    At the High-Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory, the principal irradiation system has a thermal neutron flux ({phi}) of {approximately} 4 {times} 10{sup 14} n/cm{sup 2} {center_dot} s, permitting the detection of elements via irradiation of 60 s or less. Irradiations of 6 or 7 s are acceptable for detection of elements with half-lives of as little as 30 min. However, important elements such as Al, Mg, Ti, and V have half-lives of only a few minutes. At HFIR, these can be determined with irradiation times of {approximately} 6 s, but the requirement of immediate counting leads to increased exposure to the high activity produced by irradiation in the high flux. In addition, pneumatic system timing uncertainties (about {+-} 0.5 s) make irradiations of < 6 s less reliable. Therefore, the determination of these ultra-short-lived species in mixed matrices has not generally been made at HFIR. The authors have found that very short lived activation products can be produced easily during the period after reactor shutdown (SCRAM), but prior to the removal of spent fuel elements. During this 24- to 36-h period (dubbed the ``sleeping reactor``), neutrons are produced in the beryllium reflector by the reaction {sup 9}Be({gamma},n){sup 8}Be, the gamma rays principally originating in the spent fuel. Upon reactor SCRAM, the flux drops to {approximately} 1 {times} 10{sup 10} n/cm{sup 2} {center_dot} s within 1 h. By the time the fuel elements are removed, the flux has dropped to {approximately} 6 {times} 10{sup 8}. Such fluxes are ideal for the determination of short-lived elements such as Al, Ti, Mg, and V. An important feature of the sleeping reactor is a flux that is not constant.

  12. Beam Test Facility

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

    Beam Test Facility Beam Test Facility Print Tuesday, 20 October 2009 09:36 Coming Soon

  13. Measurements - Ion Beams - Radiation Effects Facility / Cyclotron Institute

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

    / Texas A&M University Ion Beams Available Beams / Beam Change Times / Measurements / Useful Graphs Measurements The beam uniformity and flux are determined using an array of five detectors. Each detector is made up with a plastic scintillator coupled to photo-multiplier tubes. Four of the detectors are fixed in position and set up to measure beam particle counting rates continuously at four characteristic points 1.64 inches (4.71 mm) away from the beam axis. The fifth scintillator can

  14. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Treshow, M.

    1958-08-19

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

  15. REACTOR UNLOADING

    DOE Patents [OSTI]

    Leverett, M.C.

    1958-02-18

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

  16. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Wade, E.J.

    1958-09-16

    This patent relates to a reflector means for a neutronic reactor. A reflector comprised of a plurality of vertically movable beryllium control members is provided surrounding the sides of the reactor core. An absorber of fast neutrons comprised of natural uramum surrounds the reflector. An absorber of slow neutrons surrounds the absorber of fast neutrons and is formed of a plurality of beryllium blocks having natural uranium members distributcd therethrough. in addition, a movable body is positioned directly below the core and is comprised of a beryllium reflector and an absorbing member attached to the botiom thereof, the absorbing member containing a substance selected from the goup consisting of natural urantum and Th/sup 232/.

  17. Nuclear reactor

    DOE Patents [OSTI]

    Wade, Elman E.

    1979-01-01

    A nuclear reactor including two rotatable plugs and a positive top core holddown structure. The top core holddown structure is divided into two parts: a small core cover, and a large core cover. The small core cover, and the upper internals associated therewith, are attached to the small rotating plug, and the large core cover, with its associated upper internals, is attached to the large rotating plug. By so splitting the core holddown structures, under-the-plug refueling is accomplished without the necessity of enlarging the reactor pressure vessel to provide a storage space for the core holddown structure during refueling. Additionally, the small and large rotating plugs, and their associated core covers, are arranged such that the separation of the two core covers to permit rotation is accomplished without the installation of complex lifting mechanisms.

  18. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Grebe, J.J.

    1959-12-15

    A reactor which is particularly adapted tu serve as a heat source for a nuclear powered alrcraft or rocket is described. The core of this reactor consists of a porous refractory modera;or body which is impregnated with fissionable nuclei. The core is designed so that its surface forms tapered inlet and outlet ducts which are separated by the porous moderator body. In operation a gaseous working fluid is circulated through the inlet ducts to the surface of the moderator, enters and passes through the porous body, and is heated therein. The hot gas emerges into the outlet ducts and is available to provide thrust. The principle advantage is that tremendous quantities of gas can be quickly heated without suffering an excessive pressure drop.

  19. NUCLEAR REACTORS

    DOE Patents [OSTI]

    Long, E.; Ashby, J.W.

    1958-09-16

    ABS>A graphite moderator structure is presented for a nuclear reactor compriscd of an assembly of similarly orientated prismatic graphite blocks arranged on spaced longitudinal axes lying in common planes wherein the planes of the walls of the blocks are positioned so as to be twisted reintive to the planes of said axes so thatthe unlmpeded dtrect paths in direction wholly across the walls of the blocks are limited to the width of the blocks plus spacing between the blocks.

  20. Neutronic reactor

    DOE Patents [OSTI]

    Lewis, Warren R.

    1978-05-30

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

  1. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Anderson, J.B.

    1960-01-01

    A reactor is described which comprises a tank, a plurality of coaxial steel sleeves in the tank, a mass of water in the tank, and wire grids in abutting relationship within a plurality of elongated parallel channels within the steel sleeves, the wire being provided with a plurality of bends in the same plane forming adjacent parallel sections between bends, and the sections of adjacent grids being normally disposed relative to each other.

  2. REACTOR CONTROL

    DOE Patents [OSTI]

    Ruano, W.J.

    1957-12-10

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

  3. Multimegawatt Space Reactor Safety

    SciTech Connect (OSTI)

    Stanley, M.L. )

    1989-01-01

    The Multimegawatt (MMW) Space Reactor Project supports the Strategic Defense Initiative Office requirement to provide reliable, safe, cost-effective, electrical power in the MMW range. Specifically, power may be used for neutral particle beams, free electron lasers, electromagnetic launchers, and orbital transfer vehicles. This power plant technology may also apply to the electrical power required for other uses such as deep-space probes and planetary exploration. The Multimegawatt Space Reactor Project, the Thermionic Fuel Element Verification Program, and Centaurus Program all support the Multimegawatt Space Nuclear Power Program and form an important part of the US Department of Energy's (DOE's) space and defense power systems activities. A major objective of the MMW project is the development of a reference flight system design that provides the desired levels of public safety, health protection, and special nuclear material (SNM) protection when used during its designated missions. The safety requirements for the MMW project are a hierarchy of requirements that consist of safety requirements/regulations, a safety policy, general safety criteria, safety technical specifications, safety design specifications, and the system design. This paper describes the strategy and philosophy behind the development of the safety requirements imposed upon the MMW concept developers. The safety organization, safety policy, generic safety issues, general safety criteria, and the safety technical specifications are discussed.

  4. Advanced Test Reactor - A National Scientific User Facility

    SciTech Connect (OSTI)

    Clifford J. Stanley

    2008-05-01

    The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected nuclear research reactor with a maximum operating power of 250 MWth. The unique serpentine configuration of the fuel elements creates five main reactor power lobes (regions) and nine flux traps. In addition to these nine flux traps there are 68 additional irradiation positions in the reactor core reflector tank. There are also 34 low-flux irradiation positions in the irradiation tanks outside the core reflector tank. The ATR is designed to provide a test environment for the evaluation of the effects of intense radiation (neutron and gamma). Due to the unique serpentine core design each of the five lobes can be operated at different powers and controlled independently. Options exist for the individual test trains and assemblies to be either cooled by the ATR coolant (i.e., exposed to ATR coolant flow rates, pressures, temperatures, and neutron flux) or to be installed in their own independent test loops where such parameters as temperature, pressure, flow rate, neutron flux, and energy can be controlled per experimenter specifications. The full-power maximum thermal neutron flux is ~1.0 x1015 n/cm2-sec with a maximum fast flux of ~5.0 x1014 n/cm2-sec. The Advanced Test Reactor, now a National Scientific User Facility, is a versatile tool in which a variety of nuclear reactor, nuclear physics, reactor fuel, and structural material irradiation experiments can be conducted. The cumulative effects of years of irradiation in a normal power reactor can be duplicated in a few weeks or months in the ATR due to its unique design, power density, and operating flexibility.

  5. Quantum flux parametron

    SciTech Connect (OSTI)

    Hioe, W. ); Goto, E. )

    1991-01-01

    The quantum flux parametron (QFP) is an offspring of the parametron, an early flux-based logic device, and the Josephson junction. It is a single flux quantum device that works completely in the superconductive mode. While it has the speed of other Josephson devices that work on switching between the voltage and superconductive modes, its power is about one thousand times less. Hence, it promises to be an attractive alternative to both transistors and other Josephson devices. This book reports the latest research results on QFP applications as a logic device. In particular, a number of auxiliary circuits and a new logic gate are proposed for improving the device margin. Samples of these circuits and logic gate have been fabricated.

  6. High Flux Isotope Reactor | Neutron Science at ORNL

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

    HFIR is also used for medical, industrial, and research isotope production; research on severe neutron damage to materials; and neutron activation analysis to examine trace ...

  7. Modeling and Simulations for the High Flux Isotope Reactor Cycle...

    Office of Scientific and Technical Information (OSTI)

    serve as reference for the design of an LEU fuel for HFIR. ... critical experiments at HFIR, data included in the current HFIR safety analysis report, andor data from ...

  8. B Reactor | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Operational Management History Manhattan Project Signature Facilities B Reactor B Reactor B Reactor Completed in September 1944, the B Reactor was the world's first ...

  9. Analysis of the Reactor Position Independent Monitor (PIM) Diagnostic

    SciTech Connect (OSTI)

    Hayes-Sterbenz, Anna Catherine

    2014-07-17

    In this note I analyze the physics determining the proposed reactor position independent monitor (PIM), which is the ratio (240Pu/239Pu)1/3 × (135Cs/137Cs)1/2. The PIM ratios in any reactor fuel is shown to increase monotonically with the time over which the fuel is irradiated. This is because the Cs ratio determines the neutron flux, while the Pu isotopic ratio is determined by the flux times the irradiation time. If the irradiation time for all fuel rods across the reactor is fixed, the PIM ratio is approximately constant in all rods. However, no information can be extracted from the PIM ratio on Pu isotopics unless both the flux (or Cs ratio) and the irradiation time (from, say, Ru isotopics) are known separately, i.e., the PIM ratio is not a fundamental parameter of any reactor. Thus, unless the PIM ratio has been measured for the specific fuel under interrogation, no information can be deduced from measurements or reactor simulations of PIM ratios in different fuel from the same reactor. However, if a PIM measurement has been in one spent fuel rod from a given reactor, all other rods that are known to have been in the reactor for the same irradiation period can be assumed to have approximately the same PIM ratio.

  10. Nuclear reactor

    DOE Patents [OSTI]

    Pennell, William E.; Rowan, William J.

    1977-01-01

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

  11. NUCLEAR REACTORS

    DOE Patents [OSTI]

    Long, E.; Ashley, J.W.

    1958-12-16

    A graphite moderator structure is described for a gas-cooled nuclear reactor having a vertical orlentation wherein the structure is physically stable with regard to dlmensional changes due to Wigner growth properties of the graphite, and leakage of coolant gas along spaces in the structure is reduced. The structure is comprised of stacks of unlform right prismatic graphite blocks positioned in layers extending in the direction of the lengths of the blocks, the adjacent end faces of the blocks being separated by pairs of tiles. The blocks and tiles have central bores which are in alignment when assembled and are provided with cooperatlng keys and keyways for physical stability.

  12. Photocatalytic reactor

    DOE Patents [OSTI]

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

    1999-01-19

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

  13. Radiative Flux Analysis

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

    Long, Chuck [NOAA

    2008-05-14

    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.

  14. Hybrid adsorptive membrane reactor

    DOE Patents [OSTI]

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    2011-03-01

    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.

  15. H Reactor - Hanford Site

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

    About Us Projects & Facilities H Reactor About Us About Hanford Cleanup Hanford History ... 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and ...

  16. C Reactor - Hanford Site

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

    C Reactor About Us About Hanford Cleanup Hanford History Hanford Site Wide Programs ... 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and ...

  17. F Reactor - Hanford Site

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

    About Us Projects & Facilities F Reactor About Us About Hanford Cleanup Hanford History ... 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and ...

  18. N Reactor - Hanford Site

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

    Projects & Facilities N Reactor About Us About Hanford Cleanup Hanford History Hanford ... 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and ...

  19. Power flow control using distributed saturable reactors

    DOE Patents [OSTI]

    Dimitrovski, Aleksandar D.

    2016-02-13

    A magnetic amplifier includes a saturable core having a plurality of legs. Control windings wound around separate legs are spaced apart from each other and connected in series in an anti-symmetric relation. The control windings are configured in such a way that a biasing magnetic flux arising from a control current flowing through one of the plurality of control windings is substantially equal to the biasing magnetic flux flowing into a second of the plurality of control windings. The flow of the control current through each of the plurality of control windings changes the reactance of the saturable core reactor by driving those portions of the saturable core that convey the biasing magnetic flux in the saturable core into saturation. The phasing of the control winding limits a voltage induced in the plurality of control windings caused by a magnetic flux passing around a portion of the saturable core.

  20. Power Burst Facility (PBF) Reactor Reactor Decommissioning

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

    Click here to view Click here to view Reactor Decommissioning Click on an image to enlarge A crane removes the reactor vessel from the Power Burst Facility (top), then places it ...

  1. Control system for a small fission reactor

    DOE Patents [OSTI]

    Burelbach, James P.; Kann, William J.; Saiveau, James G.

    1986-01-01

    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.

  2. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Stewart, H.B.

    1958-12-23

    A nuclear reactor of the type speclfically designed for the irradiation of materials is discussed. In this design a central cyllndrical core of moderating material ls surrounded by an active portlon comprlsed of an annular tank contalning fissionable material immersed ln a liquid moderator. The active portion ls ln turn surrounded by a reflector, and a well ls provided in the center of the core to accommodate the materlals to be irradiated. The over-all dimensions of the core ln at least one plane are equal to or greater than twice the effective slowing down length and equal to or less than twlce the effective diffuslon length for neutrons in the core materials.

  3. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Daniels, F.

    1962-12-18

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

  4. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Wigner, E.P.

    1957-09-17

    A reactor of the type having coolant liquid circulated through clad fuel elements geometrically arranged in a solid moderator, such as graphite, is described. The core is enclosed in a pressure vessel and suitable shielding, wherein means is provided for circulating vapor through the core to superheat the same. This is accomplished by drawing off the liquid which has been heated in the core due to the fission of the fuel, passing it to a nozzle within a chamber where it flashes into a vapor, and then passing the vapor through separate tubes extending through the moderator to pick up more heat developed in the core due to the fission of the fuel, thereby producing superheated vapor.

  5. Nuclear reactor

    DOE Patents [OSTI]

    Yant, Howard W.; Stinebiser, Karl W.; Anzur, Gregory C.

    1977-01-01

    A nuclear reactor, particularly a liquid-metal breeder reactor, whose upper internals include outlet modules for channeling the liquid-metal coolant from selected areas of the outlet of the core vertically to the outlet plenum. The modules are composed of a highly-refractory, high corrosion-resistant alloy, for example, INCONEL-718. Each module is disposed to confine and channel generally vertically the coolant emitted from a subplurality of core-component assemblies. Each module has a grid with openings, each opening disposed to receive the coolant from an assembly of the subplurality. The grid in addition serves as a holdown for the assemblies of the corresponding subplurality preventing their excessive ejection upwardly from the core. In the region directly over the core the outlet modules are of such peripheral form that they nest forming a continuum over the core-component assemblies whose outlet coolant they confine. Each subassembly includes a chimney which confines the coolant emitted by its corresponding subassemblies to generally vertical flow between the outlet of the core and the outlet plenum. Each subplurality of assemblies whose emitted coolant is confined by an outlet module includes assemblies which emit lower-temperature coolant, for example, a control-rod assembly, or fertile assemblies, and assemblies which emit coolant of substantially higher temperature, for example, fuel-rod assemblies. The coolants of different temperatures are mixed in the chimneys reducing the effect of stripping (hot-cold temperature fluctuations) on the remainder of the upper internals which are composed typically of AISI-304 or AISI-316 stainless steel.

  6. Beam current controller for laser ion source

    DOE Patents [OSTI]

    Okamura, Masahiro

    2014-10-28

    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.

  7. Optical heat flux gauge

    DOE Patents [OSTI]

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

    1991-04-09

    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.

  8. High sensitivity charge amplifier for ion beam uniformity monitor

    DOE Patents [OSTI]

    Johnson, Gary W.

    2001-01-01

    An ion beam uniformity monitor for very low beam currents using a high-sensitivity charge amplifier with bias compensation. The ion beam monitor is used to assess the uniformity of a raster-scanned ion beam, such as used in an ion implanter, and utilizes four Faraday cups placed in the geometric corners of the target area. Current from each cup is integrated with respect to time, thus measuring accumulated dose, or charge, in Coulombs. By comparing the dose at each corner, a qualitative assessment of ion beam uniformity is made possible. With knowledge of the relative area of the Faraday cups, the ion flux and areal dose can also be obtained.

  9. IN-CORE FLUX SENSOR EVALUATIONS AT THE ATR CRITICAL FACILITY.

    SciTech Connect (OSTI)

    Troy Unruh; Benjamin Chase; Joy Rempe; David Nigg; George Imel; Jason Harris; Todd Sherman; Jean-Francois VIllard

    2014-12-01

    As part of an Idaho State University (ISU)–led Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) collaborative project that includes Idaho National Laboratory (INL) and the French Alternative Energies and Atomic Energy Commission (CEA), flux detector evaluations were completed to compare their accuracy, response time, and longduration performance. Special fixturing, developed by INL, allows real-time flux detectors to be inserted into various Advanced Test Reactor Critical Facility (ATRC) core positions to perform lobe power measurements, axial flux profile measurements, and detector crosscalibrations. Detectors initially evaluated in this program included miniature fission chambers, specialized self-powered neutron detectors (SPNDs), and specially developed commercial SPNDs. Results from this program provide important insights related to flux detector accuracy and resolution for subsequent ATR and CEA experiments and yield new flux data required for benchmarking models in the ATR Life Extension Program (LEP) Modeling Update Project.

  10. Method to Reduce Neutron Production in Small Clean Fusion Reactors Inventor

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

    --- Samuel A. Cohen | Princeton Plasma Physics Lab Method to Reduce Neutron Production in Small Clean Fusion Reactors Inventor --- Samuel A. Cohen This invention describes a method to reduce neutron production of D-3He-fueled, steady state, small FRC fusion reactors using periodic, co-streaming, energetic ion beams generated by RF. Use of this method will lessen damage to and activation of reactor components and, in doing so, can advance the development of fusion reactors for electrical

  11. Characterization of X-ray generator beam profiles.

    SciTech Connect (OSTI)

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

    2013-07-01

    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.

  12. Reactor and method of operation

    DOE Patents [OSTI]

    Wheeler, John A.

    1976-08-10

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

  13. Effect of rolling motion on critical heat flux for subcooled flow boiling in vertical tube

    SciTech Connect (OSTI)

    Hwang, J. S.; Park, I. U.; Park, M. Y.; Park, G. C.

    2012-07-01

    This paper presents defining characteristics of the critical heat flux (CHF) for the boiling of R-134a in vertical tube operation under rolling motion in marine reactor. It is important to predict CHF of marine reactor having the rolling motion in order to increase the safety of the reactor. Marine Reactor Moving Simulator (MARMS) tests are conducted to measure the critical heat flux using R-134a flowing upward in a uniformly heated vertical tube under rolling motion. MARMS was rotated by motor and mechanical power transmission gear. The CHF tests were performed in a 9.5 mm I.D. test section with heated length of 1 m. Mass fluxes range from 285 to 1300 kg m{sup -2}s{sup -1}, inlet subcooling from 3 to 38 deg. C and outlet pressures from 13 to 24 bar. Amplitudes of rolling range from 15 to 40 degrees and periods from 6 to 12 sec. To convert the test conditions of CHF test using R-134a in water, Katto's fluid-to-fluid modeling was used in present investigation. A CHF correlation is presented which accounts for the effects of pressure, mass flux, inlet subcooling and rolling angle over all conditions tested. Unlike existing transient CHF experiments, CHF ratio of certain mass flux and pressure are different in rolling motion. For the mass fluxes below 500 kg m{sup -2}s{sup -1} at 13, 16 (region of relative low mass flux), CHF ratio was decreased but was increased above that mass flux (region of relative high mass flux). Moreover, CHF tend to enhance in entire mass flux at 24 bar. (authors)

  14. ARM - Measurement - Sensible heat flux

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

    heat flux ARM Data Discovery Browse Data Comments? We would love to hear from you Send us a note below or call us at 1-888-ARM-DATA. Send Measurement : Sensible heat flux The time ...

  15. ARM - Measurement - Latent heat flux

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

    heat flux ARM Data Discovery Browse Data Comments? We would love to hear from you Send us a note below or call us at 1-888-ARM-DATA. Send Measurement : Latent heat flux The time ...

  16. CEBAF beam loss accounting

    SciTech Connect (OSTI)

    Ursic, R.; Mahoney, K.; Hovater, C.; Hutton, A.; Sinclair, C.

    1995-12-31

    This paper describes the design and implementation of a beam loss accounting system for the CEBAF electron accelerator. This system samples the beam curent throughout the beam path and measures the beam current accurately. Personnel Safety and Machine Protection systems use this system to turn off the beam when hazardous beam losses occur.

  17. Confined ion beam sputtering device and method

    DOE Patents [OSTI]

    Sharp, D.J.

    1986-03-25

    A hollow cylindrical target, lined internally with a sputter deposit material and open at both ends, surrounds a substrate on which sputtered deposition is to be obtained. An ion beam received through either one or both ends of the open cylindrical target is forced by a negative bias applied to the target to diverge so that ions impinge at acute angles at different points of the cylindrical target surface. The ion impingement results in a radially inward and downstream directed flux of sputter deposit particles that are received by the substrate. A positive bias applied to the substrate enhances divergence of the approaching ion beams to generate a higher sputtered deposition flux rate. Alternatively, a negative bias applied to the substrate induces the core portion of the ion beams to reach the substrate and provide ion polishing of the sputtered deposit thereon.

  18. Confined ion beam sputtering device and method

    DOE Patents [OSTI]

    Sharp, Donald J.

    1988-01-01

    A hollow cylindrical target, lined internally with a sputter deposit material and open at both ends, surrounds a substrate on which sputtered deposition is to be obtained. An ion beam received through either one or both ends of the open cylindrical target is forced by a negative bias applied to the target to diverge so that ions impinge at acute angles at different points of the cylindrical target surface. The ion impingement results in a radially inward and downstream directed flux of sputter deposit particles that are received by the substrate. A positive bias applied to the substrate enhances divergence of the approaching ion beams to generate a higher sputtered deposition flux rate. Alternatively, a negative bias applied to the substrate induces the core portion of the ion beams to reach the substrate and provide ion polishing of the sputtered deposit thereon.

  19. Light Water Reactor Sustainability (LWRS) Program | Department...

    Energy Savers [EERE]

    Nuclear Reactor Technologies Light Water Reactor Sustainability (LWRS) Program Light Water Reactor Sustainability (LWRS) Program Light Water Reactor Sustainability (LWRS) ...

  20. SRS Small Modular Reactors

    ScienceCinema (OSTI)

    None

    2014-05-21

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

  1. Reactor safety method

    DOE Patents [OSTI]

    Vachon, Lawrence J.

    1980-03-11

    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.

  2. Portable radiography system using a relativistic electron beam

    DOE Patents [OSTI]

    Hoeberling, R.F.

    1987-09-22

    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.

  3. Portable radiography system using a relativistic electron beam

    DOE Patents [OSTI]

    Hoeberling, Robert F.

    1990-01-01

    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.

  4. Nuclear reactor

    DOE Patents [OSTI]

    Thomson, Wallace B.

    2004-03-16

    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.

  5. High flux, narrow bandwidth compton light sources via extended laser-electron interactions

    DOE Patents [OSTI]

    Barty, V P

    2015-01-13

    New configurations of lasers and electron beams efficiently and robustly produce high flux beams of bright, tunable, polarized quasi-monoenergetic x-rays and gamma-rays via laser-Compton scattering. Specifically, the use of long-duration, pulsed lasers and closely-spaced, low-charge and low emittance bunches of electron beams increase the spectral flux of the Compton-scattered x-rays and gamma rays, increase efficiency of the laser-electron interaction and significantly reduce the overall complexity of Compton based light sources.

  6. Fast Flux Test Facility Closure Project - Project Management Plan

    SciTech Connect (OSTI)

    BEACH, R.R.

    2002-09-26

    The Fast Flux Test Facility (FFTF) Closure Project, Project Management Plan, Revision 5, provides the scope, cost, and schedule to achieve the most cost effective and expeditious closure of the FFTF to an assumed final end-state with the reactor vessel and the containment building, below the 5504 grade level, being entombed in place. Closure will be completed by December 2009 at a cost of $547 million.

  7. Failure analysis of beryllium tile assembles following high heat flux testing for the ITER program

    SciTech Connect (OSTI)

    B. C. Odegard, Jr.; C. H. Cadden; N. Y. C. Yang

    2000-05-01

    The following document describes the processing, testing and post-test analysis of two Be-Cu assemblies that have successfully met the heat load requirements for the first wall and dome sections for the ITER (International Thermonuclear Experimental Reactor) fusion reactor. Several different joint assemblies were evaluated in support of a manufacturing technology investigation aimed at diffusion bonding or brazing a beryllium armor tile to a copper alloy heat sink for fusion reactor applications. Judicious selection of materials and coatings for these assemblies was essential to eliminate or minimize interactions with the highly reactive beryllium armor material. A thin titanium layer was used as a diffusion barrier to isolate the copper heat sink from the beryllium armor. To reduce residual stresses produced by differences in the expansion coefficients between the beryllium and copper, a compliant layer of aluminum or aluminum-beryllium (AlBeMet-150) was used. Aluminum was chosen because it does not chemically react with, and exhibits limited volubility in, beryllium. Two bonding processes were used to produce the assemblies. The primary process was a diffusion bonding technique. In this case, undesirable metallurgical reactions were minimized by keeping the materials in a solid state throughout the fabrication cycle. The other process employed an aluminum-silicon layer as a brazing filler material. In both cases, a hot isostatic press (HIP) furnace was used in conjunction with vacuum-canned assemblies in order to minimize oxidation and provide sufficient pressure on the assemblies for full metal-to-metal contact and subsequent bonding. The two final assemblies were subjected to a suite of tests including: tensile tests and electron and optical metallography. Finally, high heat flux testing was conducted at the electron beam testing system (EBTS) at Sandia National Laboratories, New Mexico. Here, test mockups were fabricated and subjected to normal heat loads to

  8. Beam geometry selection using sequential beam addition

    SciTech Connect (OSTI)

    Popple, Richard A. Brezovich, Ivan A.; Fiveash, John B.

    2014-05-15

    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

  9. Neutrino Flux Prediction for the NuMI Beamline

    SciTech Connect (OSTI)

    Aliaga Soplin, Leonidas

    2016-01-01

    The determination of the neutrino flux in any conventional neutrino beam presents a challenge for the current and future short and long baseline neutrino experiments. The uncertainties associated with the production and attenuation of the hadrons in the beamline materials along with those associated with the beam optics have a big effect in the flux spectrum knowledge. For experiments like MINERvA, understanding the flux is crucial since it enters directly into every neutrino-nucleus cross-sections measurements. The foundation of this work is predicting the neutrino flux at MINERvA using dedicated measurements of hadron production in hadron-nucleus collisions and incorporating in-situ MINERvA data that can provide additional constraints. This work also includes the prospect for predicting the flux at other detectors like the NOvA Near detector. The procedure and conclusions of this thesis will have a big impact on future hadron production experiments and on determining the fl ux for the upcoming DUNE experiment.

  10. Prospects for Tokamak Fusion Reactors

    SciTech Connect (OSTI)

    Sheffield, J.; Galambos, J.

    1995-04-01

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

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

    SciTech Connect (OSTI)

    NIELSEN, D L

    2004-02-26

    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.

  12. BDDR, a new CEA technological and operating reactor database

    SciTech Connect (OSTI)

    Soldevilla, M.; Salmons, S.; Espinosa, B.

    2013-07-01

    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)

  13. Once-through CANDU reactor models for the ORIGEN2 computer code

    SciTech Connect (OSTI)

    Croff, A.G.; Bjerke, M.A.

    1980-11-01

    Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % /sup 235/U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given.

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

    SciTech Connect (OSTI)

    John Jackson; Todd Allen; Frances Marshall; Jim Cole

    2013-03-01

    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

  15. Laser beam monitoring system

    DOE Patents [OSTI]

    Weil, Bradley S.; Wetherington, Jr., Grady R.

    1985-01-01

    Laser beam monitoring systems include laser-transparent plates set at an angle to the laser beam passing therethrough and light sensor for detecting light reflected from an object on which the laser beam impinges.

  16. Laser beam monitoring system

    DOE Patents [OSTI]

    Weil, B.S.; Wetherington, G.R. Jr.

    Laser beam monitoring systems include laser-transparent plates set at an angle to the laser beam passing therethrough and light sensor for detecting light reflected from an object on which the laser beam impinges.

  17. Relativistic electron beam generator

    DOE Patents [OSTI]

    Mooney, L.J.; Hyatt, H.M.

    1975-11-11

    A relativistic electron beam generator for laser media excitation is described. The device employs a diode type relativistic electron beam source having a cathode shape which provides a rectangular output beam with uniform current density.

  18. BEAMS: Curiosity | Jefferson Lab

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

    BEAMS: Curiosity BEAMS: Curiosity January 9, 2013 BEAMS, Becoming Excited About Math and Science, is one of our education programs. In particular, it is the only one in which I ...

  19. Characterization of local heat fluxes around ICRF antennas on JET

    SciTech Connect (OSTI)

    Campergue, A.-L.; Jacquet, P.; Monakhov, I.; Arnoux, G.; Brix, M.; Sirinelli, A.; Milanesio, D.; Colas, L.; Collaboration: JET-EFDA Contributors

    2014-02-12

    When using Ion Cyclotron Range of Frequency (ICRF) heating, enhanced power deposition on Plasma-Facing Components (PFCs) close to the antennas can occur. Experiments have recently been carried out on JET with the new ITER-Like-Wall (ILW) to characterize the heat fluxes on the protection of the JET ICRF antennas, using Infra-Red (IR) thermography measurement. The measured heat flux patterns along the poloidal limiters surrounding powered antennas were compared to predictions from a simple RF sheath rectification model. The RF electric field, parallel to the static magnetic field in front of the antenna, was evaluated using the TOPICA code, integrating a 3D flattened model of the JET A2 antennas. The poloidal density variation in front of the limiters was obtained from the mapping of the Li-beam or edge reflectometry measurements using the flux surface geometry provided by EFIT equilibrium reconstruction. In many cases, this simple model can well explain the position of the maximum heat flux on the different protection limiters and the heat-flux magnitude, confirming that the parallel RF electric field and the electron plasma density in front of the antenna are the main driving parameters for ICRF-induced local heat fluxes.

  20. Beam imaging sensor

    DOE Patents [OSTI]

    McAninch, Michael D; Root, Jeffrey J

    2015-03-31

    The present invention relates generally to the field of sensors for beam imaging and, in particular, to a new and useful beam imaging sensor for use in determining, for example, the power density distribution of a beam including, but not limited to, an electron beam or an ion beam. In one embodiment, the beam imaging sensor of the present invention comprises, among other items, a circumferential slit that is either circular, elliptical or polygonal in nature.

  1. Beam imaging sensor

    DOE Patents [OSTI]

    McAninch, Michael D.; Root, Jeffrey J.

    2016-07-05

    The present invention relates generally to the field of sensors for beam imaging and, in particular, to a new and useful beam imaging sensor for use in determining, for example, the power density distribution of a beam including, but not limited to, an electron beam or an ion beam. In one embodiment, the beam imaging sensor of the present invention comprises, among other items, a circumferential slit that is either circular, elliptical or polygonal in nature.

  2. Attrition reactor system

    DOE Patents [OSTI]

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

    1993-09-28

    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.

  3. Attrition reactor system

    DOE Patents [OSTI]

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

    1993-01-01

    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.

  4. Hybrid plasmachemical reactor

    SciTech Connect (OSTI)

    Lelevkin, V. M. Smirnova, Yu. G.; Tokarev, A. V.

    2015-04-15

    A hybrid plasmachemical reactor on the basis of a dielectric barrier discharge in a transformer is developed. The characteristics of the reactor as functions of the dielectric barrier discharge parameters are determined.

  5. Technical aspects of boron neutron capture therapy at the BNL Medical Research Reactor

    SciTech Connect (OSTI)

    Holden, N.E.; Rorer, D.C.; Patti, F.J.; Liu, H.B.; Reciniello, R.; Chanana, A.D.

    1997-07-01

    The Brookhaven Medical Research Reactor, BMRR, is a 3 MW heterogeneous, tank-type, light water cooled and moderated, graphite reflected reactor, which was designed for biomedical studies. Early BNL work in Boron Neutron Capture Therapy (BNCT) used a beam of thermal neutrons for experimental treatment of brain tumors. Research elsewhere and at BNL indicated that higher energy neutrons would be required to treat deep seated brain tumors. Epithermal neutrons would be thermalized as they penetrated the brain and peak thermal neutron flux densities would occur at the depth of brain tumors. One of the two BMRR thermal port shutters was modified in 1988 to include plates of aluminum and aluminum oxide to provide an epithermal port. Lithium carbonate in polyethylene was added in 1991 around the bismuth port to reduce the neutron flux density coming from outside the port. To enhance the epithermal neutron flux density, the two vertical thimbles A-3 (core edge) and E-3 (in core) were replaced with fuel elements. There are now four fuel elements of 190 grams each and 28 fuel elements of 140 grams each for a total of 4.68 kg of {sup 235}U in the core. The authors have proposed replacing the epithermal shutter with a fission converter plate shutter. It is estimated that the new shutter would increase the epithermal neutron flux density by a factor of seven and the epithermal/fast neutron ratio by a factor of two. The modifications made to the BMRR in the past few years permit BNCT for brain tumors without the need to reflect scalp and bone flaps. Radiation workers are monitored via a TLD badge and a self-reading dosimeter during each experiment. An early concern was raised about whether workers would be subject to a significant dose rate from working with patients who have been irradiated. The gamma ray doses for the representative key personnel involved in the care of the first 12 patients receiving BNCT are listed. These workers did not receive unusually high exposures.

  6. BEAM INSTRUMENTATION FOR HIGH POWER HADRON BEAMS

    SciTech Connect (OSTI)

    Aleksandrov, Alexander V

    2013-01-01

    This presentation will describe developments in the beam diagnostics which support the understanding and operation of high power hadron accelerators. These include the measurement of large dynamic range transverse and longitudinal beam profiles, beam loss detection, and non-interceptive diagnostics.

  7. Further investigation of the “reactor anomaly”

    SciTech Connect (OSTI)

    Garvey, G. T. Hayes, A. C. Jungman, Gerard; Jonkmans, G.

    2015-07-15

    The effect of a more realistic and extensive inclusion of first forbidden beta decay into the determination of the reactor neutrino flux is investigated. Forbidden decays make up approximately 30% of all fission product decays so their possible impact on the neutrino flux should not be neglected. Because of an incomplete knowledge of the requisite nuclear structure it is not possible to incorporate the forbidden decays in an exact fashion thus a variety of scenarios are investigated. It appears that including first forbidden decays measurably modifies the anti-neutrino spectrum, and the uncertainty on the neutrino flux should be expanded beyond 4%.

  8. Reactor System Transient Code.

    Energy Science and Technology Software Center (OSTI)

    1999-07-14

    RELAP3B describes the behavior of water-cooled nuclear reactors during postulated accidents or power transients, such as large reactivity excursions, coolant losses or pump failures. The program calculates flows, mass and energy inventories, pressures, temperatures, and steam qualities along with variables associated with reactor power, reactor heat transfer, or control systems. Its versatility allows one to describe simple hydraulic systems as well as complex reactor systems.

  9. Period meter for reactors

    DOE Patents [OSTI]

    Rusch, Gordon K.

    1976-01-06

    An improved log N amplifier type nuclear reactor period meter with reduced probability for noise-induced scrams is provided. With the reactor at low power levels a sampling circuit is provided to determine the reactor period by measuring the finite change in the amplitude of the log N amplifier output signal for a predetermined time period, while at high power levels, differentiation of the log N amplifier output signal provides an additional measure of the reactor period.

  10. NEUTRONIC REACTOR POWER PLANT

    DOE Patents [OSTI]

    Metcalf, H.E.

    1962-12-25

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

  11. Development of the beam extraction synchronization system at the Fermilab Booster

    SciTech Connect (OSTI)

    Seiya, K.; Chaurize, S.; Drennan, C. C.; Pellico, W.; Sullivan, T.; Triplett, A. K.; Waller, A. M.

    2015-07-28

    The new beam extraction synchronization control system called “Magnetic Cogging” was developed at the Fermilab Booster and it replaces a system called “RF Cogging” as part of the Proton Improvement Plan (PIP). [1] The flux throughput goal for the PIP is 2.2×1017 protons per hour, which is double the present flux. Thus, the flux increase will be accomplished by doubling the number of beam cycles which, in turn, will double the beam loss in the Booster accelerator if nothing else is done.

  12. Development of the beam extraction synchronization system at the Fermilab Booster

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Seiya, K.; Chaurize, S.; Drennan, C. C.; Pellico, W.; Sullivan, T.; Triplett, A. K.; Waller, A. M.

    2015-07-28

    The new beam extraction synchronization control system called “Magnetic Cogging” was developed at the Fermilab Booster and it replaces a system called “RF Cogging” as part of the Proton Improvement Plan (PIP). [1] The flux throughput goal for the PIP is 2.2×1017 protons per hour, which is double the present flux. Thus, the flux increase will be accomplished by doubling the number of beam cycles which, in turn, will double the beam loss in the Booster accelerator if nothing else is done.

  13. Development of the beam extraction synchronization system at the Fermilab Booster

    SciTech Connect (OSTI)

    Seiya, K.; Chaurize, S.; Drennan, C. C.; Pellico, W.; Sullivan, T.; Triplett, A. K.; Waller, A. M.

    2015-07-28

    The new beam extraction synchronization control system called Magnetic Cogging was developed at the Fermilab Booster and it replaces a system called RF Cogging as part of the Proton Improvement Plan (PIP). [1] The flux throughput goal for the PIP is 2.21017 protons per hour, which is double the present flux. Thus, the flux increase will be accomplished by doubling the number of beam cycles which, in turn, will double the beam loss in the Booster accelerator if nothing else is done.

  14. Preserving physics knowledge at the fast flux test facility

    SciTech Connect (OSTI)

    Wootan, D.; Omberg, R.; Makenas, B. J.; Polzin, D. L.

    2012-07-01

    One of the goals of the Dept. 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. A disciplined and orderly approach has been developed 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. (authors)

  15. NEUTRONIC REACTOR SHIELDING

    DOE Patents [OSTI]

    Borst, L.B.

    1961-07-11

    A special hydrogenous concrete shielding for reactors is described. In addition to Portland cement and water, the concrete essentially comprises 30 to 60% by weight barytes aggregate for enhanced attenuation of fast neutrons. The biological shields of AEC's Oak Ridge Graphite Reactor and Materials Testing Reactor are particular embodiments.

  16. Advanced Test Reactor Tour

    ScienceCinema (OSTI)

    Miley, Don

    2013-05-28

    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.

  17. High solids fermentation reactor

    DOE Patents [OSTI]

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

    1993-01-01

    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.

  18. High solids fermentation reactor

    DOE Patents [OSTI]

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

    1993-03-02

    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.

  19. Improved vortex reactor system

    DOE Patents [OSTI]

    Diebold, James P.; Scahill, John W.

    1995-01-01

    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.

  20. Advanced Test Reactor Tour

    SciTech Connect (OSTI)

    Miley, Don

    2011-01-01

    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.

  1. NUCLEAR REACTOR CONTROL SYSTEM

    DOE Patents [OSTI]

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

    1959-11-01

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

  2. The New Uppsala Neutron Beam Facility

    SciTech Connect (OSTI)

    Pomp, S.; Blomgren, J.; Hildebrand, A.; Johansson, C.; Mermod, P.; Oesterlund, M.; Prokofiev, A.V.; Bystroem, O.; Ekstroem, C.; Haag, N.; Jonsson, O.; Reistad, D.; Renberg, P.-U.; Wessman, D.; Ziemann, V.; Nilsson, L.; Olsson, N.; Tippawan, U.

    2005-05-24

    A new quasi-monoenergetic neutron beam facility has been constructed at the The Svedberg Laboratory (TSL) in Uppsala, Sweden. Key features include an energy range of 20 to 175 MeV, high fluxes, and the possibility of large-area fields. Besides cross-section measurements, the new facility has been designed specifically to provide optimal conditions for testing of single-event effects in electronics and for dosimetry development. First results of the beam characterization measurements performed in early 2004 are reported.

  3. NOVEL CRYOGENIC ENGINEERING SOLUTIONS FOR THE NEW AUSTRALIAN RESEARCH REACTOR OPAL

    SciTech Connect (OSTI)

    Olsen, S. R.; Kennedy, S. J.; Kim, S.; Schulz, J. C.; Thiering, R.; Gilbert, E. P.; Lu, W.; James, M.; Robinson, R. A.

    2008-03-16

    In August 2006 the new 20MW low enriched uranium research reactor OPAL went critical. The reactor has 3 main functions, radio pharmaceutical production, silicon irradiation and as a neutron source. Commissioning on 7 neutron scattering instruments began in December 2006. Three of these instruments (Small Angle Neutron Scattering, Reflectometer and Time-of-flight Spectrometer) utilize cold neutrons.The OPAL Cold Neutron Source, located inside the reactor, is a 20L liquid deuterium moderated source operating at 20K, 330kPa with a nominal refrigeration capacity of 5 kW and a peak flux at 4.2meV (equivalent to a wavelength of 0.4nm). The Thermosiphon and Moderator Chamber are cooled by helium gas delivered at 19.8K using the Brayton cycle. The helium is compressed by two 250kW compressors (one with a variable frequency drive to lower power consumption).A 5 Tesla BSCCO (2223) horizontal field HTS magnet will be delivered in the 2{sup nd} half of 2007 for use on all the cold neutron instruments. The magnet is cooled by a pulse tube cryocooler operating at 20K. The magnet design allows for the neutron beam to pass both axially and transverse to the field. Samples will be mounted in a 4K to 800K Gifford-McMahon (GM) cryofurnace, with the ability to apply a variable electric field in-situ. The magnet is mounted onto a tilt stage. The sample can thus be studied under a wide variety of conditions.A cryogen free 7.4 Tesla Nb-Ti vertical field LTS magnet, commissioned in 2005 will be used on neutron diffraction experiments. It is cooled by a standard GM cryocooler operating at 4.2K. The sample is mounted in a 2{sup nd} GM cryocooler (4K-300K) and a variable electric field can be applied.

  4. Beam halo in mismatched proton beams.

    SciTech Connect (OSTI)

    Wangler, Thomas P.,; Allen, C. K.; Chan, D.; Colestock, P. L. ,; Crandall, K. R.; Qiang, J.; Garnett, R. W.; Lysenko, W. P.; Gilpatrick, J. D.; Schneider, J. D.; Schulze, M. E.; Sheffield, R. L.; Smith, H. V.

    2002-01-01

    Progress was made during the past decade towards a better understanding of halo formation caused by beam mismatch in high-intensity beams. To test these ideas an experiment was carried out at Los Alamos with proton beams in a 52-quadrupole focusing channel. Rms emittances and beam widths were obtained from measured beam profiles for comparison with the maximum emittance growth predictions of a free-energy model and the maximum haloamplitude predictions of a particle-core model. The experimental results are also compared with multiparticle simulations. In this paper we will present the experimental results and discuss the implications with respect to the validity of both the models and the simulations. Keywords: beam halo, emittance growth, beam profiles, simulations, space charge, mismatch

  5. PHLUX: Photographic Flux Tools for Solar Glare and Flux

    Energy Science and Technology Software Center (OSTI)

    2010-12-02

    A web-based tool to a) analytically and empirically quantify glare from reflected light and determine the potential impact (e.g., temporary flash blindness, retinal burn), and b) produce flux maps for central receivers. The tool accepts RAW digital photographs of the glare source (for hazard assessment) or the receiver (for flux mapping), as well as a photograph of the sun for intensity and size scaling. For glare hazard assessment, the tool determines the retinal irradiance (W/cm2)more » and subtended source angle for an observer and plots the glare source on a hazard spectrum (i.e., low-potential for flash blindness impact, potential for flash blindness impact, retinal burn). For flux mapping, the tool provides a colored map of the receiver scaled by incident solar flux (W/m2) and unwraps the physical dimensions of the receiver while accounting for the perspective of the photographer (e.g., for a flux map of a cylindrical receiver, the horizontal axis denotes receiver angle in degrees and the vertical axis denotes vertical position in meters; for a flat panel receiver, the horizontal axis denotes horizontal position in meters and the vertical axis denotes vertical position in meters). The flux mapping capability also allows the user to specify transects along which the program plots incident solar flux on the receiver.« less

  6. Flux growth utilizing the reaction between flux and crucible

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Yan, J. -Q.

    2015-01-22

    Flux growth involves dissolving the components of the target compound in an appropriate flux at high temperatures and then crystallizing under supersaturation controlled by cooling or evaporating the flux. A refractory crucible is generally used to contain the high temperature melt. Moreover, the reaction between the melt and crucible materials can modify the composition of the melt, which typically results in growth failure, or contaminates the crystals. Thus one principle in designing a flux growth is to select suitable flux and crucible materials thus to avoid any reaction between them. In this paper, we review two cases of flux growthmore » in which the reaction between flux and Al2O3 crucible tunes the oxygen content in the melt and helps the crystallization of desired compositions. For the case of La5Pb3O, the Al2O3 crucible oxidizes La to form a passivating La2O3 layer which not only prevents further oxidization of La in the melt but also provides [O] to the melt. Finally, in the case of La0.4Na0.6Fe2As2, it is believed that the Al2O3 crucible reacts with NaAsO2 and the reaction consumes oxygen in the melt thus maintaining an oxygen-free environment.« less

  7. Reactor vessel support system

    DOE Patents [OSTI]

    Golden, Martin P.; Holley, John C.

    1982-01-01

    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.

  8. Nuclear reactor overflow line

    DOE Patents [OSTI]

    Severson, Wayne J.

    1976-01-01

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

  9. Hanging core support system for a nuclear reactor. [LMFBR

    DOE Patents [OSTI]

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

    1984-04-26

    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.

  10. Spinning fluids reactor

    DOE Patents [OSTI]

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

    2012-11-20

    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.

  11. Reactor water cleanup system

    DOE Patents [OSTI]

    Gluntz, Douglas M.; Taft, William E.

    1994-01-01

    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.

  12. Reactor water cleanup system

    DOE Patents [OSTI]

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

    1994-12-20

    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.

  13. Straw man 900-1000 GeV crystal extraction test beam for Fermilab collider operation

    SciTech Connect (OSTI)

    Carrigan, R.A. Jr.

    1996-10-01

    A design for a 900-1000 GeV, 100 khz parasitic test beam for use during collider operations has been developed. The beam makes use of two bent crystals, one for extraction and the other one for redirecting the beam in to the present Switchyard beam system. The beam requires only a few modifications in the A0 area and largely uses existing devices. It should be straight-forward to modify one or two beam lines in the fixed target experimental areas to work above 800 GeV. Possibilities for improvements to the design,to operate at higher fluxes are discussed.

  14. Telecommunication using muon beams

    DOE Patents [OSTI]

    Arnold, Richard C.

    1976-01-01

    Telecommunication is effected by generating a beam of mu mesons or muons, varying a property of the beam at a modulating rate to generate a modulated beam of muons, and detecting the information in the modulated beam at a remote location.

  15. Ion Beam Materials Lab

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

    Ion Beam Materials Lab Ion Beam Materials Lab A new research frontier awaits! Our door is open and we thrive on mutually beneficial partnerships, collaborations that drive innovations and new technologies. April 12, 2012 Ion Beam Danfysik Implanter High Voltage Terminal. Contact Yongqiang Wang (505) 665-1596 Email Devoted to the characterization and modification of surfaces through the use of ion beams The Ion Beam Materials Laboratory (IBML) is a Los Alamos National Laboratory resource devoted

  16. BEAM HALO IN PROTON LINAC BEAMS

    SciTech Connect (OSTI)

    T. WANGLER; K. CRANDALL

    2000-08-01

    In this paper we review the present picture of km halo in proton linacs. Space-charge forces acting in mismatched beams have been identified as a major cause of beam-halo. We present a definition of halo based on a ratio of moments of the distribution of the beam coordinates. We find from our initial studies that for halo detined in this way, a beam can have rms emittance growth without halo growth, but halo growth is always accompanied by rms emittance growth. We describe the beam-halo experiment that is in preparation at Los Alamos, which will address questions about the beam profiles, maximum particle amplitudes, and rms emittance growth associated with the halo.

  17. Korea Research Reactor -1 & 2 Decommissioning Project in Korea

    SciTech Connect (OSTI)

    Park, S. K.; Chung, U. S.; Jung, K. J.; Park, J. H.

    2003-02-24

    Korea Research Reactor 1 (KRR-1), the first research reactor in Korea, has been operated since 1962, and the second one, Korea Research Reactor 2 (KRR-2) since 1972. The operation of both of them was phased out in 1995 due to their lifetime and operation of the new and more powerful research reactor, HANARO (High-flux Advanced Neutron Application Reactor; 30MW). Both are TRIGA Pool type reactors in which the cores are small self-contained units sitting in tanks filled with cooling water. The KRR-1 is a TRIGA Mark II, which could operate at a level of up to 250 kW. The second one, the KRR-2 is a TRIGA Mark III, which could operate at a level of up 2,000 kW. The decontamination and decommissioning (D & D) project of these two research reactors, the first D & D project in Korea, was started in January 1997 and will be completed to stage 3 by 2008. The aim of this decommissioning program is to decommission the KRR-1 & 2 reactors and to decontaminate the residual building structure s and the site to release them as unrestricted areas. KAERI (Korea Atomic Energy Research Institute) submitted the decommissioning plan and the environmental impact assessment reports to the Ministry of Science and Technology (MOST) for the license in December 1998, and was approved in November 2000.

  18. ARM - Measurement - Soil moisture flux

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

    moisture flux ARM Data Discovery Browse Data Comments? We would love to hear from you! Send us a note below or call us at 1-888-ARM-DATA. Send Measurement : Soil moisture flux A quantity measured according to the formula B = {lambda}(dq/dz), where {lambda} is the conductivity of the soil that the moisture is moving through. Categories Surface Properties Instruments The above measurement is considered scientifically relevant for the following instruments. Refer to the datastream (netcdf) file

  19. Reactor Chamber and Balance-of-Plant Characteristics for a Fast-Ignition Heavy-Ion Fusion Power Plant

    SciTech Connect (OSTI)

    Medin, Stanislav; Churazov, Mikhail; Koshkarev, Dmitri; Sharkov, Boris; Orlov, Yurii; Suslin, Viktor; Zemskov, Eugeni

    2003-05-15

    The concept of a fast-ignition heavy-ion fusion (FIHIF) power plant involves a cylindrical target and superhigh energy ion beams. The driver produces one plus/minus charge state multimass platinum ions with energy of 100 GeV. The driver efficiency and the target gain are taken as 0.25 and 100, respectively. The preliminary data on the energy fluxes delivered to the reactor chamber wall by the 500-MJ fusion yield are presented. The reactor chamber designed has two sections. In the first section, the microexplosions occur, and in the second section of bigger volume the expansion and condensation of vapors take place. The response of the blanket and the thin liquid film at the first-wall surface is evaluated. Lithium-lead eutectic is taken as a coolant. The evaporated mass and the condensation time are estimated, taking into account major thermophysical effects. The estimated neutron spectrum from the FIHIF target gives an average neutron energy of 11.9 MeV. The mechanical stresses in the construction material due to neutron energy release are evaluated. The outlet coolant chamber temperature is taken as 550 deg. C. The heat conversion system consisting of three coolant loops provides a net efficiency of the FIHIF power plant of 0.37.

  20. 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)

    Newton, T. H.; Wilson, E. H; Bergeron, A.; Horelik, N.; Stevens, J.

    2011-03-02

    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

  1. Design of a Modular E-Core Flux Concentrating Axial Flux Machine...

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

    Design of a Modular E-Core Flux Concentrating Axial Flux Machine Preprint Tausif Husain, 1 ... Design of a Modular E-Core Flux Concentrating Axial Flux Machine Tausif Husain (1) Yilmaz ...

  2. ION BEAM COLLIMATOR

    DOE Patents [OSTI]

    Langsdorf, A.S. Jr.

    1957-11-26

    A device is described for defining a beam of high energy particles wherein the means for defining the beam in the horizontal and vertical dimension are separately adjustable and the defining members are internally cooled. In general, the device comprises a mounting block having a central opening through which the beam is projected, means for rotatably supporting two pairs of beam- forming members, passages in each member for the flow of coolant; the beam- forming members being insulated from each other and the block, and each having an end projecting into the opening. The beam-forming members are adjustable and may be cooperatively positioned to define the beam passing between the end of the members. To assist in projecting and defining the beam, the member ends have individual means connected thereto for indicating the amount of charge collected thereon due to beam interception.

  3. Flux growth utilizing the reaction between flux and crucible

    SciTech Connect (OSTI)

    Yan, J. -Q.

    2015-01-22

    Flux growth involves dissolving the components of the target compound in an appropriate flux at high temperatures and then crystallizing under supersaturation controlled by cooling or evaporating the flux. A refractory crucible is generally used to contain the high temperature melt. Moreover, the reaction between the melt and crucible materials can modify the composition of the melt, which typically results in growth failure, or contaminates the crystals. Thus one principle in designing a flux growth is to select suitable flux and crucible materials thus to avoid any reaction between them. In this paper, we review two cases of flux growth in which the reaction between flux and Al2O3 crucible tunes the oxygen content in the melt and helps the crystallization of desired compositions. For the case of La5Pb3O, the Al2O3 crucible oxidizes La to form a passivating La2O3 layer which not only prevents further oxidization of La in the melt but also provides [O] to the melt. Finally, in the case of La0.4Na0.6Fe2As2, it is believed that the Al2O3 crucible reacts with NaAsO2 and the reaction consumes oxygen in the melt thus maintaining an oxygen-free environment.

  4. Benchmarking transition costs for the Fast Flux Test Facility

    SciTech Connect (OSTI)

    Hulvey, R.K.

    1996-12-31

    The Fast Flux Test Facility (FFTF) is a government-owned, 400-MW(thermal), sodium-cooled test reactor operated by Westinghouse Hanford Company. The reactor is shut down and is undergoing a transition to a long-term surveillance and maintenance state. The mission strategy for the FFTF transition project is to place the FFTF in a radiologically and industrially safe condition, completing the transition phase activities as soon as possible to drive down the current annual surveillance and maintenance costs from approximately $26 million/yr to roughly $1.5 million/yr. The effort to establish the shutdown and transition costs for this 7-yr, $260 million activity is a first of a kind for the U.S. Department of Energy (DOE).

  5. NEUTRONIC REACTOR SYSTEM

    DOE Patents [OSTI]

    Goett, J.J.

    1961-01-24

    A system is described which includes a neutronic reactor containing a dispersion of fissionable material in a liquid moderator as fuel and a conveyor to which a portion of the dispersion may be passed and wherein the self heat of the slurry evaporates the moderator. Means are provided for condensing the liquid moderator and returning it to the reactor and for conveying the dried fissionable material away from the reactor.

  6. Improved vortex reactor system

    DOE Patents [OSTI]

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

    1995-05-09

    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.

  7. Small Modular Reactors (SMRs) | Department of Energy

    Energy Savers [EERE]

    Reactor Technologies Small Modular Reactors (SMRs) Small Modular Reactors (SMRs) ... to the NRC by late-2016 Complete reactor module final design by mid-2019 For more ...

  8. Fast flux test facility radioisotope production and medical applications

    SciTech Connect (OSTI)

    Schenter, R.E.; Smith, S.G.; Tenforde, T.S.

    1997-12-01

    The Fast Flux Test Facility (FFTF) is a 400-MW, sodium-cooled reactor that operated successfully from 1982 to 1992, conducting work in support of the liquid-metal reactor industry by developing and testing fuel assemblies, control rods, and other core reactor components. Upon termination of this program, the primary mission of FFTF ended, and it was placed in a standby mode in 1993. However, in January 1997 the U.S. Secretary of Energy requested that FFTF be evaluated for a future mission that would consist of a primary goal of producing tritium for nuclear defense applications and a secondary goal of supplying medical isotopes for research and clinical applications. Production by FFTF of tritium for U.S. nuclear weapons would augment the dual-track strategy now under consideration for providing a long-term tritium supply in the United States (consisting of a light water reactor option and an accelerator option). A decision by the Secretary of Energy on proceeding with steps leading toward the possible reactivation of FFTF will be made before the end of 1998.

  9. Knowledge Preservation at the Fast Flux Test Facility

    SciTech Connect (OSTI)

    Wootan, David W.; Omberg, Ronald P.

    2011-12-30

    One of the goals of the Department of Energy's Office of Nuclear Energy Fuel Cycle Research and Development Program (FCRD) 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. 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. A disciplined and orderly approach has been developed 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.

  10. Beam position monitor

    DOE Patents [OSTI]

    Alkire, Randy W.; Rosenbaum, Gerold; Evans, Gwyndaf

    2003-07-22

    An apparatus for determining the position of an x-ray beam relative to a desired beam axis. Where the apparatus is positioned along the beam path so that a thin metal foil target intersects the x-ray beam generating fluorescent radiation. A PIN diode array is positioned so that a portion of the fluorescent radiation is intercepted by the array resulting in an a series of electrical signals from the PIN diodes making up the array. The signals are then analyzed and the position of the x-ray beam is determined relative to the desired beam path.

  11. Pyramid beam splitter

    DOE Patents [OSTI]

    McKeown, Mark H.; Beason, Steven C.; Fairer, George

    1992-01-01

    The apparatus of the present invention provides means for obtaining accurate, dependable, measurement of bearings and directions for geologic mapping in subterranean shafts, such as, for example, nuclear waste storage investigations. In operation, a laser beam is projected along a reference bearing. A pyramid is mounted such that the laser beam is parallel to the pyramid axis and can impinge on the apex of the pyramid thus splitting the beam several ways into several beams at right angles to each other and at right angles to the reference beam. The pyramid is also translatable and rotatable in a plane perpendicular to the reference beam.

  12. Tokamak reactor first wall

    DOE Patents [OSTI]

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

    1984-11-20

    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.

  13. Small Modular Reactors - SRSCRO

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

    smr Small Modular Reactors The Savannah River National Laboratory (SRNL) has announced several partnerships to bring refrigerator-sized modular nuclear reactors, known as Small Modular Reactors or SMRs, to the Savannah River Site facility and jump start development of the U.S. Energy Freedom CenterTM. Currently, all large commercial power reactors in the United States and most in the rest of the world are based on "light water" designs - that is, they use uranium fuel and ordinary

  14. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, Juhani

    1996-01-01

    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.

  15. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, J.

    1996-03-19

    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.

  16. HOMOGENEOUS NUCLEAR POWER REACTOR

    DOE Patents [OSTI]

    King, L.D.P.

    1959-09-01

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

  17. Method of fission heat flux determination from experimental data

    DOE Patents [OSTI]

    Paxton, Frank A.

    1999-01-01

    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.

  18. AmeriFlux US-Sta Saratoga

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

    Ewers, Brent [University of Wyoming; Pendall, Elise [University of Wyoming

    2016-01-01

    This is the AmeriFlux version of the carbon flux data for the site US-Sta Saratoga. Site Description - Sagebrush steppe ecosystem

  19. AmeriFlux US-Wdn Walden

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

    Ewers, Brent [University of Wyoming; Pendall, Elise [University of Wyoming

    2016-01-01

    This is the AmeriFlux version of the carbon flux data for the site US-Wdn Walden. Site Description - Sagebrush steppe ecosystem

  20. Non-Vacuum Electron Beam Welding

    SciTech Connect (OSTI)

    Hershcovitch, Ady

    2007-01-31

    &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.

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

    SciTech Connect (OSTI)

    Odette, G. Robert; Yamamoto, Takuya

    2013-06-17

    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

  2. Code System for Reactor Physics and Fuel Cycle Simulation.

    Energy Science and Technology Software Center (OSTI)

    1999-04-21

    Version 00 VSOP94 (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based on neutron flux synthesis with depletion and shut-down features, in-core and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employed to accelerate the iterativemore » processes and to optimize the internal data transfer. The code system has been used extensively for comparison studies of reactors, their fuel cycles, and related detailed features. In addition to its use in research and development work for the High Temperature Reactor, the system has been applied successfully to Light Water and Heavy Water Reactors.« less

  3. Background radiation measurements at high power research reactors

    SciTech Connect (OSTI)

    Ashenfelter, J.; Yeh, M.; Balantekin, B.; Baldenegro, C. X.; Band, H. R.; Barclay, G.; Bass, C. D.; Berish, D.; Bowden, N. S.; Bryan, C. D.; Cherwinka, J. J.; Chu, R.; Classen, T.; Davee, D.; Dean, D.; Deichert, G.; Dolinski, M. J.; Dolph, J.; Dwyer, D. A.; Fan, S.; Gaison, J. K.; Galindo-Uribarri, A.; Gilje, K.; Glenn, A.; Green, M.; Han, K.; Hans, S.; Heeger, K. M.; Heffron, B.; Jaffe, D. E.; Kettell, S.; Langford, T. J.; Littlejohn, B. R.; Martinez, D.; McKeown, R. D.; Morrell, S.; Mueller, P. E.; Mumm, H. P.; Napolitano, J.; Norcini, D.; Pushin, D.; Romero, E.; Rosero, R.; Saldana, L.; Seilhan, B. S.; Sharma, R.; Stemen, N. T.; Surukuchi, P. T.; Thompson, S. J.; Varner, R. L.; Wang, W.; Watson, S. M.; White, B.; White, C.; Wilhelmi, J.; Williams, C.; Wise, T.; Yao, H.; Yen, Y. -R.; Zhang, C.; Zhang, X.

    2015-10-23

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including γ-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the background fields encountered. Furthermore, the general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.

  4. Background radiation measurements at high power research reactors

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Ashenfelter, J.; Yeh, M.; Balantekin, B.; Baldenegro, C. X.; Band, H. R.; Barclay, G.; Bass, C. D.; Berish, D.; Bowden, N. S.; Bryan, C. D.; et al

    2015-10-23

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including γ-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the backgroundmore » fields encountered. Furthermore, the general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.« less

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

    SciTech Connect (OSTI)

    Snoj, L.; Sklenka, L.; Rataj, J.; Boeck, H.

    2012-07-01

    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)

  6. Microsoft Word - beam characterization and verification.doc

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

    Beam Characterization and Verification Detector Components and Arrangement The beam uniformity and flux are determined using an array of five particle detectors. Each detector consists of Bicron BC-400 scintillator, a Bicron BC-634A optical coupling pad, a Hamamatsu R1635 photomultiplier tube, and a Hamamatsu E1761-04 tube base. Four of the detectors are fixed in position as show in Figure 1 and set up to measure beam particle counting rates continuously at four characteristic points, each 1.64

  7. PARTICLE BEAM TRACKING CIRCUIT

    DOE Patents [OSTI]

    Anderson, O.A.

    1959-05-01

    >A particle-beam tracking and correcting circuit is described. Beam induction electrodes are placed on either side of the beam, and potentials induced by the beam are compared in a voltage comparator or discriminator. This comparison produces an error signal which modifies the fm curve at the voltage applied to the drift tube, thereby returning the orbit to the preferred position. The arrangement serves also to synchronize accelerating frequency and magnetic field growth. (T.R.H.)

  8. Beam Dynamics for ARIA

    SciTech Connect (OSTI)

    Ekdahl, Carl August Jr.

    2014-10-14

    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.

  9. 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...

  10. Critical Heat Flux Phenomena at HighPressure & Low Mass Fluxes: NEUP Final Report Part I: Experiments

    SciTech Connect (OSTI)

    Corradini, Michael; Wu, Qiao

    2015-04-30

    This report is a preliminary document presenting an overview of the Critical Heat Flux (CHF) phenomenon, the High Pressure Critical Heat Flux facility (HPCHF), preliminary CHF data acquired, and the future direction of the research. The HPCHF facility has been designed and built to study CHF at high pressure and low mass flux ranges in a rod bundle prototypical of conceptual Small Modular Reactor (SMR) designs. The rod bundle is comprised of four electrically heated rods in a 2x2 square rod bundle with a prototypic chopped-cosine axial power profile and equipped with thermocouples at various axial and circumferential positions embedded in each rod for CHF detection. Experimental test parameters for CHF detection range from pressures of ~80 – 160 bar, mass fluxes of ~400 – 1500 kg/m2s, and inlet water subcooling from ~30 – 70°C. The preliminary data base established will be further extended in the future along with comparisons to existing CHF correlations, models, etc. whose application ranges may be applicable to the conditions of SMRs.

  11. NEUTRONIC REACTOR BURIAL ASSEMBLY

    DOE Patents [OSTI]

    Treshow, M.

    1961-05-01

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

  12. NEUTRONIC REACTOR SYSTEM

    DOE Patents [OSTI]

    Treshow, M.

    1959-02-10

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

  13. Light water reactor program

    SciTech Connect (OSTI)

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  14. REFLECTOR FOR NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Fraas, A.P.

    1963-08-01

    A reflector for nuclear reactors that comprises an assembly of closely packed graphite rods disposed with their major axes substantially perpendicular to the interface between the reactor core and the reflector is described. Each graphite rod is round in transverse cross section at (at least) its interface end and is provided, at that end, with a coaxial, inwardly tapering hole. (AEC)

  15. Enhancing the quality of radiographic images acquired with point-like gamma-ray sources through correction of the beam divergence and attenuation

    SciTech Connect (OSTI)

    Silvani, M. I.; Almeida, G. L.; Lopes, R. T.

    2014-11-11

    Radiographic images acquired with point-like gamma-ray sources exhibit a desirable low penumbra effects specially when positioned far away from the set object-detector. Such an arrangement frequently is not affordable due to the limited flux provided by a distant source. A closer source, however, has two main drawbacks, namely the degradation of the spatial resolution - as actual sources are only approximately punctual - and the non-homogeneity of the beam hitting the detector, which creates a false attenuation map of the object being inspected. This non-homogeneity is caused by the beam divergence itself and by the different thicknesses traversed the beam even if the object were an homogeneous flat plate. In this work, radiographic images of objects with different geometries, such as flat plates and pipes have undergone a correction of beam divergence and attenuation addressing the experimental verification of the capability and soundness of an algorithm formerly developed to generate and process synthetic images. The impact of other parameters, including source-detector gap, attenuation coefficient, ratio defective-to-main hull thickness and counting statistics have been assessed for specifically tailored test-objects aiming at the evaluation of the ability of the proposed method to deal with different boundary conditions. All experiments have been carried out with an X-ray sensitive Imaging Plate and reactor-produced {sup 198}Au and {sup 165}Dy sources. The results have been compared with other technique showing a better capability to correct the attenuation map of inspected objects unveiling their inner structure otherwise concealed by the poor contrast caused by the beam divergence and attenuation, in particular for those regions far apart from the vertical of the source.

  16. Some Aspects of Reactor Theory

    DOE R&D Accomplishments [OSTI]

    Weinberg, Alvin M.

    1952-10-10

    Some general remarks are made on reactor theory, particularly the asymptotic theory and multigroup methods. Unsolved reactor problems are also briefly discussed. (B.J.H.)

  17. Reactor Materials | Department of Energy

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

    reactor materials crosscut effort will enable the development of innovative and ... Research into specific degradation modes or material needs unique to a particular reactor ...

  18. High flux solar energy transformation

    DOE Patents [OSTI]

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

    1991-04-09

    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.

  19. ARM - Measurement - Soil heat flux

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

    heat flux ARM Data Discovery Browse Data Comments? We would love to hear from you! Send us a note below or call us at 1-888-ARM-DATA. Send Measurement : Soil heat flux A quantity measured according to the formula B = {lambda}(dT/dz), where {lambda} is the conductivity of the soil that the heat is moving through. Categories Surface Properties Instruments The above measurement is considered scientifically relevant for the following instruments. Refer to the datastream (netcdf) file headers of each

  20. High flux solar energy transformation

    DOE Patents [OSTI]

    Winston, Roland; Gleckman, Philip L.; O'Gallagher, Joseph J.

    1991-04-09

    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.

  1. Neutral beam monitoring

    DOE Patents [OSTI]

    Fink, Joel H.

    1981-08-18

    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.

  2. BEAM CONTROL PROBE

    DOE Patents [OSTI]

    Chesterman, A.W.

    1959-03-17

    A probe is described for intercepting a desired portion of a beam of charged particles and for indicating the spatial disposition of the beam. The disclosed probe assembly includes a pair of pivotally mounted vanes moveable into a single plane with adjacent edges joining and a calibrated mechanical arrangement for pivoting the vancs apart. When the probe is disposed in the path of a charged particle beam, the vanes may be adjusted according to the beam current received in each vane to ascertain the dimension of the beam.

  3. SNS Sample Activation Calculator Flux Recommendations and Validation

    SciTech Connect (OSTI)

    McClanahan, Tucker C.; Gallmeier, Franz X.; Iverson, Erik B.; Lu, Wei

    2015-02-01

    The Spallation Neutron Source (SNS) at Oak Ridge National Laboratory (ORNL) uses the Sample Activation Calculator (SAC) to calculate the activation of a sample after the sample has been exposed to the neutron beam in one of the SNS beamlines. The SAC webpage takes user inputs (choice of beamline, the mass, composition and area of the sample, irradiation time, decay time, etc.) and calculates the activation for the sample. In recent years, the SAC has been incorporated into the user proposal and sample handling process, and instrument teams and users have noticed discrepancies in the predicted activation of their samples. The Neutronics Analysis Team validated SAC by performing measurements on select beamlines and confirmed the discrepancies seen by the instrument teams and users. The conclusions were that the discrepancies were a result of a combination of faulty neutron flux spectra for the instruments, improper inputs supplied by SAC (1.12), and a mishandling of cross section data in the Sample Activation Program for Easy Use (SAPEU) (1.1.2). This report focuses on the conclusion that the SAPEU (1.1.2) beamline neutron flux spectra have errors and are a significant contributor to the activation discrepancies. The results of the analysis of the SAPEU (1.1.2) flux spectra for all beamlines will be discussed in detail. The recommendations for the implementation of improved neutron flux spectra in SAPEU (1.1.3) are also discussed.

  4. Preserving Physics Knowledge at the Fast Flux Test Facility

    SciTech Connect (OSTI)

    Wootan, David W.; Omberg, Ronald P.; Makenas, Bruce J.; Nielsen, Deborah L.; Nelson, Joseph V.; Polzin, David L.

    2011-11-01

    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. A disciplined and orderly approach has been developed 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.

  5. Knowledge Management at the Fast Flux Test Facility

    SciTech Connect (OSTI)

    Wootan, David W.; Omberg, Ronald P.

    2013-06-01

    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.

  6. Gated beam imager for heavy ion beams

    SciTech Connect (OSTI)

    Ahle, Larry; Hopkins, Harvey S.

    1998-12-10

    As part of the work building a small heavy-ion induction accelerator ring, or recirculator, at Lawrence Livermore National Laboratory, a diagnostic device measuring the four-dimensional transverse phase space of the beam in just a single pulse has been developed. This device, the Gated Beam Imager (GBI), consists of a thin plate filled with an array of 100-micron diameter holes and uses a Micro Channel Plate (MCP), a phosphor screen, and a CCD camera to image the beam particles that pass through the holes after they have drifted for a short distance. By time gating the MCP, the time evolution of the beam can also be measured, with each time step requiring a new pulse.

  7. Gated beam imager for heavy ion beams

    SciTech Connect (OSTI)

    Ahle, L.; Hopkins, H.S.

    1998-12-01

    As part of the work building a small heavy-ion induction accelerator ring, or recirculator, at Lawrence Livermore National Laboratory, a diagnostic device measuring the four-dimensional transverse phase space of the beam in just a single pulse has been developed. This device, the Gated Beam Imager (GBI), consists of a thin plate filled with an array of 100-micron diameter holes and uses a Micro Channel Plate (MCP), a phosphor screen, and a CCD camera to image the beam particles that pass through the holes after they have drifted for a short distance. By time gating the MCP, the time evolution of the beam can also be measured, with each time step requiring a new pulse. {copyright} {ital 1998 American Institute of Physics.}

  8. Laser-fusion targets for reactors

    DOE Patents [OSTI]

    Nuckolls, John H.; Thiessen, Albert R.

    1987-01-01

    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.

  9. TEST-HOLE CONSTRUCTION FOR A NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Ohlinger, L.A.; Seitz, F.; Young, G.J.

    1959-02-17

    Test-hole construction is described for a reactor which provides safe and ready access to the neutron flux region for specimen materials which are to be irradiated therein. An elongated tubular thimble adapted to be inserted in the access hole through the wall of the reactor is constructed of aluminum and is provided with a plurality of holes parallel to the axis of the thimble for conveying the test specimens into position for irradiation, and a conduit for the circulation of coolant. A laminated shield formed of alternate layers of steel and pressed wood fiber is disposed lengthwise of the thimble near the outer end thereof.

  10. Digital, remote control system for a 2-MW research reactor

    SciTech Connect (OSTI)

    Battle, R.E.; Corbett, G.K.

    1988-01-01

    A fault-tolerant programmable logic controller (PLC) and operator workstations have been programmed to replace the hard-wired relay control system in the 2-MW Bulk Shielding Reactor (BSR) at Oak Ridge National Laboratory. In addition to the PLC and remote and local operator workstations, auxiliary systems for remote operation include a video system, an intercom system, and a fiber optic communication system. The remote control station, located at the High Flux Isotope Reactor 2.5 km from the BSR, has the capability of rector startup and power control. The system was designed with reliability and fail-safe features as important considerations. 4 refs., 3 figs.

  11. Nuclear reactor control column

    DOE Patents [OSTI]

    Bachovchin, Dennis M.

    1982-01-01

    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.

  12. Slurry reactor design studies

    SciTech Connect (OSTI)

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. ); Akgerman, A. ); Smith, J.M. )

    1990-06-01

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

  13. Reactor Safety Research Programs

    SciTech Connect (OSTI)

    Edler, S. K.

    1981-07-01

    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.

  14. Superconducting flux flow digital circuits

    DOE Patents [OSTI]

    Hietala, Vincent M.; Martens, Jon S.; Zipperian, Thomas E.

    1995-01-01

    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.

  15. Superconducting flux flow digital circuits

    DOE Patents [OSTI]

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

    1995-02-14

    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.

  16. Fast Breeder Reactor studies

    SciTech Connect (OSTI)

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

    1980-07-01

    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.

  17. CONTROL FOR NEUTRONIC REACTOR

    DOE Patents [OSTI]

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

    1959-03-31

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

  18. Microfluidic electrochemical reactors

    DOE Patents [OSTI]

    Nuzzo, Ralph G.; Mitrovski, Svetlana M.

    2011-03-22

    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.

  19. Nuclear reactor reflector

    DOE Patents [OSTI]

    Hopkins, Ronald J.; Land, John T.; Misvel, Michael C.

    1994-01-01

    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.

  20. Nuclear reactor reflector

    DOE Patents [OSTI]

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

    1994-06-07

    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.

  1. COOLED NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Binner, C.R.; Wilkie, C.B.

    1958-03-18

    This patent relates to a design for a reactor of the type in which a fluid coolant is flowed through the active portion of the reactor. This design provides for the cooling of the shielding material as well as the reactor core by the same fluid coolant. The core structure is a solid moderator having coolant channels in which are disposed the fuel elements in rod or slug form. The coolant fluid enters the chamber in the shield, in which the core is located, passes over the inner surface of said chamber, enters the core structure at the center, passes through the coolant channels over the fuel elements and out through exhaust ducts.

  2. Spherical torus fusion reactor

    DOE Patents [OSTI]

    Martin Peng, Y.K.M.

    1985-10-03

    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.

  3. Reactor safety assessment system

    SciTech Connect (OSTI)

    Sebo, D.E.; Bray, M.A.; King, M.A.

    1987-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSA is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category.

  4. NUCLEAR REACTOR FUEL SYSTEMS

    DOE Patents [OSTI]

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

    1959-09-15

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

  5. Displacement damage effects on CMOS APS image sensors induced by neutron irradiation from a nuclear reactor

    SciTech Connect (OSTI)

    Wang, Zujun Huang, Shaoyan; Liu, Minbo; Xiao, Zhigang; He, Baoping; Yao, Zhibin; Sheng, Jiangkun

    2014-07-15

    The experiments of displacement damage effects on CMOS APS image sensors induced by neutron irradiation from a nuclear reactor are presented. The CMOS APS image sensors are manufactured in the standard 0.35 ?m CMOS technology. The flux of neutron beams was about 1.33 10{sup 8} n/cm{sup 2}s. The three samples were exposed by 1 MeV neutron equivalent-fluence of 1 10{sup 11}, 5 10{sup 11}, and 1 10{sup 12} n/cm{sup 2}, respectively. The mean dark signal (K{sub D}), dark signal spike, dark signal non-uniformity (DSNU), noise (V{sub N}), saturation output signal voltage (V{sub S}), and dynamic range (DR) versus neutron fluence are investigated. The degradation mechanisms of CMOS APS image sensors are analyzed. The mean dark signal increase due to neutron displacement damage appears to be proportional to displacement damage dose. The dark images from CMOS APS image sensors irradiated by neutrons are presented to investigate the generation of dark signal spike.

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

    SciTech Connect (OSTI)

    Popa-Simil, L.

    2012-07-01

    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)

  7. Small Modular Nuclear Reactors: Parametric Modeling of Integrated...

    Office of Environmental Management (EM)

    Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel ... Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel ...

  8. P Reactor Grouting

    SciTech Connect (OSTI)

    2010-01-01

    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.

  9. Reactor hot spot analysis

    SciTech Connect (OSTI)

    Vilim, R.B.

    1985-08-01

    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.

  10. NEUTRONIC REACTOR STRUCTURE

    DOE Patents [OSTI]

    Daniels, F.

    1961-10-24

    A reactor core, comprised of vertical stacks of hexagonal blocks of beryllium oxide having axial cylindrical apertures extending therethrough and cylindrical rods of a sintered mixture of uranium dioxide and beryllium oxide, is described. (AEC)

  11. The Fast Flux Test Facility shutdown program plan

    SciTech Connect (OSTI)

    Guttenberg, S.; Jones, D.H.; Midgett, J.C.; Nielsen, D.L.

    1995-01-01

    The Fast Flux Test Facility (FFTF) is a 400 MWt sodium-cooled research reactor owned by the US Department of Energy (DOE) and operated by the Westinghouse Hanford Company (WHC) on the Hanford Site in southeastern Washington State. The decision was made by the DOE in December, 1993, to initiate shutdown of the FFTF. This paper describes the FFTF Transition Project Plan (1) (formerly the FFTF Shutdown Program Plan) which provides the strategy, major elements, and project baseline for transitioning the FFTF to an industrially and radiologically safe shutdown condition. The Plan, and its resource loaded schedule, indicate this transition can be achieved in a period of six to seven years at a cost of approximately $359 million. The transition activities include reactor defueling, fuel offload to dry cask storage, sodium drain and reaction, management of sodium residuals, shutdown of auxiliary systems, and preparation of appropriate environmental and regulatory documentation. Completion of these activities will involve resolution of many challenging and unique issues associated with shutdown of a large sodium reactor facility. At the conclusion of these activities, the FFTF will be in a safe condition for turnover to the Hanford Site Environmental Restoration Contractor for a long term surveillance and maintenance phase and decommissioning.

  12. Molten metal reactors

    DOE Patents [OSTI]

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

    2013-11-05

    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.

  13. Future reactor experiments

    SciTech Connect (OSTI)

    Wen, Liangjian

    2015-07-15

    The non-zero neutrino mixing angle θ{sub 13} has been discovered and precisely measured by the current generation short-baseline reactor neutrino experiments. It opens the gate of measuring the leptonic CP-violating phase and enables the neutrino mass ordering. The JUNO and RENO-50 proposals aim at resolving the neutrino mass ordering using reactors. The experiment design, physics sensitivity, technical challenges as well as the progresses of those two proposed experiments are reviewed in this paper.

  14. Compact power reactor

    DOE Patents [OSTI]

    Wetch, Joseph R.; Dieckamp, Herman M.; Wilson, Lewis A.

    1978-01-01

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

  15. K-Reactor readiness

    SciTech Connect (OSTI)

    Rice, P.D.

    1991-12-04

    This document describes some of the more significant accomplishments in the reactor restart program and details the magnitude and extent of the work completed to bring K-Reactor to a state of restart readiness. The discussion of restart achievements is organized into the three major categories of personnel, programs, and plant. Also presented is information on the scope and extent of internal and external oversight of the efforts, as well as some details on the startup plan.

  16. Particle beam injection system

    DOE Patents [OSTI]

    Jassby, Daniel L.; Kulsrud, Russell M.

    1977-01-01

    This invention provides a poloidal divertor for stacking counterstreaming ion beams to provide high intensity colliding beams. To this end, method and apparatus are provided that inject high energy, high velocity, ordered, atomic deuterium and tritium beams into a lower energy, toroidal, thermal equilibrium, neutral, target plasma column that is magnetically confined along an endless magnetic axis in a strong restoring force magnetic field having helical field lines to produce counterstreaming deuteron and triton beams that are received bent, stacked and transported along the endless axis, while a poloidal divertor removes thermal ions and electrons all along the axis to increase the density of the counterstreaming ion beams and the reaction products resulting therefrom. By balancing the stacking and removal, colliding, strong focused particle beams, reaction products and reactions are produced that convert one form of energy into another form of energy.

  17. F Reactor Inspection

    SciTech Connect (OSTI)

    Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

    2014-10-29

    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."

  18. F Reactor Inspection

    ScienceCinema (OSTI)

    Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

    2014-11-24

    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."

  19. Beam Instrumentation Workshop

    SciTech Connect (OSTI)

    Shafer, R.E. )

    1994-01-01

    The fifth annual Beam Instrumentation Workshop was hosted by Los Alamos National Laboratory in Santa Fe, New Mexico. These proceedings represent the papers presented at the Workshop. A variety of topics were covered including beam emittance diagnostics, fluorescent screens, control systems for many accelerators and photon sources. Beam monitoring was discussed in great detail. There were thirty seven papers presented at the Workshop and all have been abstracted for the Energy and Science Technology database. (AIP)

  20. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Jennifer Lyons; Wade R. Marcum; Mark D. DeHart; Sean R. Morrell

    2014-01-01

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  1. NEUTRONIC REACTOR CONSTRUCTION AND OPERATION

    DOE Patents [OSTI]

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

    1960-03-15

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

  2. Broad beam ion implanter

    DOE Patents [OSTI]

    Leung, K.N.

    1996-10-08

    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. 6 figs.

  3. Broad beam ion implanter

    DOE Patents [OSTI]

    Leung, Ka-Ngo

    1996-01-01

    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.

  4. Accelerators AND Beams

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

    ... to energies far higher than usually found on earth. ... Implanting them very precisely in metal surfaces means ... of a Facility for Rare Isotope Beams (FRIB), an ...

  5. Beam Stability Complaint Form

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

    For New Users For Current Users For Administrators MX Users APS User Portal APS Data Management Practices Find a Beamline Apply for Beam Time ESAF Contacts Calendars User...

  6. Catalyzed D-D stellarator reactor

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Sheffield, John; Spong, Donald A.

    2016-05-12

    The advantages of using the catalyzed deuterium-deuterium (D-D) approach for a fusion reactor—lower and less energetic neutron flux and no need for a tritium breeding blanket—have been evaluated in previous papers, giving examples of both tokamak and stellarator reactors. This paper presents an update for the stellarator example, taking account of more recent empirical transport scaling results and design studies of lower-aspect-ratio stellarators. We use a modified version of the Generic Magnetic Fusion Reactor model to cost a stellarator-type reactor. Recently, this model has been updated to reflect the improved science and technology base and costs in the magnetic fusionmore » program. Furthermore, it is shown that an interesting catalyzed D-D, stellarator power plant might be possible if the following parameters could be achieved: R/ ≈ 4, required improvement factor to ISS04 scaling, FR = 0.9 to 1.15, <β> ≈ 8.0% to 11.5%, Zeff ≈ 1.45 plus a relativistic temperature correction, fraction of fast ions lost ≈ 0.07, Bm ≈ 14 to 16 T, and R ≈ 18 to 24 m.« less

  7. Apparatus for an Inertial Fusion Reactor Inventor Abraham Massry |

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

    Princeton Plasma Physics Lab Apparatus for an Inertial Fusion Reactor Inventor Abraham Massry This invention is comprised of a very large vacuum chamber capable of withstanding a very high neutron flux generated by a fusion-fission reaction at the center. A blanket module on the outside of the vacuum chamber captures the neutrons and converts the energy into heat for further conversion into electrical energy. No.: M-820

  8. Power consumption and byproducts in electron beam and electrical discharge processing of volatile organic compounds

    SciTech Connect (OSTI)

    Penetrante, B.M.; Hsiao, M.C.; Bardsley, J.N.

    1996-02-20

    Among the new methods being investigated for the post-process reduction of volatile organic compounds (VOCs) in atmospheric-pressure air streams are based on non-thermal plasmas. Electron beam, pulsed corona and dielectric-barrier discharge methods are among the more extensively investigated techniques for producing non-thermal plasmas. In order to apply non-thermal plasmas in an industrial scale, it is important to establish the electrical power requirements and byproducts of the process. In this paper the authors present experimental results using a compact electron beam reactor, a pulsed corona and a dielectric-barrier discharge reactor. They have used these reactors to study the removal of a wide variety of VOCs. The effects of background gas composition and gas temperature on the decomposition chemistry have been studied. They present a description of the reactions that control the efficiency of the plasma process. They have found that pulsed corona and other types of electrical discharge reactors are most suitable only for processes requiring O radicals. For VOCs requiring copious amounts of electrons, ions, N atoms or OH radicals, the use of electron beam reactors is generally the best way of minimizing the electrical power consumption. Electron beam processing is remarkably more effective for all of the VOCs tested. For control of VOC emissions from dilute, large volume sources such as paint spray booths, cost analysis shows that the electron beam method is cost-competitive to thermal and catalytic methods that employ heat recovery or hybrid techniques.

  9. METHOD OF MEASURING THE INTEGRATED ENERGY OUTPUT OF A NEUTRONIC CHAIN REACTOR

    DOE Patents [OSTI]

    Sturm, W.J.

    1958-12-01

    A method is presented for measuring the integrated energy output of a reactor conslsting of the steps of successively irradiating calibrated thin foils of an element, such as gold, which is rendered radioactive by exposure to neutron flux for periods of time not greater than one-fifth the mean life of the induced radioactlvity and producing an indication of the radioactivity induced in each foil, each foil belng introduced into the reactor immediately upon removal of its predecessor.

  10. REACTOR GROUT THERMAL PROPERTIES

    SciTech Connect (OSTI)

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

    2011-01-28

    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.

  11. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    SciTech Connect (OSTI)

    P. Delmolino

    2005-05-06

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

  12. Vertical transport and sources in flux models

    SciTech Connect (OSTI)

    Canavan, G.H.

    1997-01-01

    Vertical transport in flux models in examined and shown to reproduce expected limits for densities and fluxes. Disparities with catalog distributions are derived and inverted to find the sources required to rectify them.

  13. Material options for a commercial fusion reactor first wall

    SciTech Connect (OSTI)

    Dabiri, A.E.

    1986-05-01

    A study has been conducted to evaluate the potential of various materials for use as first walls in high-power-density commercial fusion reactors. Operating limits for each material were obtained based on a number of criteria, including maximum allowable structural temperatures, critical heat flux, ultimate tensile strength, and design-allowable stress. The results with water as a coolant indicate that a modified alloy similar to HT-9 may be a suitable candidate for low- and medium-power-density reactor first walls with neutron loads of up to 6 MW/m/sup 2/. A vanadium or copper alloy must be used for high-power-density reactors. The neutron wall load limit for vanadium alloys is about 14 MW/sup 2/, provided a suitable coating material is chosen. The extremely limited data base for radiation effects hinders any quantitative assessment of the limits for copper alloys.

  14. High-flux solar photon processes

    SciTech Connect (OSTI)

    Lorents, D C; Narang, S; Huestis, D C; Mooney, J L; Mill, T; Song, H K; Ventura, S

    1992-06-01

    This study was commissioned by the National Renewable Energy Laboratory (NREL) for the purpose of identifying high-flux photoprocesses that would lead to beneficial national and commercial applications. The specific focus on high-flux photoprocesses is based on the recent development by NREL of solar concentrator technology capable of delivering record flux levels. We examined photolytic and photocatalytic chemical processes as well as photothermal processes in the search for processes where concentrated solar flux would offer a unique advantage. 37 refs.

  15. Laser beam generating apparatus

    DOE Patents [OSTI]

    Warner, Bruce E. (Livermore, CA); Duncan, David B. (Auburn, CA)

    1994-01-01

    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).

  16. Laser beam generating apparatus

    DOE Patents [OSTI]

    Warner, Bruce E. (Livermore, CA); Duncan, David B. (Auburn, CA)

    1993-01-01

    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).

  17. Laser beam generating apparatus

    DOE Patents [OSTI]

    Warner, B.E.; Duncan, D.B.

    1994-02-15

    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.

  18. Laser beam generating apparatus

    DOE Patents [OSTI]

    Warner, B.E.; Duncan, D.B.

    1993-12-28

    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.

  19. Picosecond beam monitor

    DOE Patents [OSTI]

    Schutt, D.W.; Beck, G.O.

    1974-01-01

    The current in the beam of a particle accelerator is monitored with picosecond resolution by causing the beam to impinge upon the center conductor of a coaxial line, generating a pulse of electromagnetic energy in response thereto. This pulse is detected by means such as a sampling oscilloscope. (Official Gazette)

  20. Advanced Reactor Concepts Technical Review Panel Report | Department...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    The concepts included five fast reactors and three thermal reactors. As to reactor coolants, there were three sodium-cooled reactors, two gas-cooled reactors, one light ...

  1. Instrumentation, Monitoring and NDE for New Fast Reactors

    SciTech Connect (OSTI)

    Bond, Leonard J.; Doctor, Steven R.; Bunch, Kyle J.; Good, Morris S.; Waltar, Alan E.

    2007-07-28

    The Global Nuclear Energy Partnership (GNEP) has been proposed as a viable system in which to close the fuel cycle in a manner consistent with markedly expanding the global role of nuclear power while significantly reducing proliferation risks. A key part of this system relies on the development of actinide transmutation, which can only be effectively accomplished in a fast-spectrum reactor. The fundamental physics for fast reactors is well established. However, to achieve higher standards of safety and reliability, operate with longer intervals between outages, and achieve high operating capacity factors, new instrumentation and on-line monitoring capabilities will be required--during both fabrication and operation. Since the Fast Flux Test Facility (FFTF) and Experimental Breeder Reactor – II (EBR-II) reactors were operational in the USA, there have been major advances in instrumentation, not the least being the move to digital systems. Some specific capabilities have been developed outside the USA, but new or at least re-established capabilities will be required. In many cases the only available information is in reports and papers. New and improved sensors and instrumentation will be required. Advanced instrumentation has been developed for high-temperature/high-flux conditions in some cases, but most of the original researchers and manufacturers are retired or no longer in business.

  2. REACTOR AND NOVEL METHOD

    DOE Patents [OSTI]

    Young, G.J.; Ohlinger, L.A.

    1958-06-24

    A nuclear reactor of the type which uses a liquid fuel and a method of controlling such a reactor are described. The reactor is comprised essentially of a tank for containing the liquid fuel such as a slurry of discrete particles of fissionnble material suspended in a heavy water moderator, and a control means in the form of a disc of neutron absorbirg material disposed below the top surface of the slurry and parallel thereto. The diameter of the disc is slightly smaller than the diameter of the tank and the disc is perforated to permit a flow of the slurry therethrough. The function of the disc is to divide the body of slurry into two separate portions, the lower portion being of a critical size to sustain a nuclear chain reaction and the upper portion between the top surface of the slurry and the top surface of the disc being of a non-critical size. The method of operation is to raise the disc in the reactor until the lower portion of the slurry has reached a critical size when it is desired to initiate the reaction, and to lower the disc in the reactor to reduce the size of the lower active portion the slurry to below criticality when it is desired to stop the reaction.

  3. Beam director design report

    SciTech Connect (OSTI)

    Younger, F.C.

    1986-08-01

    A design and fabrication effort for a beam director is documented. The conceptual design provides for the beam to pass first through a bending and focusing system (or ''achromat''), through a second achromat, through an air-to-vacuum interface (the ''beam window''), and finally through the vernier steering system. Following an initial concept study for a beam director, a prototype permanent magnet 30/sup 0/ beam-bending achromat and prototype vernier steering magnet were designed and built. In volume II, copies are included of the funding instruments, requests for quotations, purchase orders, a complete set of as-built drawings, magnetic measurement reports, the concept design report, and the final report on the design and fabrication project. (LEW)

  4. Laser beam alignment system

    DOE Patents [OSTI]

    Kasner, William H.; Racki, Daniel J.; Swenson, Clark E.

    1984-01-01

    A plurality of pivotal reflectors direct a high-power laser beam onto a workpiece, and a rotatable reflector is movable to a position wherein it intercepts the beam and deflects a major portion thereof away from its normal path, the remainder of the beam passing to the pivotal reflectors through an aperture in the rotating reflector. A plurality of targets are movable to positions intercepting the path of light traveling to the pivotal reflectors, and a preliminary adjustment of the latter is made by use of a low-power laser beam reflected from the rotating reflector, after which the same targets are used to make a final adjustment of the pivotal reflectors with the portion of the high-power laser beam passed through the rotating reflector.

  5. Light Water Reactor Sustainability Technical Documents | Department...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    in Light Water Reactors: Life After 60 Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high...

  6. Zero Power Reactor simulation | Argonne National Laboratory

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

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

  7. Apparatus for measuring a flux of neutrons

    DOE Patents [OSTI]

    Stringer, James L.

    1977-01-01

    A flux of neutrons is measured by disposing a detector in the flux and applying electronic correlation techniques to discriminate between the electrical signals generated by the neutron detector and the unwanted interfering electrical signals generated by the incidence of a neutron flux upon the cables connecting the detector to the electronic measuring equipment at a remote location.

  8. Scaling of solid state lasers for satellite power beaming applications

    SciTech Connect (OSTI)

    Friedman, H.W.; Albrecht, G.F.; Beach, R.J.

    1994-01-01

    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.

  9. Electron Lens for Beam-Beam Compensation at LHC

    SciTech Connect (OSTI)

    Valishev, A.; Shiltsev, V.; /Fermilab

    2009-05-01

    Head-on beam-beam effect may become a major performance limitation for the LHC in some of the upgrade scenarios. Given the vast experience gained from the operation of Tevatron electron lenses, a similar device provides significant potential for mitigation of beam-beam effects at the LHC. In this report we present the results of simulation studies of beam-beam compensation and analyze potential application of electron lense at LHC and RHIC.

  10. Methanation assembly using multiple reactors

    DOE Patents [OSTI]

    Jahnke, Fred C.; Parab, Sanjay C.

    2007-07-24

    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.

  11. Breazeale Reactor Modernization Program

    SciTech Connect (OSTI)

    Davison, C. C.

    2003-04-16

    The Penn State Breazeale Nuclear Reactor is the longest operating licensed research reactor in the nation. The facility has played a key role in educating scientists, engineers and in providing facilities and services to researchers in many different disciplines. In order to remain a viable and effective research and educational institution, a multi-phase modernization project was proposed. Phase I was the replacement of the 25-year old reactor control and safety system along with associated wiring and hardware. This phase was fully funded by non-federal funds. Tasks identified in Phases II-V expand upon and complement the work done in Phase I to strategically implement state-of-the-art technologies focusing on identified national needs and priorities of the future.

  12. Reactor for exothermic reactions

    DOE Patents [OSTI]

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

    1993-03-02

    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.

  13. Reactor for exothermic reactions

    DOE Patents [OSTI]

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

    1993-01-01

    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.

  14. Dynamic bed reactor

    DOE Patents [OSTI]

    Stormo, Keith E. (Moscow, ID)

    1996-07-02

    A dynamic bed reactor is disclosed in which a compressible open cell foam matrix is periodically compressed and expanded to move a liquid or fluid through the matrix. In preferred embodiments, the matrix contains an active material such as an enzyme, biological cell, chelating agent, oligonucleotide, adsorbent or other material that acts upon the liquid or fluid passing through the matrix. The active material may be physically immobilized in the matrix, or attached by covalent or ionic bonds. Microbeads, substantially all of which have diameters less than 50 microns, can be used to immobilize the active material in the matrix and further improve reactor efficiency. A particularly preferred matrix is made of open cell polyurethane foam, which adsorbs pollutants such as polychlorophenol or o-nitrophenol. The reactors of the present invention allow unidirectional non-laminar flow through the matrix, and promote intimate exposure of liquid reactants to active agents such as microorganisms immobilized in the matrix.

  15. Nuclear reactor safety device

    DOE Patents [OSTI]

    Hutter, Ernest

    1986-01-01

    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.

  16. A NEUTRONIC REACTOR

    DOE Patents [OSTI]

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

    1959-09-01

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

  17. Thermionic Reactor Design Studies

    SciTech Connect (OSTI)

    Schock, Alfred

    1994-08-01

    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.

  18. STUDY OF ELECTRON -PROTON BEAM-BEAM INTERACTION IN ERHIC

    SciTech Connect (OSTI)

    HAO,Y.; LITVINENKO, V.N.; MONTAG, C.; POZDEYEV, E.; PTITSYN, V.

    2007-06-25

    Beam-beam effects present one of major factors limiting the luminosity of colliders. In the linac-ring option of eRHIC design, an electron beam accelerated in a superconducting energy recovery linac collides with a proton beam circulating in the RHIC ring. There are some features of beam-beam effects, which require careful examination in linac-ring configuration. First, the beam-beam interaction can induce specific head-tail type instability of the proton beam referred to as a ''kink'' instability. Thus, beam stability conditions should be established to avoid proton beam loss. Also, the electron beam transverse disruption by collisions has to be evaluated to ensure beam quality is good enough for the energy recovery pass. In addition, fluctuations of electron beam current and/or electron beam size, as well as transverse offset, can cause proton beam emittance growth. The tolerances for those factors should be determined and possible countermeasures should be developed to mitigate the emittance growth. In this paper, a soft Gaussian strong-strong simulation is used to study all of mentioned beam-beam interaction features and possible techniques to reduce the emittance growth.

  19. Fast quench reactor method

    DOE Patents [OSTI]

    Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.; Berry, Ray A.

    1999-01-01

    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.

  20. NEUTRONIC REACTOR SYSTEM

    DOE Patents [OSTI]

    Daniels, F.

    1957-10-15

    Gas-cooled solid-moderator type reactors wherein the fissionable fuel and moderator materials are each in the form of solid pebbles, or discrete particles, and are substantially homogeneously mixed in the proper proportion and placed within the core of the reactor are described. The shape of these discrete particles must be such that voids are present between them when mixed together. Helium enters the bottom of the core and passes through the voids between the fuel and moderator particles to absorb the heat generated by the chain reaction. The hot helium gas is drawn off the top of the core and may be passed through a heat exchanger to produce steam.