National Library of Energy BETA

Sample records for thermal neutron flux

  1. Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux

    DOE Patents [OSTI]

    Bowman, Charles D.

    1992-01-01

    Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux. High thermal neutron fluxes generated from the action of a high power proton accelerator on a spallation target allows the efficient burn-up of higher actinide nuclear waste by a two-step process. Additionally, rapid burn-up of fission product waste for nuclides having small thermal neutron cross sections, and the practicality of small material inventories while achieving significant throughput derive from employment of such high fluxes. Several nuclear technology problems are addressed including 1. nuclear energy production without a waste stream requiring storage on a geological timescale, 2. the burn-up of defense and commercial nuclear waste, and 3. the production of defense nuclear material. The apparatus includes an accelerator, a target for neutron production surrounded by a blanket region for transmutation, a turbine for electric power production, and a chemical processing facility. In all applications, the accelerator power may be generated internally from fission and the waste produced thereby is transmuted internally so that waste management might not be required beyond the human lifespan.

  2. Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux

    DOE Patents [OSTI]

    Bowman, C.D.

    1992-11-03

    Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux. High thermal neutron fluxes generated from the action of a high power proton accelerator on a spallation target allows the efficient burn-up of higher actinide nuclear waste by a two-step process. Additionally, rapid burn-up of fission product waste for nuclides having small thermal neutron cross sections, and the practicality of small material inventories while achieving significant throughput derive from employment of such high fluxes. Several nuclear technology problems are addressed including 1. nuclear energy production without a waste stream requiring storage on a geological timescale, 2. the burn-up of defense and commercial nuclear waste, and 3. the production of defense nuclear material. The apparatus includes an accelerator, a target for neutron production surrounded by a blanket region for transmutation, a turbine for electric power production, and a chemical processing facility. In all applications, the accelerator power may be generated internally from fission and the waste produced thereby is transmuted internally so that waste management might not be required beyond the human lifespan.

  3. 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)

  4. Thermal neutron detection system

    DOE Patents [OSTI]

    Peurrung, Anthony J. (Richland, WA); Stromswold, David C. (West Richland, WA)

    2000-01-01

    According to the present invention, a system for measuring a thermal neutron emission from a neutron source, has a reflector/moderator proximate the neutron source that reflects and moderates neutrons from the neutron source. The reflector/moderator further directs thermal neutrons toward an unmoderated thermal neutron detector.

  5. Method and apparatus for determining the content and distribution of a thermal neutron absorbing material in an object

    DOE Patents [OSTI]

    Crane, Thomas W.

    1986-01-01

    The disclosure is directed to an apparatus and method for determining the content and distribution of a thermal neutron absorbing material within an object. Neutrons having an energy higher than thermal neutrons are generated and thermalized. The thermal neutrons are detected and counted. The object is placed between the neutron generator and the neutron detector. The reduction in the neutron flux corresponds to the amount of thermal neutron absorbing material in the object. The object is advanced past the neutron generator and neutron detector to obtain neutron flux data for each segment of the object. The object may comprise a space reactor heat pipe and the thermal neutron absorbing material may comprise lithium.

  6. Method and apparatus for determining the content and distribution of a thermal neutron absorbing material in an object

    DOE Patents [OSTI]

    Crane, T.W.

    1983-12-21

    The disclosure is directed to an apparatus and method for determining the content and distribution of a thermal neutron absorbing material within an object. Neutrons having an energy higher than thermal neutrons are generated and thermalized. The thermal neutrons are detected and counted. The object is placed between the neutron generator and the neutron detector. The reduction in the neutron flux corresponds to the amount of thermal neutron absorbing material in the object. The object is advanced past the neutron generator and neutron detector to obtain neutron flux data for each segment of the object. The object may comprise a space reactor heat pipe and the thermal neutron absorbing material may comprise lithium.

  7. Dual neutron flux/temperature measurement sensor

    DOE Patents [OSTI]

    Mihalczo, J.T.; Simpson, M.L.; McElhaney, S.A.

    1994-10-04

    Simultaneous measurement of neutron flux and temperature is provided by a single sensor which includes a phosphor mixture having two principal constituents. The first constituent is a neutron sensitive 6LiF and the second is a rare-earth activated Y203 thermophosphor. The mixture is coated on the end of a fiber optic, while the opposite end of the fiber optic is coupled to a light detector. The detected light scintillations are quantified for neutron flux determination, and the decay is measured for temperature determination. 3 figs.

  8. NEUTRONIC REACTOR HAVING LOCALIZED AREAS OF HIGH THERMAL NEUTRON DENSITIES

    DOE Patents [OSTI]

    Newson, H.W.

    1958-06-01

    A nuclear reactor for the irradiation of materials designed to provide a localized area of high thermal neutron flux density in which the materials to be irradiated are inserted is described. The active portion of the reactor is comprised of a cubicle graphite moderator of about 25 feet in length along each axis which has a plurality of cylindrical channels for accommodatirg elongated tubular-shaped fuel elements. The fuel elements have radial fins for spacing the fuel elements from the channel walls, thereby providing spaces through which a coolant may be passed, and also to serve as a heatconductirg means. Ducts for accommnodating the sample material to be irradiated extend through the moderator material perpendicular to and between parallel rows of fuel channels. The improvement is in the provision of additional fuel element channels spaced midway between 2 rows of the regular fuel channels in the localized area surrounding the duct where the high thermal neutron flux density is desired. The fuel elements normally disposed in the channels directly adjacent the duct are placed in the additional channels, and the channels directly adjacent the duct are plugged with moderator material. This design provides localized areas of high thermal neutron flux density without the necessity of providing additional fuel material.

  9. 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.

  10. High intensity, pulsed thermal neutron source

    DOE Patents [OSTI]

    Carpenter, J.M.

    1973-12-11

    This invention relates to a high intensity, pulsed thermal neutron source comprising a neutron-producing source which emits pulses of fast neutrons, a moderator block adjacent to the last neutron source, a reflector block which encases the fast neutron source and the moderator block and has a thermal neutron exit port extending therethrough from the moderator block, and a neutron energy- dependent decoupling reflector liner covering the interior surfaces of the thermal neutron exit port and surrounding all surfaces of the moderator block except the surface viewed by the thermal neutron exit port. (Official Gazette)

  11. 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.

  12. Thermal conductivity changes upon neutron transmutation of {sup 10}B doped diamond

    SciTech Connect (OSTI)

    Jagannadham, K., E-mail: jag-kasichainula@ncsu.edu [Materials Science and Engineering, North Carolina State University, Raleigh, North Carolina 27695 (United States); Verghese, K. [Nuclear Engineering, North Carolina State University, Raleigh, North Carolina 27695 (United States); Butler, J. E. [Code 6174, Naval research Laboratory, Washington, District of Columbia 20375 (United States)

    2014-08-28

    {sup 10}B doped p-type diamond samples were subjected to neutron transmutation reaction using thermal neutron flux of 0.9 10{sup 13} cm{sup ?2} s{sup ?1} and fast neutron flux of 0.09 10{sup 13} cm{sup ?2} s{sup ?1}. Another sample of epilayer grown on type IIa (110) single crystal diamond substrate was subjected to equal thermal and fast neutron flux of 10{sup 14}?cm{sup ?2} s{sup ?1}. The defects in the diamond samples were previously characterized by different methods. In the present work, thermal conductivity of these diamond samples was determined at room temperature by transient thermoreflectance method. The thermal conductivity change in the samples as a function of neutron fluence is explained by the phonon scattering from the point defects and disordered regions. The thermal conductivity of the diamond samples decreased more rapidly initially and less rapidly for larger neutron fluence. In addition, the thermal conductivity in type IIb diamond decreased less rapidly with thermal neutron fluence compared to the decrease in type IIa diamond subjected to fast neutron fluence. It is concluded that the rate of production of defects during transmutation reaction is slower when thermal neutrons are used. The thermal conductivity of epilayer of diamond subjected to high thermal and fast neutron fluence is associated with the covalent carbon network in the composite structure consisting of disordered carbon and sp{sup 2} bonded nanocrystalline regions.

  13. 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. [Dept. of Nuclear Engineering, Univ. of Tennessee Knoxville, Knoxville, TN 37996-2300 (United States)

    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)

  14. Dosimetry in Thermal Neutron Irradiation Facility at BMRR

    SciTech Connect (OSTI)

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

    2014-05-23

    Radiation dosimetry for Neutron Capture Therapy (NCT) has been performed since 1959 at Thermal Neutron Irradiation Facility (TNIF) of the three-megawatt light-water cooled Brookhaven Medical Research Reactor (BMRR). In the early 1990s when more effective drug carriers were developed for NCT, in which the eye melanoma and brain tumors in rats were irradiated in situ, extensive clinical trials of small animals began using a focused thermal neutron beam. To improve the dosimetry at irradiation facility, a series of innovative designs and major modifications made to enhance the beam intensity and to ease the experimental sampling at BMRR were performed; including (1) in-core fuel addition to increase source strength and balance flux of neutrons towards two ports, (2) out of core moderator remodeling, done by replacing thicker D2O tanks at graphite-shutter interfacial areas, to expedite neutron thermalization, (3) beam shutter upgrade to reduce strayed neutrons and gamma dose, (4) beam collimator redesign to optimize the beam flux versus dose for animal treatment, (5) beam port shielding installation around the shutter opening area (lithium-6 enriched polyester-resin in boxes, attached with polyethylene plates) to reduce prompt gamma and fast neutron doses, (6) sample holder repositioning to optimize angle versus distance for a single organ or whole body irradiation, and (7) holder wall buildup with neutron reflector materials to increase dose and dose rate from scattered thermal neutrons. During the facility upgrade, reactor dosimetry was conducted using thermoluminescent dosimeters TLD for gamma dose estimate, using ion chambers to confirm fast neutron and gamma dose rate, and by the activation of gold-foils with and without cadmium-covers, for fast and thermal neutron flux determination. Based on the combined effect from the size and depth of tumor cells and the location and geometry of dosimeters, the measured flux from cadmium-difference method was 4 - 7 % lower than the statistical mean derived from the Monte-Carlo modeling (5% uncertainty). The dose rate measured by ion chambers was 6 - 10 % lower than the output tallies (7% uncertainty). The detailed dosimetry that was performed at the TNIF for the NCT will be described.

  15. Thermal neutron shield and method of manufacture

    DOE Patents [OSTI]

    Brindza, Paul Daniel; Metzger, Bert Clayton

    2013-05-28

    A thermal neutron shield comprising concrete with a high percentage of the element Boron. The concrete is least 54% Boron by weight which maximizes the effectiveness of the shielding against thermal neutrons. The accompanying method discloses the manufacture of Boron loaded concrete which includes enriching the concrete mixture with varying grit sizes of Boron Carbide.

  16. Thermal neutron shield and method of manufacture

    SciTech Connect (OSTI)

    Metzger, Bert Clayton; Brindza, Paul Daniel

    2014-03-04

    A thermal neutron shield comprising boron shielding panels with a high percentage of the element Boron. The panel is least 46% Boron by weight which maximizes the effectiveness of the shielding against thermal neutrons. The accompanying method discloses the manufacture of boron shielding panels which includes enriching the pre-cursor mixture with varying grit sizes of Boron Carbide.

  17. Transport of thermal neutrons in different forms of liquid hydrogen and the production of intense beams of cold neutrons

    SciTech Connect (OSTI)

    Swaminathan, K.; Tewari, S.P.

    1982-10-01

    From their studies the authors find that the thermal neutron inelastic scattering kernel incorporating the chemical binding energy in liquid hydrogen is able to successfully explain various neutron transport studies such as pulsed neutron and steady-state neutron spectra. For an infinite-sized assembly, D/sub 2/ at 4 K yields a very intense flux of cold and ultracold neutrons. For the practicable finite assembly corresponding to B/sup 2/ = 0.0158 cm/sup -2/, it is found that liquid hydrogen at 11 K gives the most intense beam of cold neutrons.

  18. 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.

  19. Absorption of Thermal Neutrons in Uranium

    DOE R&D Accomplishments [OSTI]

    Creutz, E. C.; Wilson, R. R.; Wigner, E. P.

    1941-09-26

    A knowledge of the absorption processes for neutrons in uranium is important for planning a chain reaction experiment. The absorption of thermal neutrons in uranium and uranium oxide has been studied. Neutrons from the cyclotron were slowed down by passage through a graphite block. A uranium or uranium oxide sphere was placed at various positions in the block. The neutron intensity at different points in the sphere and in the graphite was measured by observing the activity induced in detectors or uranium oxide or manganese. It was found that both the fission activity in the uranium oxide and the activity induced in manganese was affected by non-thermal neutrons. An experimental correction for such effects was made by making measurements with the detectors surrounded by cadmium. After such corrections the results from three methods of procedure with the uranium oxide detectors and from the manganese detectors were consistent to within a few per cent.

  20. Compound Refractive Lenses for Thermal Neutron Applications

    SciTech Connect (OSTI)

    Gary, Charles K.

    2013-11-12

    This project designed and built compound refractive lenses (CRLs) that are able to focus, collimate and image using thermal neutrons. Neutrons are difficult to manipulate compared to visible light or even x rays; however, CRLs can provide a powerful tool for focusing, collimating and imaging neutrons. Previous neutron CRLs were limited to long focal lengths, small fields of view and poor resolution due to the materials available and manufacturing techniques. By demonstrating a fabrication method that can produce accurate, small features, we have already dramatically improved the focal length of thermal neutron CRLs, and the manufacture of Fresnel lens CRLs that greatly increases the collection area, and thus efficiency, of neutron CRLs. Unlike a single lens, a compound lens is a row of N lenslets that combine to produce an N-fold increase in the refraction of neutrons. While CRLs can be made from a variety of materials, we have chosen to mold Teflon lenses. Teflon has excellent neutron refraction, yet can be molded into nearly arbitrary shapes. We designed, fabricated and tested Teflon CRLs for neutrons. We demonstrated imaging at wavelengths as short as 1.26 ? with large fields of view and achieved resolution finer than 250 ?m which is better than has been previously shown. We have also determined designs for Fresnel CRLs that will greatly improve performance.

  1. Operation Greenhouse. Scientific Director's report of atomic weapon tests at Eniwetok, 1951. Annex 1. 5. Neutron measurements. Part 2. External neutron- and gamma flux measurements by sample activation. Section 1

    SciTech Connect (OSTI)

    Biggers, W.A.; Brown, L.J.

    1985-09-01

    The Greenhouse operation consisted of a series of four shots conducted at Eniwetok during the Srping of 1951. The external neutron threshold measurements consisted of the use of good samples to measure integrated thermal neutron fluxes and sulfur, iodine, and zirconium samples to measure fluxes of higher-energy neutrons. The iodine also measured high-energy gamma-ray intensity. Measurements were also made on slow- and fast-neutron intensities as a function of time.

  2. Imaging of Diesel Particulate Filters using a High-Flux Neutron...

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

    Imaging of Diesel Particulate Filters using a High-Flux Neutron Source Imaging of Diesel Particulate Filters using a High-Flux Neutron Source Detailed images of deposits identified ...

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

    SciTech Connect (OSTI)

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

    2013-04-19

    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.

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

    Office of Scientific and Technical Information (OSTI)

    Reactor (Journal Article) | SciTech Connect 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 High Flux Isotope Reactor Neutron scattering at the Oak Ridge National Laboratory dates back to 1945 when Ernest Wollan installed a modified x-ray diffractometer on a beam port of the original graphite reactor. Subsequently, Wollan and Clifford Shull pioneered neutron

  6. 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)

  7. Methods for absorbing neutrons

    DOE Patents [OSTI]

    Guillen, Donna P.; Longhurst, Glen R.; Porter, Douglas L.; Parry, James R.

    2012-07-24

    A conduction cooled neutron absorber may include a metal matrix composite that comprises a metal having a thermal neutron cross-section of at least about 50 barns and a metal having a thermal conductivity of at least about 1 W/cmK. Apparatus for providing a neutron flux having a high fast-to-thermal neutron ratio may include a source of neutrons that produces fast neutrons and thermal neutrons. A neutron absorber positioned adjacent the neutron source absorbs at least some of the thermal neutrons so that a region adjacent the neutron absorber has a fast-to-thermal neutron ratio of at least about 15. A coolant in thermal contact with the neutron absorber removes heat from the neutron absorber.

  8. Thermal Neutron Capture y's (CapGam)

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

    The National Nuclear Data Center (NNDC) presents two tables showing energy and photon intensity with uncertainties of gamma rays as seen in thermal-neutron capture. One table is organized in ascending order of gamma energy, and the second is organized by Z, A of the target. In the energy-ordered table the three strongest transitions are indicated in each case. The nuclide given is the target nucleus in the capture reaction. The gamma energies given are in keV. The gamma intensities given are relative to 100 for the strongest transition. %I? (per 100 n-captures) for the strongest transition is given, where known. All data are taken from the Evaluated Nuclear Structure Data File (ENSDF), a computer file of evaluated nuclear structure data and from the eXperimental Unevaluated Nuclear Data List (XUNDL). (Specialized Interface)

  9. Thermal Neutron Capture y's (CapGam)

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

    The National Nuclear Data Center (NNDC) presents two tables showing energy and photon intensity with uncertainties of gamma rays as seen in thermal-neutron capture.  One table is organized in ascending order of gamma energy, and the second is organized by Z, A of the target. In the energy-ordered table the three strongest transitions are indicated in each case. The nuclide given is the target nucleus in the capture reaction. The gamma energies given are in keV. The gamma intensities given are relative to 100 for the strongest transition. %Iγ (per 100 n-captures) for the strongest transition is given, where known. All data are taken from the Evaluated Nuclear Structure Data File (ENSDF), a computer file of evaluated nuclear structure data and from the eXperimental Unevaluated Nuclear Data List (XUNDL). (Specialized Interface)

  10. ROTATIONAL CORRECTIONS TO NEUTRON-STAR RADIUS MEASUREMENTS FROM THERMAL SPECTRA

    SciTech Connect (OSTI)

    Baubck, Michi; zel, Feryal; Psaltis, Dimitrios; Morsink, Sharon M.

    2015-01-20

    We calculate the rotational broadening in the observed thermal spectra of neutron stars spinning at moderate rates in the Hartle-Thorne approximation. These calculations accurately account for the effects of the second-order Doppler boosts as well as for the oblate shapes and the quadrupole moments of the neutron stars. We find that fitting the spectra and inferring the bolometric fluxes under the assumption that a star is not rotating causes an underestimate of the inferred fluxes and, thus, radii. The correction depends on the stellar spin, mass, radius, and the observer's inclination. For a 10km, 1.4 M {sub ?} neutron star spinning at 600Hz, the rotational correction to the flux is ?1%-4%, while for a 15km neutron star with the same spin period, the correction ranges from 2% for pole-on sources to 12% for edge-on sources. We calculate the inclination-averaged corrections to inferred radii as a function of the neutron-star radius and mass and provide an empirical formula for the corrections. For realistic neutron-star parameters (1.4 M {sub ?}, 12km, 600Hz), the stellar radius is on the order of 4% larger than the radius inferred under the assumption that the star is not spinning.

  11. Theoretical analysis of integral neutron transport equation using collision probability method with quadratic flux approach

    SciTech Connect (OSTI)

    Shafii, Mohammad Ali Meidianti, Rahma Wildian, Fitriyani, Dian; Tongkukut, Seni H. J.; Arkundato, Artoto

    2014-09-30

    Theoretical analysis of integral neutron transport equation using collision probability (CP) method with quadratic flux approach has been carried out. In general, the solution of the neutron transport using the CP method is performed with the flat flux approach. In this research, the CP method is implemented in the cylindrical nuclear fuel cell with the spatial of mesh being conducted into non flat flux approach. It means that the neutron flux at any point in the nuclear fuel cell are considered different each other followed the distribution pattern of quadratic flux. The result is presented here in the form of quadratic flux that is better understanding of the real condition in the cell calculation and as a starting point to be applied in computational calculation.

  12. Neutron diffraction study and anomalous negative thermal expansion in

    Office of Scientific and Technical Information (OSTI)

    non-superconducting PrFe1-xRuxAsO (Journal Article) | SciTech Connect Neutron diffraction study and anomalous negative thermal expansion in non-superconducting PrFe1-xRuxAsO Citation Details In-Document Search Title: Neutron diffraction study and anomalous negative thermal expansion in non-superconducting PrFe1-xRuxAsO Neutron powder diraction has been used to investigate the structural and magnetic behavior of the isoelectronically doped Fe pnictide material PrFe1-xRuxAsO. Substitution of

  13. SHARP Neutronics Expanded | Department of Energy

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

    Neutronics Expanded SHARP Neutronics Expanded January 29, 2013 - 1:28pm Addthis Fully heterogeneous predictions of thermal neutron flux in a hypothetical metal-oxide-fueled PWR Fully heterogeneous predictions of thermal neutron flux in a hypothetical metal-oxide-fueled PWR SHARP neutronics Module Development The SHARP neutronics module, PROTEUS, includes neutron and gamma transport solvers and cross-section processing tools as well as the capability for depletion and fuel cycle analysis. The

  14. EFFECT OF FAST NEUTRON IRRADIATION ON SINTERED ALUMINA AND MAGNESIA...

    Office of Scientific and Technical Information (OSTI)

    IRRADIATION; LATTICES; MAGNESIUM OXIDES; MONOCRYSTALS; NEUTRON FLUX; RADIATION DOSES; RADIATION EFFECTS; SINTERED MATERIALS; TEMPERATURE; THERMAL CONDUCTIVITY ALUMINUM OXIDES

  15. 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.

  16. Lithium-containing scintillators for thermal neutron, fast neutron, and gamma detection

    DOE Patents [OSTI]

    Zaitseva, Natalia P.; Carman, M. Leslie; Faust, Michelle A.

    2016-03-01

    In one embodiment, a scintillator includes a scintillator material; a primary fluor, and a Li-containing compound, where the Li-containing compound is soluble in the primary fluor, and where the scintillator exhibits an optical response signature for thermal neutrons that is different than an optical response signature for fast neutrons and gamma rays.

  17. 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.

  18. 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.

  19. Three-dimensional boron particle loaded thermal neutron detector

    DOE Patents [OSTI]

    Nikolic, Rebecca J.; Conway, Adam M.; Graff, Robert T.; Kuntz, Joshua D.; Reinhardt, Catherine; Voss, Lars F.; Cheung, Chin Li; Heineck, Daniel

    2014-09-09

    Three-dimensional boron particle loaded thermal neutron detectors utilize neutron sensitive conversion materials in the form of nano-powders and micro-sized particles, as opposed to thin films, suspensions, paraffin, etc. More specifically, methods to infiltrate, intersperse and embed the neutron nano-powders to form two-dimensional and/or three-dimensional charge sensitive platforms are specified. The use of nano-powders enables conformal contact with the entire charge-collecting structure regardless of its shape or configuration.

  20. Defect annealing and thermal desorption of deuterium in low dose HFIR neutron-irradiated tungsten

    SciTech Connect (OSTI)

    Masashi Shimada; M. Hara; T. Otsuka; Y. Oya; Y. Hatano

    2014-05-01

    Accurately estimating tritium retention in plasma facing components (PFCs) and minimizing its uncertainty are key safety issues for licensing future fusion power reactors. D-T fusion reactions produce 14.1 MeV neutrons that activate PFCs and create radiation defects throughout the bulk of the material of these components. Recent studies show that tritium migrates and is trapped in bulk (>> 10 m) tungsten beyond the detection range of nuclear reaction analysis technique [1-2], and thermal desorption spectroscopy (TDS) technique becomes the only established diagnostic that can reveal hydrogen isotope behavior in in bulk (>> 10 m) tungsten. Radiation damage and its recovery mechanisms in neutron-irradiated tungsten are still poorly understood, and neutron-irradiation data of tungsten is very limited. In this paper, systematic investigations with repeated plasma exposures and thermal desorption are performed to study defect annealing and thermal desorption of deuterium in low dose neutron-irradiated tungsten. Three tungsten samples (99.99 at. % purity from A.L.M.T. Co., Japan) irradiated at High Flux Isotope Reactor at Oak Ridge National Laboratory were exposed to high flux (ion flux of (0.5-1.0)x1022 m-2s-1 and ion fluence of 1x1026 m-2) deuterium plasma at three different temperatures (100, 200, and 500 C) in Tritium Plasma Experiment at Idaho National Laboratory. Subsequently, thermal desorption spectroscopy (TDS) was performed with a ramp rate of 10 C/min up to 900 C, and the samples were annealed at 900 C for 0.5 hour. These procedures were repeated three (for 100 and 200 C samples) and four (for 500 C sample) times to uncover damage recovery mechanisms and its effects on deuterium behavior. The results show that deuterium retention decreases approximately 90, 75, and 66 % for 100, 200, and 500 C, respectively after each annealing. When subjected to the same TDS recipe, the desorption temperature shifts from 800 C to 600 C after 1st annealing for the sample exposed to TPE at 500 C. Tritium Migration Analysis Program (TMAP) analysis reveals that the detrapping energy decreases from 1.8 eV to 1.4 eV, indicating the changes in trapping mechanisms. This paper also summarizes deuterium behavior studies in HFIR neutron-irradiated tungsten under US-Japan TITAN program.

  1. Ortho- and para-hydrogen in neutron thermalization

    SciTech Connect (OSTI)

    Daemen, L. L.; Brun, T. O.

    1998-01-01

    The large difference in neutron scattering cross-section at low neutron energies between ortho- and para-hydrogen was recognized early on. In view of this difference (more than an order of magnitude), one might legitimately ask whether the ortho/para ratio has a significant effect on the neutron thermalization properties of a cold hydrogen moderator. Several experiments performed in the 60`s and early 70`s with a variety of source and (liquid hydrogen) moderator configurations attempted to investigate this. The results tend to show that the ortho/para ratio does indeed have an effect on the energy spectrum of the neutron beam produced. Unfortunately, the results are not always consistent with each other and much unknown territory remains to be explored. The problem has been approached from a computational standpoint, but these isolated efforts are far from having examined the ortho/para-hydrogen problem in neutron moderation in all its complexity. Because of space limitations, the authors cannot cover, even briefly, all the aspects of the ortho/para question here. This paper will summarize experiments meant to investigate the effect of the ortho/para ratio on the neutron energy spectrum produced by liquid hydrogen moderators.

  2. Method for manufacturing solid-state thermal neutron detectors with simultaneous high thermal neutron detection efficiency (>50%) and neutron to gamma discrimination (>1.0E4)

    DOE Patents [OSTI]

    Nikolic, Rebecca J.; Conway, Adam M.; Heineck, Daniel; Voss, Lars F.; Wang, Tzu Fang; Shao, Qinghui

    2013-10-15

    Methods for manufacturing solid-state thermal neutron detectors with simultaneous high thermal neutron detection efficiency (>50%) and neutron to gamma discrimination (>10.sup.4) are provided. A structure is provided that includes a p+ region on a first side of an intrinsic region and an n+ region on a second side of the intrinsic region. The thickness of the intrinsic region is minimized to achieve a desired gamma discrimination factor of at least 1.0E+04. Material is removed from one of the p+ region or the n+ region and into the intrinsic layer to produce pillars with open space between each pillar. The open space is filed with a neutron sensitive material. An electrode is placed in contact with the pillars and another electrode is placed in contact with the side that is opposite of the intrinsic layer with respect to the first electrode.

  3. Minimum activation martensitic alloys for surface disposal after exposure to neutron flux

    DOE Patents [OSTI]

    Lechtenberg, Thomas

    1985-01-01

    Steel alloys for long-term exposure to neutron flux have a martensitic microstructure and contain chromium, carbon, tungsten, vanadium and preferably titanium. Activation of the steel is held to within acceptable limits for eventual surface disposal by stringently controlling the impurity levels of Ni, Mo, Cu, N, Co, Nb, Al and Mn.

  4. Thermal Neutron Capture for Nuclei A = 3 - 20

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

    Thermal Neutron Capture Evaluated Data for Nuclei A = 3 - 20 Go to the Text Only below if you prefer to view the nuclides in a text list. 19Ne 20Ne 18F 19F 20F 15O 16O 17O 18O 19O 13N 14N 15N 16N 11C 12C 13C 14C 15C 8B 9B 10B 11B 12B 7Be 8Be 9Be 10Be 11Be 6Li 7Li 8Li 9Li 4He p 2H 3H n Color-code Key: Experiments have been performed Others Note: Comments, and corrections are welcome. Please email us. List of available thermal neutron capture data for A = 2 - 20 nuclides: Hydrogen: 2H, 3H Helium:

  5. Accurate Development of Thermal Neutron Scattering Cross Section Libraries

    SciTech Connect (OSTI)

    Hawari, Ayman; Dunn, Michael

    2014-06-10

    The objective of this project is to develop a holistic (fundamental and accurate) approach for generating thermal neutron scattering cross section libraries for a collection of important enutron moderators and reflectors. The primary components of this approach are the physcial accuracy and completeness of the generated data libraries. Consequently, for the first time, thermal neutron scattering cross section data libraries will be generated that are based on accurate theoretical models, that are carefully benchmarked against experimental and computational data, and that contain complete covariance information that can be used in propagating the data uncertainties through the various components of the nuclear design and execution process. To achieve this objective, computational and experimental investigations will be performed on a carefully selected subset of materials that play a key role in all stages of the nuclear fuel cycle.

  6. Measurement of Neutron and Muon Fluxes 100~m Underground with the SciBath Detector

    SciTech Connect (OSTI)

    Garrison, Lance

    2014-01-01

    The SciBath detector is an 80 liter liquid scintillator detector read out by a three dimensional grid of 768 wavelength-shifting fibers. Initially conceived as a fine-grained charged particle detector for neutrino studies that could image charged particle tracks in all directions, it is also sensitive to fast neutrons (15-200 MeV). In fall of 2011 the apparatus performed a three month run to measure cosmic-induced muons and neutrons 100~meters underground in the FNAL MINOS near-detector area. Data from this run has been analyzed and resulted in measurements of the cosmic muon flux as \

  7. Wide dynamic range neutron flux monitor having fast time response for the Large Helical Device

    SciTech Connect (OSTI)

    Isobe, M. Takeiri, Y.; Ogawa, K.; Miyake, H.; Hayashi, H.; Kobuchi, T.; Nakano, Y.; Watanabe, K.; Uritani, A.; Misawa, T.; Nishitani, T.; Tomitaka, M.; Kumagai, T.; Mashiyama, Y.; Ito, D.; Kono, S.; Yamauchi, M.

    2014-11-15

    A fast time response, wide dynamic range neutron flux monitor has been developed toward the LHD deuterium operation by using leading-edge signal processing technologies providing maximum counting rate up to ∼5 × 10{sup 9} counts/s. Because a maximum total neutron emission rate over 1 × 10{sup 16} n/s is predicted in neutral beam-heated LHD plasmas, fast response and wide dynamic range capabilities of the system are essential. Preliminary tests have demonstrated successful performance as a wide dynamic range monitor along the design.

  8. Coupled Monte Carlo neutronics and thermal hydraulics for power reactors

    SciTech Connect (OSTI)

    Bernnat, W.; Buck, M.; Mattes, M.; Zwermann, W.; Pasichnyk, I.; Velkov, K.

    2012-07-01

    The availability of high performance computing resources enables more and more the use of detailed Monte Carlo models even for full core power reactors. The detailed structure of the core can be described by lattices, modeled by so-called repeated structures e.g. in Monte Carlo codes such as MCNP5 or MCNPX. For cores with mainly uniform material compositions, fuel and moderator temperatures, there is no problem in constructing core models. However, when the material composition and the temperatures vary strongly a huge number of different material cells must be described which complicate the input and in many cases exceed code or memory limits. The second problem arises with the preparation of corresponding temperature dependent cross sections and thermal scattering laws. Only if these problems can be solved, a realistic coupling of Monte Carlo neutronics with an appropriate thermal-hydraulics model is possible. In this paper a method for the treatment of detailed material and temperature distributions in MCNP5 is described based on user-specified internal functions which assign distinct elements of the core cells to material specifications (e.g. water density) and temperatures from a thermal-hydraulics code. The core grid itself can be described with a uniform material specification. The temperature dependency of cross sections and thermal neutron scattering laws is taken into account by interpolation, requiring only a limited number of data sets generated for different temperatures. Applications will be shown for the stationary part of the Purdue PWR benchmark using ATHLET for thermal- hydraulics and for a generic Modular High Temperature reactor using THERMIX for thermal- hydraulics. (authors)

  9. Thermal Neutron Detectors with Discrete Anode Pad Readout

    SciTech Connect (OSTI)

    Yu,B.; Schaknowski, N.A., Smith, G.C., DeGeronimo, G., Vernon, E.O.

    2008-10-19

    A new two-dimensional thermal neutron detector concept that is capable of very high rates is being developed. It is based on neutron conversion in {sup 3}He in an ionization chamber (unity gas gain) that uses only a cathode and anode plane; there is no additional electrode such as a Frisch grid. The cathode is simply the entrance window, and the anode plane is composed of discrete pads, each with their own readout electronics implemented via application specific integrated circuits. The aim is to provide a new generation of detectors with key characteristics that are superior to existing techniques, such as higher count rate capability, better stability, lower sensitivity to background radiation, and more flexible geometries. Such capabilities will improve the performance of neutron scattering instruments at major neutron user facilities. In this paper, we report on progress with the development of a prototype device that has 48 x 48 anode pads and a sensitive area of 24cm x 24cm.

  10. Neutron activation analysis system

    DOE Patents [OSTI]

    Taylor, M.C.; Rhodes, J.R.

    1973-12-25

    A neutron activation analysis system for monitoring a generally fluid media, such as slurries, solutions, and fluidized powders, including two separate conduit loops for circulating fluid samples within the range of radiation sources and detectors is described. Associated with the first loop is a neutron source that emits s high flux of slow and thermal neutrons. The second loop employs a fast neutron source, the flux from which is substantially free of thermal neutrons. Adjacent to both loops are gamma counters for spectrographic determination of the fluid constituents. Other gsmma sources and detectors are arranged across a portion of each loop for deterMining the fluid density. (Official Gazette)

  11. Fiber optic thermal/fast neutron and gamma ray scintillation detector

    DOE Patents [OSTI]

    Neal, John S.; Mihalczo, John T.

    2006-11-28

    A detector system that combines a .sup.6Li loaded glass fiber scintillation thermal neutron detector with a fast scintillation detector in a single layered structure. Detection of thermal and fast neutrons and ionizing electromagnetic radiation is achieved in the unified detector structure. The fast scintillator replaces the polyethelene moderator layer adjacent the .sup.6Li loaded glass fiber panel of the neutron detector and acts as the moderator for the glass fibers. Fast neutrons, x-rays and gamma rays are detected in the fast scintillator. Thermal neutrons, x-rays and gamma rays are detected in the glass fiber scintillator.

  12. Double helix boron-10 powder thermal neutron detector

    DOE Patents [OSTI]

    Wang, Zhehui; Morris, Christopher L.; Bacon, Jeffrey D.

    2015-06-02

    A double-helix Boron-10 powder detector having intrinsic thermal neutron detection efficiency comparable to 36'' long, 2-in diameter, 2-bar Helium-3 detectors, and which can be used to replace such detectors for use in portal monitoring, is described. An embodiment of the detector includes a metallic plate coated with Boron-10 powder for generating alpha and Lithium-7 particles responsive to neutrons impinging thereon supported by insulators affixed to at least two opposing edges; a grounded first wire wound in a helical manner around two opposing insulators; and a second wire having a smaller diameter than that of the first wire, wound in a helical manner around the same insulators and spaced apart from the first wire, the second wire being positively biased. A gas, disposed within a gas-tight container enclosing the plate, insulators and wires, and capable of stopping alpha and Lithium-7 particles and generating electrons produces a signal on the second wire which is detected and subsequently related to the number of neutrons impinging on the plate.

  13. 3D neutronic/thermal-hydraulic coupled analysis of MYRRHA

    SciTech Connect (OSTI)

    Vazquez, M.; Martin-Fuertes, F.

    2012-07-01

    The current tendency in multiphysics calculations applied to reactor physics is the use of already validated computer codes, coupled by means of an iterative approach. In this paper such an approach is explained concerning neutronics and thermal-hydraulics coupled analysis with MCNPX and COBRA-IV codes using a driver program and file exchange between codes. MCNPX provides the neutronic analysis of heterogeneous nuclear systems, both in critical and subcritical states, while COBRA-IV is a subchannel code that can be used for rod bundles or core thermal-hydraulics analysis. In our model, the MCNP temperature dependence of nuclear data is handled via pseudo-material approach, mixing pre-generated cross section data set to obtain the material with the desired cross section temperature. On the other hand, COBRA-IV has been updated to allow for the simulation of liquid metal cooled reactors. The coupled computational tool can be applied to any geometry and coolant, as it is the case of single fuel assembly, at pin-by-pin level, or full core simulation with the average pin of each fuel-assembly. The coupling tool has been applied to the critical core layout of the SCK-CEN MYRRHA concept, an experimental LBE cooled fast reactor presently in engineering design stage. (authors)

  14. Numerical studies of the flux-to-current ratio method in the KIPT neutron source facility

    SciTech Connect (OSTI)

    Cao, Y.; Gohar, Y.; Zhong, Z.

    2013-07-01

    The reactivity of a subcritical assembly has to be monitored continuously in order to assure its safe operation. In this paper, the flux-to-current ratio method has been studied as an approach to provide the on-line reactivity measurement of the subcritical system. Monte Carlo numerical simulations have been performed using the KIPT neutron source facility model. It is found that the reactivity obtained from the flux-to-current ratio method is sensitive to the detector position in the subcritical assembly. However, if multiple detectors are located about 12 cm above the graphite reflector and 54 cm radially, the technique is shown to be very accurate in determining the k{sub eff} this facility in the range of 0.75 to 0.975. (authors)

  15. Validation of the fast neutron spectrum in the coupled fast-thermal system HERBE

    SciTech Connect (OSTI)

    Avdic, S.; Pesic, M.; Marinkovic, P.

    1995-12-31

    Methods applied in the calculation and interpretation of the measurements of the fast neutron spectrum in the NERBE coupled fast-thermal system are validated in this paper. When advantages and disadvantages of a He-filled semi-conductor-sandwich detector are compared to other neutron detectors, the former is found more appropriate. The neutron detection is based on the reaction {sup 3}He(n,p)T + 0.764 MeV and simultaneous detection of the reaction products in the silicon diodes. The pulses from the diodes are amplified and shaped in separate {open_quotes}energy{close_quotes} channels and summed to produce a single pulse with height proportional to the energy of the incident neutron plus the Q value of the reaction. A well-known measuring system of the He neutron spectrometer is used for the HERBE fast neutron spectrum measurement and calibration in a thermal neutron field.

  16. Fiber optic thermal/fast neutron and gamma ray scintillation detector

    DOE Patents [OSTI]

    Neal, John S. (Knoxville, TN); Mihalczo, John T (Oak Ridge, TN)

    2007-10-30

    A system for detecting fissile and fissionable material originating external to the system includes: a .sup.6Li loaded glass fiber scintillator for detecting thermal neutrons, x-rays and gamma rays; a fast scintillator for detecting fast neutrons, x-rays and gamma rays, the fast scintillator conjoined with the glass fiber scintillator such that the fast scintillator moderates fast neutrons prior to their detection as thermal neutrons by the glass fiber scintillator; and a coincidence detection system for processing the time distributions of arriving signals from the scintillators.

  17. IMPROVED COMPUTATIONAL CHARACTERIZATION OF THE THERMAL NEUTRON SOURCE FOR NEUTRON CAPTURE THERAPY RESEARCH AT THE UNIVERSITY OF MISSOURI

    SciTech Connect (OSTI)

    Stuart R. Slattery; David W. Nigg; John D. Brockman; M. Frederick Hawthorne

    2010-05-01

    Parameter studies, design calculations and initial neutronic performance measurements have been completed for a new thermal neutron beamline to be used for neutron capture therapy cell and small-animal radiobiology studies at the University of Missouri Research Reactor. The beamline features the use of single-crystal silicon and bismuth sections for neutron filtering and for reduction of incident gamma radiation. The computational models used for the final beam design and performance evaluation are based on coupled discrete-ordinates and Monte Carlo techniques that permit detailed modeling of the neutron transmission properties of the filtering crystals with very few approximations. This is essential for detailed dosimetric studies required for the anticipated research program.

  18. Direct thermal-neutron capture by {sup 30}Si (Journal Article) | SciTech

    Office of Scientific and Technical Information (OSTI)

    Connect Direct thermal-neutron capture by {sup 30}Si Citation Details In-Document Search Title: Direct thermal-neutron capture by {sup 30}Si Five primary electric-dipole transitions in {sup 31}Si account for the bulk ({approx}98%) of the total thermal-neutron capture cross section (107{+-}3 mb) of {sup 30}Si. We have recalculated the partial cross sections for these transitions using direct-capture theory and reliable spectroscopic factors for the (d,p) reaction, which have become available

  19. METHOD OF TESTING THERMAL NEUTRON FISSIONABLE MATERIAL FOR PURITY

    DOE Patents [OSTI]

    Fermi, E.; Anderson, H.L.

    1961-01-24

    A process is given for determining the neutronic purity of fissionable material by the so-called shotgun test. The effect of a standard neutron absorber of known characteristics and amounts on a neutronic field also of known characteristics is measured and compared with the effect which the impurities derived from a known quantity of fissionable material has on the same neutronic field. The two readings are then made the basis of calculation from which the amount of impurities can be computed.

  20. Absorbed dose rates in tissue from prompt gamma emissions from near-thermal neutron absorption

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

    Schwahn, Scott O.

    2015-10-01

    Prompt gamma emission data from the International Atomic Energy Agency s Prompt Gamma-ray Neutron Activation Analysis database are analyzed to determine the absorbed dose rates in tissue to be expected when natural elements are exposed in a near-thermal neutron environment.

  1. Analysis of the OECD/NRC BWR Turbine Trip Transient Benchmark with the Coupled Thermal-Hydraulics and Neutronics Code TRAC-M/PARCS

    SciTech Connect (OSTI)

    Lee, Deokjung; Downar, Thomas J.; Ulses, Anthony; Akdeniz, Bedirhan; Ivanov, Kostadin N.

    2004-10-15

    An analysis of the Peach Bottom Unit 2 Turbine Trip 2 (TT2) experiment has been performed using the U.S. Nuclear Regulatory Commission coupled thermal-hydraulics and neutronics code TRAC-M/PARCS. The objective of the analysis was to assess the performance of TRAC-M/PARCS on a BWR transient with significance in two-phase flow and spatial variations of the neutron flux. TRAC-M/PARCS results are found to be in good agreement with measured plant data for both steady-state and transient phases of the benchmark. Additional analyses of four fictitious extreme scenarios are performed to provide a basis for code-to-code comparisons and comprehensive testing of the thermal-hydraulics/neutronics coupling. The obtained results of sensitivity studies on the effect of direct moderator heating on transient simulation indicate the importance of this modeling aspect.

  2. Experimental investigation of neutron emissions during thermal cycling of TiD{sub x} (x {approx} 2.00)

    SciTech Connect (OSTI)

    Fernandez, J.F.; Cuevas, F.; Alguero, M.; Sanchez, C.

    1997-03-01

    The production of neutrons from D + D reactions in thermally cycled titanium deuteride (TiD{sub x}) (x {approx} 2) is investigated in depth. Special attention is given to cubic-tetragonal ({delta} - {epsilon}) phase transition that TiD{sub x} experiences near room temperature as a possible triggering mechanism of `cold nuclear fusion reactions.` The TiD{sub x} (x {approx} 2.00) samples, possessing well-known properties about the {delta} - {epsilon} transition, are cycled at temperatures (from -60 to 60{degree} C) where the phase transition takes place. The cold fusion signature is investigated by measuring the neutron flux of the sample during the experiments. No significant neutron signal above the background level is found during thermal cycling of the TiD{sub x} samples. It is concluded that in the samples investigated, no correlation exists between the {delta} - {epsilon} transition and the trigger of the D + D reactions. Background deviations give an upper limit of the rate of the D + D {yields} {sup 3}He + n reaction of {lambda} < 10{sup -23} fusion/p-d.s. 32 refs., 8 figs., 4 tabs.

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

    Office of Scientific and Technical Information (OSTI)

    Neutron scattering at the Oak Ridge National Laboratory ... diffraction and laid the foundation for an active neutron ... Sponsoring Org: SC USDOE - Office of Science (SC) Country of ...

  4. Optimizing Neutron Thermal Scattering Effects in very High Temperature Reactors. Final Report

    SciTech Connect (OSTI)

    Hawari, Ayman; Ougouag, Abderrafi

    2014-07-08

    This project aims to develop a holistic understanding of the phenomenon of neutron thermalization in the VHTR. Neutron thermalization is dependent on the type and structure of the moderating material. The fact that the moderator (and reflector) in the VHTR is a solid material will introduce new and interesting considerations that do not apply in other (e.g. light water) reactors. The moderator structure is expected to undergo radiation induced changes as the irradiation (or burnup) history progresses. In this case, the induced changes in structure will have a direct impact on many properties including the neutronic behavior. This can be easily anticipated if one recognizes the dependence of neutron thermalization on the scattering law of the moderator. For the pebble bed reactor, it is anticipated that the moderating behavior can be tailored, e.g. using moderators that consist of composite materials, which could allow improved optimization of the moderator-to-fuel ratio.

  5. Thermal flux limited electron Kapitza conductance in copper-niobium multilayers

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

    Cheaito, Ramez; Hattar, Khalid Mikhiel; Gaskins, John T.; Yadav, Ajay K.; Duda, John C.; Beechem, III, Thomas Edwin; Ihlefeld, Jon; Piekos, Edward S.; Baldwin, Jon K.; Misra, Amit; et al

    2015-03-05

    The interplay between the contributions of electron thermal flux and interface scattering to the Kapitza conductance across metal-metal interfaces through measurements of thermal conductivity of copper-niobium multilayers was studied. Thermal conductivities of copper-niobium multilayer films of period thicknesses ranging from 5.4 to 96.2 nm and sample thicknesses ranging from 962 to 2677 nm are measured by time-domain thermoreflectance over a range of temperatures from 78 to 500 K. The Kapitza conductances between the Cu and Nb interfaces in multilayer films are determined from the thermal conductivities using a series resistor model and are in good agreement with the electron diffusemore » mismatch model. The results for the thermal boundary conductance between Cu and Nb are compared to literature values for the thermal boundary conductance across Al-Cu and Pd-Ir interfaces, and demonstrate that the interface conductance in metallic systems is dictated by the temperature derivative of the electron energy flux in the metallic layers, rather than electron mean free path or scattering processes at the interface.« less

  6. Thermal flux limited electron Kapitza conductance in copper-niobium multilayers

    SciTech Connect (OSTI)

    Cheaito, Ramez; Hattar, Khalid Mikhiel; Gaskins, John T.; Yadav, Ajay K.; Duda, John C.; Beechem, III, Thomas Edwin; Ihlefeld, Jon; Piekos, Edward S.; Baldwin, Jon K.; Misra, Amit; Hopkins, Patrick E.

    2015-03-05

    The interplay between the contributions of electron thermal flux and interface scattering to the Kapitza conductance across metal-metal interfaces through measurements of thermal conductivity of copper-niobium multilayers was studied. Thermal conductivities of copper-niobium multilayer films of period thicknesses ranging from 5.4 to 96.2 nm and sample thicknesses ranging from 962 to 2677 nm are measured by time-domain thermoreflectance over a range of temperatures from 78 to 500 K. The Kapitza conductances between the Cu and Nb interfaces in multilayer films are determined from the thermal conductivities using a series resistor model and are in good agreement with the electron diffuse mismatch model. The results for the thermal boundary conductance between Cu and Nb are compared to literature values for the thermal boundary conductance across Al-Cu and Pd-Ir interfaces, and demonstrate that the interface conductance in metallic systems is dictated by the temperature derivative of the electron energy flux in the metallic layers, rather than electron mean free path or scattering processes at the interface.

  7. Imaging of Diesel Particulate Filters using a High-Flux Neutron Source

    Broader source: Energy.gov [DOE]

    Detailed images of deposits identified inside automotive DPFs using neutrons show how the deposits of soot, ash, and washcoat occurs within the filter.

  8. Subsurface Temperature, Moisture, Thermal Conductivity and Heat Flux, Barrow, Area A, B, C, D

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

    Cable, William; Romanovsky, Vladimir

    Subsurface temperature data are being collected along a transect from the center of the polygon through the trough (and to the center of the adjacent polygon for Area D). Each transect has five 1.5m vertical array thermistor probes with 16 thermistors each. This dataset also includes soil pits that have been instrumented for temperature, water content, thermal conductivity, and heat flux at the permafrost table. Area C has a shallow borehole of 2.5 meters depth is instrumented in the center of the polygon.

  9. Subsurface Temperature, Moisture, Thermal Conductivity and Heat Flux, Barrow, Area A, B, C, D

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

    Cable, William; Romanovsky, Vladimir

    2014-03-31

    Subsurface temperature data are being collected along a transect from the center of the polygon through the trough (and to the center of the adjacent polygon for Area D). Each transect has five 1.5m vertical array thermistor probes with 16 thermistors each. This dataset also includes soil pits that have been instrumented for temperature, water content, thermal conductivity, and heat flux at the permafrost table. Area C has a shallow borehole of 2.5 meters depth is instrumented in the center of the polygon.

  10. Thermal neutron detection using a silicon pad detector and {sup 6}LiF removable converters

    SciTech Connect (OSTI)

    Barbagallo, Massimo; Cosentino, Luigi; Marchetta, Carmelo; Pappalardo, Alfio; Scire, Carlotta; Scire, Sergio; Schillaci, Maria; Vecchio, Gianfranco; Finocchiaro, Paolo; Forcina, Vittorio; Peerani, Paolo; Vaccaro, Stefano

    2013-03-15

    A semiconductor detector coupled with a neutron converter is a good candidate for neutron detection, especially for its compactness and reliability if compared with other devices, such as {sup 3}He tubes, even though its intrinsic efficiency is rather lower. In this paper we show a neutron detector design consisting of a 3 cm Multiplication-Sign 3 cm silicon pad detector coupled with one or two external {sup 6}LiF layers, enriched in {sup 6}Li at 95%, placed in contact with the Si active surfaces. This prototype, first characterized and tested at INFN Laboratori Nazionali del Sud and then at JRC Ispra, was successfully shown to detect thermal neutrons with the expected efficiency and an outstanding gamma rejection capability.

  11. Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

    DOE Patents [OSTI]

    Schreiber, R.B.; Fero, A.H.; Sejvar, J.

    1997-12-16

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor. 8 figs.

  12. Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

    DOE Patents [OSTI]

    Schreiber, Roger B.; Fero, Arnold H.; Sejvar, James

    1997-01-01

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor.

  13. ROTATING SOLAR JETS IN SIMULATIONS OF FLUX EMERGENCE WITH THERMAL CONDUCTION

    SciTech Connect (OSTI)

    Fang, Fang; Fan, Yuhong; McIntosh, Scott W.

    2014-07-01

    We study the formation of coronal jets through numerical simulation of the emergence of a twisted magnetic flux rope into a pre-existing open magnetic field. Reconnection inside the emerging flux rope in addition to that between the emerging and pre-existing fields give rise to the violent eruption studied. The simulated event closely resembles the coronal jets ubiquitously observed by the X-Ray Telescope on board Hinode and demonstrates that heated plasma is driven into the extended atmosphere above. Thermal conduction implemented in the model allows us to qualitatively compare simulated and observed emission from such events. We find that untwisting field lines after the reconnection drive spinning outflows of plasma in the jet column. The Poynting flux in the simulated jet is dominated by the untwisting motions of the magnetic fields loaded with high-density plasma. The simulated jet is comprised of ''spires'' of untwisting field that are loaded with a mixture of cold and hot plasma and exhibit rotational motion of order 20 km s{sup –1} and match contemporary observations.

  14. Detection of thermal neutrons with a CMOS pixel sensor for a future dosemeter

    SciTech Connect (OSTI)

    Vanstalle, M.; Husson, D.; Higueret, S.; Le, T. D.; Nourreddine, A. M.

    2011-07-01

    The RaMsEs group (Radioprotection et Mesures Environnementales) is developing a new compact device for operational neutron dosimetry. The electronic part of the detector is made of an integrated active pixel sensor, originally designed for tracking in particle physics. This device has useful features for neutrons, such as high detection efficiency for charged particles, good radiation resistance, high readout speed, low power consumption and high rejection of photon background. A good response of the device to fast neutrons has already been demonstrated [1]. In order to test the sensibility of the detector to thermal neutrons, experiments have been carried out with a 512 x 512 pixel CMOS sensor on a californium source moderated with heavy water (Cf.D{sub 2}O) on the Van Gogh irradiator at the LMDN, IRSN, Cadarache (France)). A thin boron converter is used to benefit from the significant cross section of the {sup 10}B (n,{alpha}) {sup 7}Li reaction. Results show a high detection efficiency (around 10{sup -3}) of the device to thermal neutrons. Our measurements are in good agreement with GEANT4 Monte Carlo simulations. (authors)

  15. Coupling of Time-Dependent Neutron Transport Theory with the Thermal Hydraulics Code ATHLET and Application to the Research Reactor FRM-II

    SciTech Connect (OSTI)

    Pautz, Andreas; Birkhofer, Adolf

    2003-11-15

    We introduce a new coupled neutronics/thermal hydraulics code system for analyzing transients of nuclear power plants and research reactors, based on a neutron transport theory approach. For the neutron kinetics, we have developed the code DORT-TD, a time-dependent extension of the well-known discrete ordinates code DORT. DORT-TD uses a fully implicit time integration scheme and is coupled via a general interface to the thermal hydraulics system code ATHLET, a generally applicable code for the analyses of LWR accident scenarios. Feedback is accounted for by interpolating multigroup cross sections from precalculated libraries, which are generated in advance for user-specified, discrete sets of thermal hydraulic parameters, e.g., fuel and coolant temperature. The coupled code system is applied to the high-flux research reactor FRM-II (Germany). Several design basis accidents are considered, namely the unintended control rod withdrawal, the loss of offsite power, and the loss of the secondary heat sink as well as a hypothetical transient with large reactivity insertion.

  16. The alanine detector in BNCT dosimetry: Dose response in thermal and epithermal neutron fields

    SciTech Connect (OSTI)

    Schmitz, T.; Bassler, N.; Blaickner, M.; Ziegner, M.; Hsiao, M. C.; Liu, Y. H.; Koivunoro, H.; Auterinen, I.; Sern, T.; Kotiluoto, P.; Palmans, H.; Sharpe, P.; Langguth, P.; Hampel, G.

    2015-01-15

    Purpose: The response of alanine solid state dosimeters to ionizing radiation strongly depends on particle type and energy. Due to nuclear interactions, neutron fields usually also consist of secondary particles such as photons and protons of diverse energies. Various experiments have been carried out in three different neutron beams to explore the alanine dose response behavior and to validate model predictions. Additionally, application in medical neutron fields for boron neutron capture therapy is discussed. Methods: Alanine detectors have been irradiated in the thermal neutron field of the research reactor TRIGA Mainz, Germany, in five experimental conditions, generating different secondary particle spectra. Further irradiations have been made in the epithermal neutron beams at the research reactors FiR 1 in Helsinki, Finland, and Tsing Hua open pool reactor in HsinChu, Taiwan ROC. Readout has been performed with electron spin resonance spectrometry with reference to an absorbed dose standard in a {sup 60}Co gamma ray beam. Absorbed doses and dose components have been calculated using the Monte Carlo codes FLUKA and MCNP. The relative effectiveness (RE), linking absorbed dose and detector response, has been calculated using the Hansen and Olsen alanine response model. Results: The measured dose response of the alanine detector in the different experiments has been evaluated and compared to model predictions. Therefore, a relative effectiveness has been calculated for each dose component, accounting for its dependence on particle type and energy. Agreement within 5% between model and measurement has been achieved for most irradiated detectors. Significant differences have been observed in response behavior between thermal and epithermal neutron fields, especially regarding dose composition and depth dose curves. The calculated dose components could be verified with the experimental results in the different primary and secondary particle fields. Conclusions: The alanine detector can be used without difficulty in neutron fields. The response has been understood with the model used which includes the relative effectiveness. Results and the corresponding discussion lead to the conclusion that application in neutron fields for medical purpose is limited by its sensitivity but that it is a useful tool as supplement to other detectors and verification of neutron source descriptions.

  17. 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.

  18. Determination of Thermal Neutron Capture Cross-Sections at Budapest PGAA Facility

    SciTech Connect (OSTI)

    Revay, Zsolt; Belgya, Tamas; Firestone, Richard B.

    2007-10-26

    Prompt gamma activation analysis (PGAA) is a powerful nuclear analytical technique to determine the elemental and isotopic composition of materials. The PGAA facility at Budapest, Hungary is one of the leading laboratories of the world, determining spectroscopic data for chemical analysis to be used in other laboratories. These partial gamma-ray production cross-sections and k{sub 0} values, being proportional to the analytical sensitivities of the chemical elements, can be transformed into thermal neutron capture cross-sections, i.e. the probabilities of the (n,{gamma}) reactions, which are of broader interest in different fields of nuclear physics. Some preliminary results on thermal neutron capture cross-sections are presented.

  19. High-efficiency scintillation detector for combined of thermal and fast neutrons and gamma radiation

    DOE Patents [OSTI]

    Chiles, Marion M. (Knoxville, TN); Mihalczo, John T. (Oak Ridge, TN); Blakeman, Edward D. (Oak Ridge, TN)

    1989-01-01

    A scintillation based radiation detector for the combined detection of thermal neutrons, high-energy neutrons and gamma rays in a single detecting unit. The detector consists of a pair of scintillators sandwiched together and optically coupled to the light sensitive face of a photomultiplier tube. A light tight radiation pervious housing is disposed about the scintillators and a portion of the photomultiplier tube to hold the arrangement in assembly and provides a radiation window adjacent the outer scintillator through which the radiation to be detected enters the detector. The outer scintillator is formed of a material in which scintillations are produced by thermal-neutrons and the inner scintillator is formed of a material in which scintillations are produced by high-energy neutrons and gamma rays. The light pulses produced by events detected in both scintillators are coupled to the photomultiplier tube which produces a current pulse in response to each detected event. These current pulses may be processed in a conventional manner to produce a count rate output indicative of the total detected radiation even count rate. Pulse discrimination techniques may be used to distinguish the different radiations and their energy distribution.

  20. RBMK coupled neutronics/thermal-hydraulics analyses by two independent code systems

    SciTech Connect (OSTI)

    Parisi, C.; D'Auria, F.; Malofeev, V.; Ivanov, B.; Ivanov, K.

    2006-07-01

    This paper presents the coupled neutronics/thermal-hydraulics activities carried out in the framework of the part B of the TACIS project R2.03/97, 'Software development for accident analysis of RBMK reactors in Russia'. Two independent code systems were assembled, one from the Russian side and the other from the Western side, for studying RBMK core transients. The Russian code system relies on the use of code UNK for neutron data libraries generation and the three-dimensional neutron kinetics thermal-hydraulics coupled codes BARS-KORSAR for plant transient analyses. The Western code system is instead based on the lattice physics code HELIOS and on the RELAP5-3D C code. Several activities were performed for testing code system's capabilities: the neutron data libraries were calculated and verified by precise Monte Carlo calculations, the coupled codes' steady state results were compared with plant detectors' data, and calculations of several transients were compared. Finally, both code systems proved to have all the capabilities for addressing reliable safety analyses of RBMK reactors. (authors)

  1. High-efficiency scintillation detector for combined detection of thermal and fast neutrons and gamma radiation

    DOE Patents [OSTI]

    Chiles, M.M.; Mihalczo, J.T.; Blakeman, E.D.

    1987-02-27

    A scintillation based radiation detector for the combined detection of thermal neutrons, high-energy neutrons and gamma rays in a single detecting unit. The detector consists of a pair of scintillators sandwiched together and optically coupled to the light sensitive face of a photomultiplier tube. A light tight radiation pervious housing is disposed about the scintillators and a portion of the photomultiplier tube to hold the arrangement in assembly and provides a radiation window adjacent the outer scintillator through which the radiation to be detected enters the detector. The outer scintillator is formed of a material in which scintillations are produced by thermal-neutrons and the inner scintillator is formed of a material in which scintillations are produced by high-energy neutrons and gamma rays. The light pulses produced by events detected in both scintillators are coupled to the photomultiplier tube which produces a current pulse in response to each detected event. These current pulses may be processed in a conventional manner to produce a count rate output indicative of the total detected radiation event count rate. Pulse discrimination techniques may be used to distinguish the different radiations and their energy distribution.

  2. Neutron, gamma ray and post-irradiation thermal annealing effects on power semiconductor switches

    SciTech Connect (OSTI)

    Schwarze, G.E.; Frasca, A.J.

    1994-09-01

    The effects of neutrons and gamma rays on the electrical and switching characteristics of power semiconductor switches must be known and understood by the designer of the power conditioning, control, and transmission subsystem of space nuclear power systems. The SP-100 radiation requirements at 25 m from the nuclear source are a neutron fluence of 10{sup 13} n/cm {sup 2} and a gamma dose of 0.5 Mrads. Experimental data showing the effects of neutrons and gamma rays on the performance characteristics of power-type NPN Bipolar Junction Transistors (BJTs), Metal-Oxide-Semiconductor Field Effect Transistors (MOSFETs), and Static Induction Transistors (SITs) are given in this paper. These three types of devices were tested at radiation levels which met or exceeded the SP-100 requirements. For the SP-100 radiation requirements, the BJTs were found to be most sensitive to neutrons, the MOSFETs were most sensitive to gamma rays, and the SITs were only slightly sensitive to neutrons. Post-irradiation thermal anneals at 300 K and up to 425 K were done on these devices and the effectiveness of these anneals are also discussed.

  3. Method of assaying uranium with prompt fission and thermal neutron borehole logging adjusted by borehole physical characteristics. [Patient application

    DOE Patents [OSTI]

    Barnard, R.W.; Jensen, D.H.

    1980-11-05

    Uranium formations are assayed by prompt fission neutron logging techniques. The uranium in the formation is proportional to the ratio of epithermal counts to thermal or epithermal dieaway. Various calibration factors enhance the accuracy of the measurement.

  4. Method of assaying uranium with prompt fission and thermal neutron borehole logging adjusted by borehole physical characteristics

    DOE Patents [OSTI]

    Barnard, Ralston W. (Albuquerque, NM); Jensen, Dal H. (Albuquerque, NM)

    1982-01-01

    Uranium formations are assayed by prompt fission neutron logging techniques. The uranium in the formation is proportional to the ratio of epithermal counts to thermal or eqithermal dieaway. Various calibration factors enhance the accuracy of the measurement.

  5. Neutronic reactor

    DOE Patents [OSTI]

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

    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.

  6. Neutron emission effects on fragment mass and kinetic energy distribution from fission of 239{sup Pu} induced by thermal neutrons

    SciTech Connect (OSTI)

    Montoya, M.; Rojas, J.; Lobato, I.

    2010-08-04

    The average of fragment kinetic energy (E-bar sign*) and the multiplicity of prompt neutrons ({nu}(bar sign)) as a function of fragment mass (m*), as well as the fragment mass yield (Y(m*)) from thermal neutron-induced fission of {sup 239}Pu have been measured by Tsuchiya et al.. In that work the mass and kinetic energy are calculated from the measured kinetic energy of one fragment and the difference of time of flight of the two complementary fragments. However they do not present their results about the standard deviation {sigma}{sub E}*(m*). In this work we have made a numerical simulation of that experiment which reproduces its results, assuming an initial distribution of the primary fragment kinetic energy (E(A)) with a constant value of the standard deviation as function of fragment mass ({sigma}{sub E}(A)). As a result of the simulation we obtain the dependence {sigma}{sub E}*(m*) which presents an enhancement between m* = 92 and m* = 110, and a peak at m* = 121.

  7. Supplementary neutron-flux calculations for the ORNL Pool Critical Assembly Pressure Vessel Facility

    SciTech Connect (OSTI)

    Maudlin, P.J.; Maerker, R.E.

    1982-01-01

    A three-dimensional Monte Carlo calculation using the MORSE code was performed to validate a procedure previously adopted in the ORNL discrete ordinate analysis of measurements made in the ORNL Pool Critical Assembly Pressure Vessel Facility. The results of these flux calculations agree, within statistical undertainties of about 5%, with those obtained from a discrete ordinate analysis employing the same procedure. This study therefore concludes that the procedure for combining several one- and two-dimensional discrete ordinate calculations into a three-dimensional flux is sufficiently accurate that it does not account for the existing discrepancies observed between calculations and measurements in this facility.

  8. Extraction of protactinium-233 and separation from thermal neutron-irradiated thorium-232 using crown ethers

    SciTech Connect (OSTI)

    Jalhoom, Moayyed G.; Mohammed, Dawood A.; Khalaf, Jumah S.

    2008-07-01

    A new method was developed for the extraction and separation of {sup 233}Pa from thermal neutron-irradiated {sup 232}Th. Solutions of Pa{sup 233} were prepared in LiCI-HCl solutions from which appreciable extraction was obtained using dibenzo-18-crown-6 in 1,2-dichloroethane. The effects of cavity size, substitutions on the crown ring, type of the organic solvent, and temperature on extraction are discussed. Very high separation factors were obtained for the pairs {sup 233}Pa/{sup 232}Th (>105), {sup 233}Pa/{sup 233}U (> 1000), and {sup 232}U/{sup 232}Th (>60). (authors)

  9. 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.

  10. On the thermal neutron transport processes in liquid H/sub 2/O-D/sub 2/O mixtures

    SciTech Connect (OSTI)

    Barnsal, R.M.; Tewari, S.P.

    1983-06-01

    Using the recently developed thermal neutron scattering kernels for water and heavy water, which incorporate both the collective and the molecular modes present in water and heavy water, the thermal neutron transport studies of asymptotic decay constants lambda/sub 0/, diffusion coefficient D/sub 0/, diffusion cooling coefficient C, and the transport mean-free-path lambda /SUB tr/ are studied for liquid H/sub 2/O-D/sub 2/O mixtures with varying molecular contents and for various assembly sizes at 21 and 5/sup 0/C. The calculated values of the physical constants, lambda/sub 0/, D/sub 0/, C, and lambda /SUB tr/ are found to be in good agreement with the corresponding experimental results. Both the collective motion and the molecular modes present in the liquid H/sub 2/O-D/sub 2/O mixtures play significant roles in the thermal neutron transport processes.

  11. Proceedings of the OECD/CSNI workshop on transient thermal-hydraulic and neutronic codes requirements

    SciTech Connect (OSTI)

    Ebert, D.

    1997-07-01

    This is a report on the CSNI Workshop on Transient Thermal-Hydraulic and Neutronic Codes Requirements held at Annapolis, Maryland, USA November 5-8, 1996. This experts` meeting consisted of 140 participants from 21 countries; 65 invited papers were presented. The meeting was divided into five areas: (1) current and prospective plans of thermal hydraulic codes development; (2) current and anticipated uses of thermal-hydraulic codes; (3) advances in modeling of thermal-hydraulic phenomena and associated additional experimental needs; (4) numerical methods in multi-phase flows; and (5) programming language, code architectures and user interfaces. The workshop consensus identified the following important action items to be addressed by the international community in order to maintain and improve the calculational capability: (a) preserve current code expertise and institutional memory, (b) preserve the ability to use the existing investment in plant transient analysis codes, (c) maintain essential experimental capabilities, (d) develop advanced measurement capabilities to support future code validation work, (e) integrate existing analytical capabilities so as to improve performance and reduce operating costs, (f) exploit the proven advances in code architecture, numerics, graphical user interfaces, and modularization in order to improve code performance and scrutibility, and (g) more effectively utilize user experience in modifying and improving the codes.

  12. Development and Validation of Temperature Dependent Thermal Neutron Scattering Laws for Applications and Safety Implications in Generation IV Reactor Designs

    SciTech Connect (OSTI)

    Ayman Hawari

    2008-06-20

    The overall obljectives of this project are to critically review the currently used thermal neutron scattering laws for various moderators as a function of temperature, select as well documented and representative set of experimental data sensitive to the neutron spectra to generate a data base of benchmarks, update models and models parameters by introducing new developments in thermalization theory and condensed matter physics into various computational approaches in establishing the scattering laws, benchmark the results against the experimentatl set. In the case of graphite, a validation experiment is performed by observing nutron slowing down as a function of temperatures equal to or greater than room temperature.

  13. Effects of Neutron Emission on Fragment Mass and Kinetic Energy Distribution from Thermal Neutron-Induced Fission of {sup 235}U

    SciTech Connect (OSTI)

    Montoya, M.; Rojas, J.; Saetone, E.

    2007-10-26

    The mass and kinetic energy distribution of nuclear fragments from thermal neutron-induced fission of {sup 235}U(n{sub th},f) have been studied using a Monte-Carlo simulation. Besides reproducing the pronounced broadening in the standard deviation of the kinetic energy at the final fragment mass number around m = 109, our simulation also produces a second broadening around m = 125. These results are in good agreement with the experimental data obtained by Belhafaf et al. and other results on yield of mass. We conclude that the obtained results are a consequence of the characteristics of the neutron emission, the sharp variation in the primary fragment kinetic energy and mass yield curves. We show that because neutron emission is hazardous to make any conclusion on primary quantities distribution of fragments from experimental results on final quantities distributions.

  14. Safety analysis of high pressure 3He-filled micro-channels for thermal neutron detection.

    SciTech Connect (OSTI)

    Ferko, Scott M.; Galambos, Paul C.; Derzon, Mark Steven; Renzi, Ronald F.

    2008-11-01

    This document is a safety analysis of a novel neutron detection technology developed by Sandia National Laboratories. This technology is comprised of devices with tiny channels containing high pressure {sup 3}He. These devices are further integrated into large scale neutron sensors. Modeling and preliminary device testing indicates that the time required to detect the presence of special nuclear materials may be reduced under optimal conditions by several orders of magnitude using this approach. Also, these devices make efficient use of our {sup 3}He supply by making individual devices more efficient and/or extending the our limited {sup 3}He supply. The safety of these high pressure devices has been a primary concern. We address these safety concerns for a flat panel configuration intended for thermal neutron detection. Ballistic impact tests using 3 g projectiles were performed on devices made from FR4, Silicon, and Parmax materials. In addition to impact testing, operational limits were determined by pressurizing the devices either to failure or until they unacceptably leaked. We found that (1) sympathetic or parasitic failure does not occur in pressurized FR4 devices (2) the Si devices exhibited benign brittle failure (sympathetic failure under pressure was not tested) and (3) the Parmax devices failed unacceptably. FR4 devices were filled to pressures up to 4000 + 100 psig, and the impacts were captured using a high speed camera. The brittle Si devices shattered, but were completely contained when wrapped in thin tape, while the ductile FR4 devices deformed only. Even at 4000 psi the energy density of the compressed gas appears to be insignificant compared to the impact caused by the incoming projectile. In conclusion, the current FR4 device design pressurized up to 4000 psi does not show evidence of sympathetic failure, and these devices are intrinsically safe.

  15. Space nuclear-power reactor design based on combined neutronic and thermal-fluid analyses

    SciTech Connect (OSTI)

    Koenig, D.R.; Gido, R.G.; Brandon, D.I.

    1985-01-01

    The design and performance analysis of a space nuclear-power system requires sophisticated analytical capabilities such as those developed during the nuclear rocket propulsion (Rover) program. In particular, optimizing the size of a space nuclear reactor for a given power level requires satisfying the conflicting requirements of nuclear criticality and heat removal. The optimization involves the determination of the coolant void (volume) fraction for which the reactor diameter is a minimum and temperature and structural limits are satisfied. A minimum exists because the critical diameter increases with increasing void fraction, whereas the reactor diameter needed to remove a specified power decreases with void fraction. The purpose of this presentation is to describe and demonstrate our analytical capability for the determination of minimum reactor size. The analysis is based on combining neutronic criticality calculations with OPTION-code thermal-fluid calculations.

  16. 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.

  17. An in-line thermal-neutron coincidence counter for WIPP (Waste Isolation Pilot Plant) certification measurements

    SciTech Connect (OSTI)

    Krick, M.S.; Osborne, L.; Polk, P.J.; Atencio, J.D.; Bjork, C.

    1989-10-01

    A custom-designed, in-line, thermal-neutron coincidence counter has been constructed for the certification of plutonium waste intended for storage at the Waste Isolation Pilot Plant. The mechanical and electrical components of the system and its performance characteristics are described. 6 refs., 16 figs.

  18. Investigation of the possibility to use a fine-mesh solver for resolving coupled neutronics and thermal-hydraulics

    SciTech Connect (OSTI)

    Jareteg, K.; Vinai, P.; Demaziere, C.

    2013-07-01

    The development of a fine-mesh coupled neutronic/thermal-hydraulic solver is touched upon in this paper. The reported work investigates the feasibility of using finite volume techniques to discretize a set of conservation equations modeling neutron transport, fluid dynamics, and heat transfer within a single numerical tool. With the long-term objective of developing fine-mesh computing capabilities for a few selected fuel assemblies in a nuclear core, this preliminary study considers an infinite array of a single fuel assembly having a finite height. Thermal-hydraulic conditions close to the ones existing in PWRs are taken as a first test case. The neutronic modeling relies on the diffusion approximation in a multi-energy group formalism, with cross-sections pre-calculated and tabulated at the sub-pin level using a Monte Carlo technique. The thermal-hydraulics is based on the Navier-Stokes equations, complemented by an energy conservation equation. The non-linear coupling terms between the different conservation equations are fully resolved using classical iteration techniques. Early tests demonstrate that the numerical tool provides an unprecedented level of details of the coupled solution estimated within the same numerical tool and thus avoiding any external data transfer, using fully consistent models between the neutronics and the thermal-hydraulics. (authors)

  19. A semi-experimental nodal synthesis method for the on-line reconstruction of three-dimensional neutron flux-shapes and reactivity

    SciTech Connect (OSTI)

    Jacqmin, R.P.

    1991-12-10

    The safety and optimal performance of large, commercial, light-water reactors require the knowledge at all time of the neutron-flux distribution in the core. In principle, this information can be obtained by solving the time-dependent neutron diffusion equations. However, this approach is complicated and very expensive. Sufficiently accurate, real-time calculations (time scale of approximately one second) are not yet possible on desktop computers, even with fast-running, nodal kinetics codes. A semi-experimental, nodal synthesis method which avoids the solution of the time-dependent, neutron diffusion equations is described. The essential idea of this method is to approximate instantaneous nodal group-fluxes by a linear combination of K, precomputed, three-dimensional, static expansion-functions. The time-dependent coefficients of the combination are found from the requirement that the reconstructed flux-distribution agree in a least-squares sense with the readings of J ({ge}K) fixed, prompt-responding neutron-detectors. Possible numerical difficulties with the least-squares solution of the ill-conditioned, J-by-K system of equations are brought under complete control by the use of a singular-value-decomposition technique. This procedure amounts to the rearrangement of the original, linear combination of K expansion functions into an equivalent more convenient, linear combination of R ({le}K) orthogonalized modes'' of decreasing magnitude. Exceedingly small modes are zeroed to eliminate any risk of roundoff-error amplification, and to assure consistency with the limited accuracy of the data. Additional modes are zeroed when it is desirable to limit the sensitivity of the results to measurement noise.

  20. A semi-experimental nodal synthesis method for the on-line reconstruction of three-dimensional neutron flux-shapes and reactivity. Final report

    SciTech Connect (OSTI)

    Jacqmin, R.P.

    1991-12-10

    The safety and optimal performance of large, commercial, light-water reactors require the knowledge at all time of the neutron-flux distribution in the core. In principle, this information can be obtained by solving the time-dependent neutron diffusion equations. However, this approach is complicated and very expensive. Sufficiently accurate, real-time calculations (time scale of approximately one second) are not yet possible on desktop computers, even with fast-running, nodal kinetics codes. A semi-experimental, nodal synthesis method which avoids the solution of the time-dependent, neutron diffusion equations is described. The essential idea of this method is to approximate instantaneous nodal group-fluxes by a linear combination of K, precomputed, three-dimensional, static expansion-functions. The time-dependent coefficients of the combination are found from the requirement that the reconstructed flux-distribution agree in a least-squares sense with the readings of J ({ge}K) fixed, prompt-responding neutron-detectors. Possible numerical difficulties with the least-squares solution of the ill-conditioned, J-by-K system of equations are brought under complete control by the use of a singular-value-decomposition technique. This procedure amounts to the rearrangement of the original, linear combination of K expansion functions into an equivalent more convenient, linear combination of R ({le}K) orthogonalized ``modes`` of decreasing magnitude. Exceedingly small modes are zeroed to eliminate any risk of roundoff-error amplification, and to assure consistency with the limited accuracy of the data. Additional modes are zeroed when it is desirable to limit the sensitivity of the results to measurement noise.

  1. Neutron source

    DOE Patents [OSTI]

    Cason, J.L. Jr.; Shaw, C.B.

    1975-10-21

    A neutron source which is particularly useful for neutron radiography consists of a vessel containing a moderating media of relatively low moderating ratio, a flux trap including a moderating media of relatively high moderating ratio at the center of the vessel, a shell of depleted uranium dioxide surrounding the moderating media of relatively high moderating ratio, a plurality of guide tubes each containing a movable source of neutrons surrounding the flux trap, a neutron shield surrounding one part of each guide tube, and at least one collimator extending from the flux trap to the exterior of the neutron source. The shell of depleted uranium dioxide has a window provided with depleted uranium dioxide shutters for each collimator. Reflectors are provided above and below the flux trap and on the guide tubes away from the flux trap.

  2. Neutron dosimetry

    DOE Patents [OSTI]

    Quinby, Thomas C.

    1976-07-27

    A method of measuring neutron radiation within a nuclear reactor is provided. A sintered oxide wire is disposed within the reactor and exposed to neutron radiation. The induced radioactivity is measured to provide an indication of the neutron energy and flux within the reactor.

  3. Study on in situ calibration for neutron flux monitor in the Large Helical Device based on Monte Carlo calculations

    SciTech Connect (OSTI)

    Nakano, Y. Yamazaki, A.; Watanabe, K.; Uritani, A.; Ogawa, K.; Isobe, M.

    2014-11-15

    Neutron monitoring is important to manage safety of fusion experiment facilities because neutrons are generated in fusion reactions. Monte Carlo simulations play an important role in evaluating the influence of neutron scattering from various structures and correcting differences between deuterium plasma experiments and in situ calibration experiments. We evaluated these influences based on differences between the both experiments at Large Helical Device using Monte Carlo simulation code MCNP5. A difference between the both experiments in absolute detection efficiency of the fission chamber between O-ports is estimated to be the biggest of all monitors. We additionally evaluated correction coefficients for some neutron monitors.

  4. Novel Materials and Devices for Solid-State Neutron Detection

    SciTech Connect (OSTI)

    Manginell, Ronald P.; Pfeifer, Kent B.

    2015-11-01

    There is a need in many fields, such as nuclear medicine, non-proliferation, energy exploration, national security, homeland security, nuclear energy, etc, for miniature, thermal neutron detectors. Until recently, thermal neutron detection has required physically large devices to provide sufficient neutron interaction and transduction signal. Miniaturization would allow broader use in the fields just mentioned and open up other applications potentially. Recent research shows promise in creating smaller neutron detectors through the combination of high-neutron-cross-section converter materials and solid-state devices. Yet, till recently it is difficult to measure low neutron fluxes by solidstate means given the need for optimized converter materials (purity, chemical composition and thickness) and a lack of designs capable of efficient transduction of the neutron conversion products (x-rays, electrons, gamma rays). Gadolinium-based semiconductor heterojunctions have detected electrons produced by Gd-neutron reactions but only at high neutron fluxes. One of the main limitations to this type of approach is the use of thin converter layers and the inability to utilize all the conversion products. In this LDRD we have optimized the converter material thickness and chemical composition to improve capture of conversion electrons and have detected thermal neutrons with high fidelity at low flux. We are also examining different semiconductor materials and converter materials to attempt to capture a greater percentage of the conversion electrons, both low and higher energy varieties. We have studied detector size and bias scaling, and cross-sensitivity to xrays and shown that we can detect low fluxes of thermal neutrons in less than 30 minutes with high selectivity by our approach. We are currently studying improvements in performance with direct placement of the Gd converter on the detector. The advancement of sensitive, miniature neutron detectors will have benefits in energy production, nonproliferation and medicine.

  5. 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.

  6. A New On-the-Fly Sampling Method for Incoherent Inelastic Thermal Neutron Scattering Data in MCNP6

    SciTech Connect (OSTI)

    Pavlou, Andrew Theodore; Brown, Forrest B.; Ji, Wei

    2014-09-02

    At thermal energies, the scattering of neutrons in a system is complicated by the comparable velocities of the neutron and target, resulting in competing upscattering and downscattering events. The neutron wavelength is also similar in size to the target's interatomic spacing making the scattering process a quantum mechanical problem. Because of the complicated nature of scattering at low energies, the thermal data files in ACE format used in continuous-energy Monte Carlo codes are quite large { on the order of megabytes for a single temperature and material. In this paper, a new storage and sampling method is introduced that is orders of magnitude less in size and is used to sample scattering parameters at any temperature on-the-fly. In addition to the reduction in storage, the need to pre-generate thermal scattering data tables at fine temperatures has been eliminated. This is advantageous for multiphysics simulations which may involve temperatures not known in advance. A new module was written for MCNP6 that bypasses the current S(?,?) table lookup in favor of the new format. The new on-the-fly sampling method was tested for graphite for two benchmark problems at ten temperatures: 1) an eigenvalue test with a fuel compact of uranium oxycarbide fuel homogenized into a graphite matrix, 2) a surface current test with a \\broomstick" problem with a monoenergetic point source. The largest eigenvalue difference was 152pcm for T= 1200K. For the temperatures and incident energies chosen for the broomstick problem, the secondary neutron spectrum showed good agreement with the traditional S(?,?) sampling method. These preliminary results show that sampling thermal scattering data on-the-fly is a viable option to eliminate both the storage burden of keeping thermal data at discrete temperatures and the need to know temperatures before simulation runtime.

  7. Estimating Liquid Fluxes in Thermally Perturbed Fractured Rock Using Measured Temperature Profiles

    SciTech Connect (OSTI)

    J.T. Birkholzer

    2005-02-14

    A new temperature-profile method was recently developed for analyzing perturbed flow conditions in superheated porous media. The method uses high-resolution temperature data to estimate the magnitude of the heat-driven liquid and gas fluxes that form as a result of boiling, condensation, and recirculation of pore water. In this paper, we evaluate the applicability of this new method to the more complex flow behavior in fractured formations with porous rock matrix. In such formations, with their intrinsic heterogeneity, the porous but low-permeable matrix provides most of the mass and heat storage capacity, and dominates conductive heat transfer, Fractures, on the other hand, offer highly effective conduits for gas and liquid flow, thereby generating significant convective heat transfer. After establishing the accuracy of the temperature-profile method for fractured porous formations, we apply the method in analyzing the perturbed flow conditions in a large-scale underground heater test conducted in unsaturated fractured porous tuff. The flux estimates for this test indicate a significant reflux of water near the heat source, on the order of a few hundred millimeter per year-much larger than the ambient percolation flux of only a few millimeter per year.

  8. A Combined Neutronic-Thermal Hydraulic Model of CERMET NTR Reactor

    SciTech Connect (OSTI)

    Jonathan A. Webb; Brian Gross; William T. Taitano

    2011-02-01

    Abstract. Two different CERMET fueled Nuclear Thermal Propulsion reactors were modeled to determine the optimum coolant channel surface area to volume ratio required to cool a 25,000 lbf rocket engine operating at a specific impulse of 940 seconds. Both reactor concepts were computationally fueled with hexagonal cross section fuel elements having a flat-to-flat distance of 3.51 cm and containing 60 vol.% UO2 enriched to 93wt.%U235 and 40 vol.% tungsten. Coolant channel configuration consisted of a 37 coolant channel fuel element and a 61 coolant channel model representing 0.3 and 0.6 surface area to volume ratios respectively. The energy deposition from decelerating fission products and scattered neutrons and photons was determined using the MCNP monte carlo code and then imported into the STAR-CCM+ computational fluid dynamics code. The 37 coolant channel case was shown to be insufficient in cooling the core to a peak temperature of 3000 K; however, the 61 coolant channel model shows promise for maintaining a peak core temperature of 3000 K, with no more refinements to the surface area to volume ratio. The core was modeled to have a power density of 9.34 GW/m3 with a thrust to weight ratio of 5.7.

  9. Methodology of Internal Assessment of Uncertainty and Extension to Neutron Kinetics/Thermal-Hydraulics Coupled Codes

    SciTech Connect (OSTI)

    Petruzzi, A.; D'Auria, F.; Giannotti, W.; Ivanov, K.

    2005-02-15

    The best-estimate calculation results from complex system codes are affected by approximations that are unpredictable without the use of computational tools that account for the various sources of uncertainty.The code with (the capability of) internal assessment of uncertainty (CIAU) has been previously proposed by the University of Pisa to realize the integration between a qualified system code and an uncertainty methodology and to supply proper uncertainty bands each time a nuclear power plant (NPP) transient scenario is calculated. The derivation of the methodology and the results achieved by the use of CIAU are discussed to demonstrate the main features and capabilities of the method.In a joint effort between the University of Pisa and The Pennsylvania State University, the CIAU method has been recently extended to evaluate the uncertainty of coupled three-dimensional neutronics/thermal-hydraulics calculations. The result is CIAU-TN. The feasibility of the approach has been demonstrated, and sample results related to the turbine trip transient in the Peach Bottom NPP are shown. Notwithstanding that the full implementation and use of the procedure requires a database of errors not available at the moment, the results give an idea of the errors expected from the present computational tools.

  10. SMITHERS: An object-oriented modular mapping methodology for MCNP-based neutronicthermal hydraulic multiphysics

    SciTech Connect (OSTI)

    Richard, Joshua; Galloway, Jack; Fensin, Michael; Trellue, Holly

    2015-04-04

    A novel object-oriented modular mapping methodology for externally coupled neutronicsthermal hydraulics multiphysics simulations was developed. The Simulator using MCNP with Integrated Thermal-Hydraulics for Exploratory Reactor Studies (SMITHERS) code performs on-the-fly mapping of material-wise power distribution tallies implemented by MCNP-based neutron transport/depletion solvers for use in estimating coolant temperature and density distributions with a separate thermal-hydraulic solver. The key development of SMITHERS is that it reconstructs the hierarchical geometry structure of the material-wise power generation tallies from the depletion solver automatically, with only a modicum of additional information required from the user. In addition, it performs the basis mapping from the combinatorial geometry of the depletion solver to the required geometry of the thermal-hydraulic solver in a generalizable manner, such that it can transparently accommodate varying levels of thermal-hydraulic solver geometric fidelity, from the nodal geometry of multi-channel analysis solvers to the pin-cell level of discretization for sub-channel analysis solvers.

  11. Mixing in thermally stratified nonlinear spin-up with uniform boundary fluxes

    SciTech Connect (OSTI)

    Baghdasarian, Meline; Pacheco-Vega, Arturo; Pacheco, J. Rafael; Verzicco, Roberto

    2014-09-15

    Studies of stratified spin-up experiments in enclosed cylinders have reported the presence of small pockets of well-mixed fluids but quantitative measurements of the mixedness of the fluid has been lacking. Previous numerical simulations have not addressed these measurements. Here we present numerical simulations that explain how the combined effect of spin-up and thermal boundary conditions enhances or hinders mixing of a fluid in a cylinder. The energy of the system is characterized by splitting the potential energy into diabatic and adiabatic components, and measurements of efficiency of mixing are based on both, the ratio of dissipation of available potential energy to forcing and variance of temperature. The numerical simulations of the NavierStokes equations for the problem with different sets of thermal boundary conditions at the horizontal walls helped shed some light on the physical mechanisms of mixing, for which a clear explanation was absent.

  12. Thermal neutron steady-state spectra in light water reactor fuel assemblies poisoned with various non-1/v absorbers of different concentrations

    SciTech Connect (OSTI)

    Swaminathan, K.; Chandra, S.; Jha, R.C.; Tewari, S.P. )

    1991-07-01

    This paper reports on the thermal neutron scattering kernel that explicitly incorporates the presence of chemical binding energy and the collective oscillations in the dynamics of water, the steady-state thermal neutron spectra in light water reactor fuel assemblies poisoned with non-1/v absorbers, such as cadmium, samarium, erbium, and gadolinium, in various concentrations have been computed at 298 K. The calculated spectra are in reasonable agreement with the corresponding experimental spectra for realistic source terms.

  13. Fission Product Yields of {sup 233}U, {sup 235}U, {sup 238}U and {sup 239}Pu in Fields of Thermal Neutrons, Fission Neutrons and 14.7-MeV Neutrons

    SciTech Connect (OSTI)

    Laurec, J.; Adam, A.; Bruyne, T. de [Commissariat a l'Energie Atomique, Centre DAM-Ile de France (CEA DAM DIF), 91297 Arpajon (France); Bauge, E., E-mail: eric.bauge@cea.f [Commissariat a l'Energie Atomique, Centre DAM-Ile de France (CEA DAM DIF), 91297 Arpajon (France); Granier, T.; Aupiais, J.; Bersillon, O.; Le Petit, G. [Commissariat a l'Energie Atomique, Centre DAM-Ile de France (CEA DAM DIF), 91297 Arpajon (France); Authier, N.; Casoli, P. [Commissariat a l'Energie Atomique, Centre de Valduc, 21120 Is-sur-Tille (France)

    2010-12-15

    The yields of more than fifteen fission products have been carefully measured using radiochemical techniques, for {sup 235}U(n,f), {sup 239}Pu(n,f) in a thermal spectrum, for {sup 233}U(n,f), {sup 235}U(n,f), and {sup 239}Pu(n,f) reactions in a fission neutron spectrum, and for {sup 233}U(n,f), {sup 235}U(n,f), {sup 238}U(n,f), and {sup 239}Pu(n,f) for 14.7 MeV monoenergetic neutrons. Irradiations were performed at the EL3 reactor, at the Caliban and Prospero critical assemblies, and at the Lancelot electrostatic accelerator in CEA-Valduc. Fissions were counted in thin deposits using fission ionization chambers. The number of fission products of each species were measured by gamma spectrometry of co-located thick deposits.

  14. ARCADIA{sup R} - A New Generation of Coupled Neutronics / Core Thermal- Hydraulics Code System at AREVA NP

    SciTech Connect (OSTI)

    Curca-Tivig, Florin; Merk, Stephan; Pautz, Andreas; Thareau, Sebastien

    2007-07-01

    Anticipating future needs of our customers and willing to concentrate synergies and competences existing in the company for the benefit of our customers, AREVA NP decided in 2002 to develop the next generation of coupled neutronics/ core thermal-hydraulic (TH) code systems for fuel assembly and core design calculations for both, PWR and BWR applications. The global CONVERGENCE project was born: after a feasibility study of one year (2002) and a conceptual phase of another year (2003), development was started at the beginning of 2004. The present paper introduces the CONVERGENCE project, presents the main feature of the new code system ARCADIA{sup R} and concludes on customer benefits. ARCADIA{sup R} is designed to meet AREVA NP market and customers' requirements worldwide. Besides state-of-the-art physical modeling, numerical performance and industrial functionality, the ARCADIA{sup R} system is featuring state-of-the-art software engineering. The new code system will bring a series of benefits for our customers: e.g. improved accuracy for heterogeneous cores (MOX/ UOX, Gd...), better description of nuclide chains, and access to local neutronics/ thermal-hydraulics and possibly thermal-mechanical information (3D pin by pin full core modeling). ARCADIA is a registered trademark of AREVA NP. (authors)

  15. Measured thermal and fast neutron fluence rates, ATR Cycle 100-BC, April 23, 1993--May 13, 1993

    SciTech Connect (OSTI)

    Smith, L.D.; Murray, R.K.; Rogers, J.W.

    1993-07-01

    This report contains the thermal (2200 m/s) and fast (E>1MeV) neutron fluence rate data for ATR Cycle 100-BC which were measured by the Radiation Measurements Laboratory (RML) as requested by the Power Reactor Programs (ATR Experiments) Radiation Measurements Work Order. This report contains fluence rate values corresponding to the particular elevations (relative to the 80 ft. core elevation) where the measurements were taken. The data in this report consists of (1) a table of the ATR power history and distribution, (2) a hard copy listing of all thermal and fast neutron fluence rates, (3) plots of both the thermal and fast neutron fluence rates, and (4) a magnetic record (3.5 inch diskette) containing a listing of only the fast neutron fluence rates, their assigned elevations and proper header identification of all monitor positions contained herein. The fluence rates reported are for the average power levels given in the table of power history and distribution. All {open_quotes}H{close_quotes} holder monitor wires for this cycle are 54 inches long. All {open_quotes}SR{close_quotes} holder monitor wires for this cycle are 55 inches long. This length allows measurement of the full core region and makes the first count elevation 24.73 inches above core midplane. Due to the safety rod problems in the west lobe, {open_quotes}BR{close_quotes} holders were used in the W-1, 2, 3, and 4 positions. All {open_quotes}BR{close_quotes} holder monitor wires for this cycle are 56.25 inches long. The distance from the end of the wires to the first count position was 4.25 inches for all wires counted from this cycle. The results from the measurements in the W-1, 2, 3, 4 monitor positions indicate that the safety rod followers were rotated to a different azimuthal orientation relative to the normal orientation. The results indicate that the rotation was counterclockwise from their normal orientation. This is the same condition observed starting with Cycle 99-B.

  16. Neutron cross-section libraries in the AMPX master interface format for thermal and fast reactors

    SciTech Connect (OSTI)

    Bjerke, M.A.; Webster, C.C.

    1981-12-01

    Neutron cross-section libraries in the AMPX master interface format have been created for three reactor types. Included are an 84-group library for use with light-water reactors, a 27-group library for use with heavy-water CANDU reactors and a 126-group library for use with liquid metal fast breeder reactors. In general, ENDF/B data were used in the creation of these libraries, and the nuclides included in each library should be sufficient for most neutronic analyses of reactors of that type. Each library has been used successfully in fuel depletion calculations.

  17. Thermal stability of photovoltaic a-Si:H determined by neutron reflectometry

    SciTech Connect (OSTI)

    Qviller, A. J. Haug, H.; You, C. C.; Hasle, I. M.; Marstein, E. S.; Frommen, C.; Hauback, B. C.; Dennison, A. J. C.; Vorobiev, A.; streng, E.; Fjellvg, H.; Hjrvarsson, B.

    2014-12-08

    Neutron and X-ray reflectometry were used to determine the layer structure and hydrogen content of thin films of amorphous silicon (a-Si:H) deposited onto crystalline silicon (Si) wafers for surface passivation in solar cells. The combination of these two reflectometry techniques is well suited for non-destructive probing of the structure of a-Si:H due to being able to probe buried interfaces and having sub-nanometer resolution. Neutron reflectometry is also unique in its ability to allow determination of density gradients of light elements such as hydrogen (H). The neutron scattering contrast between Si and H is strong, making it possible to determine the H concentration in the deposited a-Si:H. In order to correlate the surface passivation properties supplied by the a-Si:H thin films, as quantified by obtainable effective minority carrier lifetime, photoconductance measurements were also performed. It is shown that the minority carrier lifetime falls sharply when H has been desorbed from a-Si:H by annealing.

  18. 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.

  19. The OECD/NEA/NSC PBMR coupled neutronics/thermal hydraulics transient benchmark: The PBMR-400 core design

    SciTech Connect (OSTI)

    Reitsma, F.; Ivanov, K.; Downar, T.; De Haas, H.; Gougar, H. D.

    2006-07-01

    The Pebble Bed Modular Reactor (PBMR) is a High-Temperature Gas-cooled Reactor (HTGR) concept to be built in South Africa. As part of the verification and validation program the definition and execution of code-to-code benchmark exercises are important. The Nuclear Energy Agency (NEA) of the Organisation for Economic Cooperation and Development (OECD) has accepted, through the Nuclear Science Committee (NSC), the inclusion of the Pebble-Bed Modular Reactor (PBMR) coupled neutronics/thermal hydraulics transient benchmark problem in its program. The OECD benchmark defines steady-state and transients cases, including reactivity insertion transients. It makes use of a common set of cross sections (to eliminate uncertainties between different codes) and includes specific simplifications to the design to limit the need for participants to introduce approximations in their models. In this paper the detailed specification is explained, including the test cases to be calculated and the results required from participants. (authors)

  20. Verification of the coupled 3-D neutronics and thermal-hydraulic code SKETCH-INS/TRAC-P

    SciTech Connect (OSTI)

    Nakajima, T.; Takeuchi, A.

    2006-07-01

    In order to analyze the complex transients with coupled interactions between core behavior and plant dynamics, the three-dimensional neutronics code SKETCH-INS was coupled with the thermal-hydraulic code TRAC-P. The capability SKETCH-INS code was verified against the 3-D transient benchmark problem. The capability of the coupled code SKETCH-INS/TRAC-P was verified against the NEACRP 3-D LWR core transient benchmark and OECD MSLB benchmark problems. The results of analyses were in reasonable agreement with the reference and different codes results of benchmarks. This paper provides the outline of the coupled code SKETCH-INS/TRAC-P and the results of benchmarks. (authors)

  1. The new Cold Neutron Chopper Spectrometer at the Spallation Neutron Source -- Design and Performance

    SciTech Connect (OSTI)

    Ehlers, Georg; Podlesnyak, Andrey A.; Niedziela, Jennifer L.; Iverson, Erik B.; Sokol, Paul E.

    2011-01-01

    The design and performance of the new cold neutron chopper spectrometer (CNCS) at the Spallation Neutron Source in Oak Ridge are described. CNCS is a direct-geometry inelastic time-of-flight spectrometer, designed essentially to cover the same energy and momentum transfer ranges as IN5 at ILL, LET at ISIS, DCS at NIST, TOFTOF at FRM-II, AMATERAS at J-PARC, PHAROS at LANSCE, and NEAT at HZB, at similar energy resolution. Measured values of key figures such as neutron flux at sample position and energy resolution are compared between measurements and ray tracing Monte Carlo simulations, and good agreement (better than 20% of absolute numbers) has been achieved. The instrument performs very well in the cold and thermal neutron energy ranges, and promises to become a workhorse for the neutron scattering community for quasielastic and inelastic scattering experiments.

  2. The new cold neutron chopper spectrometer at the Spallation Neutron Source: Design and performance

    SciTech Connect (OSTI)

    Ehlers, G.; Podlesnyak, A. A.; Niedziela, J. L.; Iverson, E. B.; Sokol, P. E.

    2011-08-15

    The design and performance of the new cold neutron chopper spectrometer (CNCS) at the Spallation Neutron Source in Oak Ridge are described. CNCS is a direct-geometry inelastic time-of-flight spectrometer, designed essentially to cover the same energy and momentum transfer ranges as IN5 at ILL, LET at ISIS, DCS at NIST, TOFTOF at FRM-II, AMATERAS at J-PARC, PHAROS at LANSCE, and NEAT at HZB, at similar energy resolution. Measured values of key figures such as neutron flux at sample position and energy resolution are compared between measurements and ray tracing Monte Carlo simulations, and good agreement (better than 20% of absolute numbers) has been achieved. The instrument performs very well in the cold and thermal neutron energy ranges, and promises to become a workhorse for the neutron scattering community for quasielastic and inelastic scattering experiments.

  3. Accuracy and Efficiency of a Coupled Neutronics and Thermal Hydraulics Model

    SciTech Connect (OSTI)

    Vincent A. Mousseau; Michael A. Pope

    2007-09-01

    The accuracy requirements for modern nuclear reactor simulation are steadily increasing due to the cost and regulation of relevant experimental facilities. Because of the increase in the cost of experiments and the decrease in the cost of simulation, simulation will play a much larger role in the design and licensing of new nuclear reactors. Fortunately as the work load of simulation increases, there are better physics models, new numerical techniques, and more powerful computer hardware that will enable modern simulation codes to handle the larger workload. This manuscript will discuss a numerical method where the six equations of two-phase flow, the solid conduction equations, and the two equations that describe neutron diffusion and precursor concentration are solved together in a tightly coupled, nonlinear fashion for a simplified model of a nuclear reactor core. This approach has two important advantages. The first advantage is a higher level of accuracy. Because the equations are solved together in a single nonlinear system, the solution is more accurate than the traditional operator split approach where the two-phase flow equations are solved first, the heat conduction is solved second and the neutron diffusion is solved third, limiting the temporal accuracy to 1st order because the nonlinear coupling between the physics is handled explicitly. The second advantage of the method described in this manuscript is that the time step control in the fully implicit system can be based on the timescale of the solution rather than a stability-based time step restriction like the material Courant. Results are presented from a simulated control rod movement and a rod ejection that address temporal accuracy for the fully coupled solution and demonstrate how the fastest timescale of the problem can change between the state variables of neutronics, conduction and two-phase flow during the course of a transient.

  4. Accuracy and Efficiency of a Coupled Neutronics and Thermal Hydraulics Model

    SciTech Connect (OSTI)

    Michael A. Pope; Vincent A. Mousseau

    2009-09-01

    The accuracy requirements for modern nuclear reactor simulation are steadily increasing due to the cost and regulation of relevant experimental facilities. Because of the increase in the cost of experiments and the decrease in the cost of simulation, simulation will play a much larger role in the design and licensing of new nuclear reactors. Fortunately as the work load of simulation increases, there are better physics models, new numerical techniques, and more powerful computer hardware that will enable modern simulation codes to handle this larger workload. This manuscript will discuss a numerical method where the six equations of two-phase flow, the solid conduction equations, and the two equations that describe neutron diffusion and precursor concentration are solved together in a tightly coupled, nonlinear fashion for a simplified model of a nuclear reactor core. This approach has two important advantages. The first advantage is a higher level of accuracy. Because the equations are solved together in a single nonlinear system, the solution is more accurate than the traditional operator split approach where the two-phase flow equations are solved first, the heat conduction is solved second and the neutron diffusion is solved third, limiting the temporal accuracy to 1st order because the nonlinear coupling between the physics is handled explicitly. The second advantage of the method described in this manuscript is that the time step control in the fully implicit system can be based on the timescale of the solution rather than a stability-based time step restriction like the material Courant. Results are presented from a simulated control rod movement and a rod ejection that address temporal accuracy for the fully coupled solution and demonstrate how the fastest timescale of the problem can change between the state variables of neutronics, conduction and two-phase flow during the course of a transient.

  5. Format requirements of thermal neutron scattering data in a nuclear data format to succeed the ENDF format

    SciTech Connect (OSTI)

    Brown, D.

    2014-03-31

    In November 2012, the Working Party on Evaluation Cooperation Subgroup 38 (WPEC-SG38) began with the task of developing a nuclear data format and supporting infrastructure to replace the now nearly 50 year old ENDF format. The first step in this process is to develop requirements for the new format and infrastructure. In this talk, I will review the status of ENDF's Thermal Scattering Law (TSL) formats as well as support for this data in the GND format (from which the new format is expected to evolve). Finally, I hope to begin a dialog with members of the thermal neutron scattering community so that their data needs can be accurately and easily accommodated by the new format and tools, as captured by the requirements document. During this discussion, we must keep in mind that the new tools and format must; Support what is in existing data files; Support new things we want to put in data files; and Be flexible enough for us to adapt it to future unanticipated challenges.

  6. Leap Frog and Time Step Sub-Cycle Scheme for Coupled Neutronics and Thermal-Hydraulic Codes

    SciTech Connect (OSTI)

    Lu, S.

    2002-07-01

    As the result of the advancing TCP/IP based inter-process communication technology, more and more legacy thermal-hydraulic codes have been coupled with neutronics codes to provide best-estimate capabilities for reactivity related reactor transient analysis. Most of the coupling schemes are based on closely coupled serial or parallel approaches. Therefore, the execution of the coupled codes usually requires significant CPU time, when a complicated system is analyzed. Leap Frog scheme has been used to reduce the run time. The extent of the decoupling is usually determined based on a trial and error process for a specific analysis. It is the intent of this paper to develop a set of general criteria, which can be used to invoke the automatic Leap Frog algorithm. The algorithm will not only provide the run time reduction but also preserve the accuracy. The criteria will also serve as the base of an automatic time step sub-cycle scheme when a sudden reactivity change is introduced and the thermal-hydraulic code is marching with a relatively large time step. (authors)

  7. Critical experiments on an enriched uranium solution system containing periodically distributed strong thermal neutron absorbers

    SciTech Connect (OSTI)

    Rothe, R.E.

    1996-09-30

    A series of 62 critical and critical approach experiments were performed to evaluate a possible novel means of storing large volumes of fissile solution in a critically safe configuration. This study is intended to increase safety and economy through use of such a system in commercial plants which handle fissionable materials in liquid form. The fissile solution`s concentration may equal or slightly exceed the minimum-critical-volume concentration; and experiments were performed for high-enriched uranium solution. Results should be generally applicable in a wide variety of plant situations. The method is called the `Poisoned Tube Tank` because strong neutron absorbers (neutron poisons) are placed inside periodically spaced stainless steel tubes which separate absorber material from solution, keeping the former free of contamination. Eight absorbers are investigated. Both square and triangular pitched lattice patterns are studied. Ancillary topics which closely model typical plant situations are also reported. They include the effect of removing small bundles of absorbers as might occur during inspections in a production plant. Not taking the tank out of service for these inspections would be an economic advantage. Another ancillary topic studies the effect of the presence of a significant volume of unpoisoned solution close to the Poisoned Tube Tank on the critical height. A summary of the experimental findings is that boron compounds were excellent absorbers, as expected. This was true for granular materials such as Gerstley Borate and Borax; but it was also true for the flexible solid composed of boron carbide and rubber, even though only thin sheets were used. Experiments with small bundles of absorbers intentionally removed reveal that quite reasonable tanks could be constructed that would allow a few tubes at a time to be removed from the tank for inspection without removing the tank from production service.

  8. Neutron detector

    DOE Patents [OSTI]

    Stephan, Andrew C.; Jardret; Vincent D.

    2011-04-05

    A neutron detector has a volume of neutron moderating material and a plurality of individual neutron sensing elements dispersed at selected locations throughout the moderator, and particularly arranged so that some of the detecting elements are closer to the surface of the moderator assembly and others are more deeply embedded. The arrangement captures some thermalized neutrons that might otherwise be scattered away from a single, centrally located detector element. Different geometrical arrangements may be used while preserving its fundamental characteristics. Different types of neutron sensing elements may be used, which may operate on any of a number of physical principles to perform the function of sensing a neutron, either by a capture or a scattering reaction, and converting that reaction to a detectable signal. High detection efficiency, an ability to acquire spectral information, and directional sensitivity may be obtained.

  9. Apparatus and method for identification of matrix materials in which transuranic elements are embedded using thermal neutron capture gamma-ray emission

    DOE Patents [OSTI]

    Close, D.A.; Franks, L.A.; Kocimski, S.M.

    1984-08-16

    An invention is described that enables the quantitative simultaneous identification of the matrix materials in which fertile and fissile nuclides are embedded to be made along with the quantitative assay of the fertile and fissile materials. The invention also enables corrections for any absorption of neutrons by the matrix materials and by the measurement apparatus by the measurement of the prompt and delayed neutron flux emerging from a sample after the sample is interrogated by simultaneously applied neutrons and gamma radiation. High energy electrons are directed at a first target to produce gamma radiation. A second target receives the resulting pulsed gamma radiation and produces neutrons from the interaction with the gamma radiation. These neutrons are slowed by a moderator surrounding the sample and bathe the sample uniformly, generating second gamma radiation in the interaction. The gamma radiation is then resolved and quantitatively detected, providing a spectroscopic signature of the constituent elements contained in the matrix and in the materials within the vicinity of the sample. (LEW)

  10. Supercritical Water Nuclear Steam Supply System: Innovations In Materials, Neutronics & Thermal-Hydraulics

    SciTech Connect (OSTI)

    Mark Anderson; M.L. Corradini; K. Sridharan; P. WIlson; D. Cho; T.K. Kim; S. Lomperski

    2004-09-02

    In the 1990's supercritical light-water reactors were considered in conceptual designs. A nuclear reactor cooled by supercritical waster would have a much higher thermal efficiency with a once-through direct power cycle, and could be based on standardized water reactor components (light water or heavy water). The theoretical efficiency could be improved by more than 33% over that of other water reactors and could be simplified with higher reliability; e.g., a boiling water reactor without steam separators or dryers.

  11. Neutrons - 88-Inch Cyclotron

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

    Neutrons Neutron beams are available at the 88-Inch Cyclotron. Available energies range of from 8 to 30 MeV, with fluxes of up to 1E8 neutrons/cm^2/sec. For more information, please contact Mike Johnson via e-mail at MBJohnson@lbl.gov, or by phone at at (510) 486-4389.

  12. 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.

  13. Organic metal neutron detector

    DOE Patents [OSTI]

    Butler, Michael A.; Ginley, David S.

    1987-01-01

    A device for detecting neutrons comprises a layer of conductive polymer sandwiched between electrodes, which may be covered on each face with a neutron transmissive insulating material layer. Conventional electrodes are used for a non-imaging integrating total neutron fluence-measuring embodiment, while wire grids are used in an imaging version of the device. The change in conductivity of the polymer after exposure to a neutron flux is determined in either case to provide the desired data. Alternatively, the exposed conductive polymer layer may be treated with a chemical reagent which selectively binds to the sites altered by neutrons to produce an image of the flux detected.

  14. FAST NEUTRON DOSIMETER FOR HIGH TEMPERATURE OPERATION BY MEASUREMENT OF THE AMOUNT OF CESIUM 137 FORMED FROM A THORIUM WIRE

    DOE Patents [OSTI]

    McCune, D.A.

    1964-03-17

    A method and device for measurement of integrated fast neutron flux in the presence of a large thermal neutron field are described. The device comprises a thorium wire surrounded by a thermal neutron attenuator that is, in turn, enclosed by heat-resistant material. The method consists of irradiating the device in a neutron field whereby neutrons with energies in excess of 1.1 Mev cause fast fissions in the thorium, then removing the thorium wire, separating the cesium-137 fission product by chemical means from the thorium, and finally counting the radioactivity of the cesium to determine the number of fissions which have occurred so that the integrated fast flux may be obtained. (AEC)

  15. 2010 Neutron Review: ORNL Neutron Sciences Progress Report (Technical

    Office of Scientific and Technical Information (OSTI)

    Report) | SciTech Connect Technical Report: 2010 Neutron Review: ORNL Neutron Sciences Progress Report Citation Details In-Document Search Title: 2010 Neutron Review: ORNL Neutron Sciences Progress Report During 2010, the Neutron Sciences Directorate focused on producing world-class science, while supporting the needs of the scientific community. As the instrument, sample environment, and data analysis tools at High Flux Isotope Reactor (HFIR ) and Spallation Neutron Source (SNS) have grown

  16. Innovative and Advanced Coupled Neutron Transport and Thermal Hydraulic Method (Tool) for the Design, Analysis and Optimization of VHTR/NGNP Prismatic Reactors

    SciTech Connect (OSTI)

    Rahnema, Farzad; Garimeela, Srinivas; Ougouag, Abderrafi; Zhang, Dingkang

    2013-11-29

    This project will develop a 3D, advanced coarse mesh transport method (COMET-Hex) for steady- state and transient analyses in advanced very high-temperature reactors (VHTRs). The project will lead to a coupled neutronics and thermal hydraulic (T/H) core simulation tool with fuel depletion capability. The computational tool will be developed in hexagonal geometry, based solely on transport theory without (spatial) homogenization in complicated 3D geometries. In addition to the hexagonal geometry extension, collaborators will concurrently develop three additional capabilities to increase the codes versatility as an advanced and robust core simulator for VHTRs. First, the project team will develop and implement a depletion method within the core simulator. Second, the team will develop an elementary (proof-of-concept) 1D time-dependent transport method for efficient transient analyses. The third capability will be a thermal hydraulic method coupled to the neutronics transport module for VHTRs. Current advancements in reactor core design are pushing VHTRs toward greater core and fuel heterogeneity to pursue higher burn-ups, efficiently transmute used fuel, maximize energy production, and improve plant economics and safety. As a result, an accurate and efficient neutron transport, with capabilities to treat heterogeneous burnable poison effects, is highly desirable for predicting VHTR neutronics performance. This research projects primary objective is to advance the state of the art for reactor analysis.

  17. High Conduction Neutron Absorber to Simulate Fast Reactor Environment in an Existing Test Reactor

    SciTech Connect (OSTI)

    Guillen, Donna; Greenwood, Lawrence R.; Parry, James

    2014-06-22

    A need was determined for a thermal neutron absorbing material that could be cooled in a gas reactor environment without using large amounts of a coolant that would thermalize the neutron flux. A new neutron absorbing material was developed that provided high conduction so a small amount of water would be sufficient for cooling thereby thermalizing the flux as little as possible. An irradiation experiment was performed to assess the effects of radiation and the performance of a new neutron absorbing material. Neutron fluence monitors were placed inside specially fabricated holders within a set of drop-in capsules and irradiated for up to four cycles in the Advanced Test Reactor. Following irradiation, the neutron fluence monitor wires were analyzed by gamma and x-ray spectrometry to determine the activities of the activation products. The adjusted neutron fluences were calculated and grouped into three bins – thermal, epithermal and fast to evaluate the spectral shift created by the new material. Fluence monitors were evaluated after four different irradiation periods to evaluate the effects of burn-up in the absorbing material. Additionally, activities of the three highest activity isotopes present in the specimens are given.

  18. Fission Product Data Measured at Los Alamos for Fission Spectrum and Thermal Neutrons on {sup 239}Pu, {sup 235}U, {sup 238}U

    SciTech Connect (OSTI)

    Selby, H.D.; Mac Innes, M.R.; Barr, D.W.; Keksis, A.L.; Meade, R.A.; Burns, C.J.; Chadwick, M.B.; Wallstrom, T.C.

    2010-12-15

    We describe measurements of fission product data at Los Alamos that are important for determining the number of fissions that have occurred when neutrons are incident on plutonium and uranium isotopes. The fission-spectrum measurements were made using a fission chamber designed by the National Institute for Standards and Technology (NIST) in the BIG TEN critical assembly, as part of the Inter-laboratory Liquid Metal Fast Breeder Reactor (LMFBR) Reaction Rate (ILRR) collaboration. The thermal measurements were made at Los Alamos' Omega West Reactor. A related set of measurements were made of fission-product ratios (so-called R-values) in neutron environments provided by a number of Los Alamos critical assemblies that range from having average energies causing fission of 400-600 keV (BIG TEN and the outer regions of the Flattop-25 assembly) to higher energies (1.4-1.9 MeV) in the Jezebel, and in the central regions of the Flattop-25 and Flattop-Pu, critical assemblies. From these data we determine ratios of fission product yields in different fuel and neutron environments (Q-values) and fission product yields in fission spectrum neutron environments for {sup 99}Mo, {sup 95}Zr, {sup 137}Cs, {sup 140}Ba, {sup 141,143}Ce, and {sup 147}Nd. Modest incident-energy dependence exists for the {sup 147}Nd fission product yield; this is discussed in the context of models for fission that include thermal and dynamical effects. The fission product data agree with measurements by Maeck and other authors using mass-spectrometry methods, and with the ILRR collaboration results that used gamma spectroscopy for quantifying fission products. We note that the measurements also contradict earlier 1950s historical Los Alamos estimates by {approx}5-7%, most likely owing to self-shielding corrections not made in the early thermal measurements. Our experimental results provide a confirmation of the England-Rider ENDF/B-VI evaluated fission-spectrum fission product yields that were carried over to the ENDF/B-VII.0 library, except for {sup 99}Mo where the present results are about 4%-relative higher for neutrons incident on {sup 239}Pu and {sup 235}U. Additionally, our results illustrate the importance of representing the incident energy dependence of fission product yields over the fast neutron energy range for high-accuracy work, for example the {sup 147}Nd from neutron reactions on plutonium. An upgrade to the ENDF library, for ENDF/B-VII.1, based on these and other data, is described in a companion paper to this work.

  19. Proton recoil scintillator neutron rem meter

    DOE Patents [OSTI]

    Olsher, Richard H. (Los Alamos, NM); Seagraves, David T. (Los Alamos, NM)

    2003-01-01

    A neutron rem meter utilizing proton recoil and thermal neutron scintillators to provide neutron detection and dose measurement. In using both fast scintillators and a thermal neutron scintillator the meter provides a wide range of sensitivity, uniform directional response, and uniform dose response. The scintillators output light to a photomultiplier tube that produces an electrical signal to an external neutron counter.

  20. Solid state neutron detector array

    DOE Patents [OSTI]

    Seidel, John G. (Pittsburgh, PA); Ruddy, Frank H. (Monroeville, PA); Brandt, Charles D. (Mount Lebanon, PA); Dulloo, Abdul R. (Pittsburgh, PA); Lott, Randy G. (Pittsburgh, PA); Sirianni, Ernest (Monroeville, PA); Wilson, Randall O. (Greensburg, PA)

    1999-01-01

    A neutron detector array is capable of measuring a wide range of neutron fluxes. The array includes multiple semiconductor neutron detectors. Each detector has a semiconductor active region that is resistant to radiation damage. In one embodiment, the array preferably has a relatively small size, making it possible to place the array in confined locations. The ability of the array to detect a wide range of neutron fluxes is highly advantageous for many applications such as detecting neutron flux during start up, ramp up and full power of nuclear reactors.

  1. Solid state neutron detector array

    DOE Patents [OSTI]

    Seidel, J.G.; Ruddy, F.H.; Brandt, C.D.; Dulloo, A.R.; Lott, R.G.; Sirianni, E.; Wilson, R.O.

    1999-08-17

    A neutron detector array is capable of measuring a wide range of neutron fluxes. The array includes multiple semiconductor neutron detectors. Each detector has a semiconductor active region that is resistant to radiation damage. In one embodiment, the array preferably has a relatively small size, making it possible to place the array in confined locations. The ability of the array to detect a wide range of neutron fluxes is highly advantageous for many applications such as detecting neutron flux during start up, ramp up and full power of nuclear reactors. 7 figs.

  2. R-matrix analysis of the {sup 240}Pu neutron cross sections in the thermal to 5700 eV energy range

    SciTech Connect (OSTI)

    Derrien, H.; Bouland, O.; Larson, N.M.; Leal, L.C.

    1997-08-01

    Resonance analysis of high resolution neutron transmission data and of fission cross sections were performed in the neutron energy range from the thermal regions to 5,700 eV by using the Reich-Moore Bayesian code SAMMY. The experimental data base is described and the method of analysis is given. The experimental data were carefully examined in order to identify more resonances than those found in the current evaluated data files. The statistical properties of the resonance parameters are given. A new set of the average values of the parameters is proposed, which could be used for calculation of the average cross sections in the unresolved resonance region. The resonance parameters are available IN ENDF-6 format at the national or international data centers.

  3. Analysis of palladium coatings to remove hydrogen isotopes from zirconium fuel rods in Canada deuterium uranium-pressurized heavy water reactors; Thermal and neutron diffusion effects

    SciTech Connect (OSTI)

    Stokes, C.L.; Buxbaum, R.E. )

    1992-05-01

    This paper reports that, in pressurized heavy water nuclear reactors of the type standardly used in Canada (Canada deuterium uranium-pressurized heavy water reactors), the zirconium alloy pressure tubes of the core absorb deuterium produced by corrosion reactions. This deuterium weakens the tubes through hydrogen embrittlement. Thin palladium coatings on the outside of the zirconium are analyzed as a method for deuterium removal. This coating is expected to catalyze the reaction D{sub 2} + 1/2O{sub 2} {r reversible} D{sub 2}O when O{sub 2} is added to the annular (insulating) gas in the tubes. Major reductions in the deuterium concentration and, hence, hydrogen embrittlement are predicted. Potential problems such as plating the tube geometry, neutron absorption, catalyst deactivation, radioactive waste production, and oxygen corrosion are shown to be manageable. Also, a simple set of equations are derived to calculate the effect on diffusion caused by neutron interactions. Based on calculations of ordinary and neutron flux induced diffusion, a palladium coating of 1 {times} 10{sup {minus}6} m is recommended. This would cost approximately $60,000 per reactor unit and should more than double reactor lifetime. Similar coatings and similar interdiffusion calculations might have broad applications.

  4. Summary of comparison and analysis of results from exercises 1 and 2 of the OECD PBMR coupled neutronics/thermal hydraulics transient benchmark

    SciTech Connect (OSTI)

    Mkhabela, P.; Han, J.; Tyobeka, B.; Ivanov, K.; Reitsma, F.; Sartori, E.

    2006-07-01

    The Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) has accepted, through the Nuclear Science Committee (NSC), the inclusion of the Pebble-Bed Modular Reactor 400 MW design (PBMR-400) coupled neutronics/thermal hydraulics transient benchmark problem as part of their official activities. The scope of the benchmark is to establish a well-defined problem, based on a common given library of cross sections, to compare methods and tools in core simulation and thermal hydraulics analysis with a specific focus on transient events through a set of multi-dimensional computational test problems. The benchmark includes three steady state exercises and six transient exercises. This paper describes the first two steady state exercises, their objectives and the international participation in terms of organization, country and computer code utilized. This description is followed by a comparison and analysis of the participants' results submitted for these two exercises. The comparison of results from different codes allows for an assessment of the sensitivity of a result to the method employed and can thus help to focus the development efforts on the most critical areas. The two first exercises also allow for removing of user-related modeling errors and prepare core neutronics and thermal-hydraulics models of the different codes for the rest of the exercises in the benchmark. (authors)

  5. Neutronic reactor thermal shield

    DOE Patents [OSTI]

    Lowe, Paul E.

    1976-06-15

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

  6. FABRICATION OF NEUTRON SOURCES

    DOE Patents [OSTI]

    Birden, J.H.

    1959-01-20

    A method is presented for preparing a more efficient neutron source comprising inserting in a container a quantity of Po-210, inserting B powder coated with either Ag, Pt, or Ni. The container is sealed and then slowly heated to about 450 C to volatilize the Po and effect combination of the coated powder with the Po. The neutron flux emitted by the unit is moritored and the heating step is terminated when the flux reaches a maximum or selected level.

  7. Transport of thermal neutrons in D/sub 2/O in the temperature Range 5 to 60/sup 0/C based on a new scattering kernel

    SciTech Connect (OSTI)

    Bansal, R.M.; Kothari, L.S.; Tewari, S.P.

    1980-10-01

    A new scattering kernel for heavy water has been proposed. The kernel takes into account the chemical binding energy effects and also includes the rotational and intramolecular vibrational modes. Using this scattering kernel, various neutron transport processes in the temperature range 5 to 60/sup 0/C have been studied and compared with the corresponding experimental results. The calculated results include total neutron scattering cross section at 20/sup 0/C; asymptotic decay of neutron pulses in the temperature range 5 to 60/sup 0/C and temperature variation of the diffusion coefficient and diffusion cooling coefficient; timedependent spectra inside finite-sized assemblies of heavy water at 20 and 43.3/sup 0/C thermalization time; and diffusion length and space-dependent study in pure and poisoned assemblies of heavy water. The calculated results are in good agreement with the experimental results. At some places notable differences are observed between the results obtained using our scattering kernel and those based on the Honeck kernel.

  8. Experiments and Simulations of the Use of Time-Correlated Thermal Neutron Counting to Determine the Multiplication of an Assembly of Highly Enriched Uranium

    SciTech Connect (OSTI)

    David L. Chichester; Mathew T. Kinlaw; Scott M. Watson; Jeffrey M. Kalter; Eric C. Miller; William A. Noonan

    2014-11-01

    A series of experiments and numerical simulations using thermal-neutron time-correlated measurements has been performed to determine the neutron multiplication, M, of assemblies of highly enriched uranium available at Idaho National Laboratory. The experiments used up to 14.4 kg of highly-enriched uranium, including bare assemblies and assemblies reflected with high-density polyethylene, carbon steel, and tungsten. A small 252Cf source was used to initiate fission chains within the assembly. Both the experiments and the simulations used 6-channel and 8-channel detector systems, each consisting of 3He proportional counters moderated with polyethylene; data was recorded in list mode for analysis. 'True' multiplication values for each assembly were empirically derived using basic neutron production and loss values determined through simulation. A total of one-hundred and sixteen separate measurements were performed using fifty-seven unique measurement scenarios, the multiplication varied from 1.75 to 10.90. This paper presents the results of these comparisons and discusses differences among the various cases.

  9. Consideration of a ultracold neutron source in two-dimensional cylindrical geometry by taking simulated boundaries

    SciTech Connect (OSTI)

    Gheisari, R.; Firoozabadi, M. M.; Mohammadi, H.

    2014-01-15

    A new idea to calculate ultracold neutron (UCN) production by using Monte Carlo simulation method to calculate the cold neutron (CN) flux and an analytical approach to calculate the UCN production from the simulated CN flux was given. A super-thermal source (UCN source) was modeled based on an arrangement of D{sub 2}O and solid D{sub 2} (sD{sub 2}). The D{sub 2}O was investigated as the neutron moderator, and sD{sub 2} as the converter. In order to determine the required parameters, a two-dimensional (2D) neutron balance equation written in Matlab was combined with the MCNPX simulation code. The 2D neutron-transport equation in cylindrical (ρ − z) geometry was considered for 330 neutron energy groups in the sD{sub 2}. The 2D balance equation for UCN and CN was solved using simulated CN flux as boundary value. The UCN source dimensions were calculated for the development of the next UCN source. In the optimal condition, the UCN flux and the UCN production rate (averaged over the sD{sub 2} volume) equal to 6.79 × 10{sup 6} cm{sup −2}s{sup −1} and 2.20 ×10{sup 5} cm{sup −3}s{sup −1}, respectively.

  10. HYSPEC : A CRYSTAL TIME OF FLIGHT HYBRID SPECTROMETER FOR THE SPALLATION NEUTRON SOURCE.

    SciTech Connect (OSTI)

    SHAPIRO,S.M.; ZALIZNYAK,I.A.

    2002-12-30

    This document lays out a proposal by the Instrument Development Team (IDT) composed of scientists from leading Universities and National Laboratories to design and build a conceptually new high-flux inelastic neutron spectrometer at the pulsed Spallation Neutron Source (SNS) at Oak Ridge. This instrument is intended to supply users of the SNS and scientific community, of which the IDT is an integral part, with a platform for ground-breaking investigations of the low-energy atomic-scale dynamical properties of crystalline solids. It is also planned that the proposed instrument will be equipped with a polarization analysis capability, therefore becoming the first polarized beam inelastic spectrometer in the SNS instrument suite, and the first successful polarized beam inelastic instrument at a pulsed spallation source worldwide. The proposed instrument is designed primarily for inelastic and elastic neutron spectroscopy of single crystals. In fact, the most informative neutron scattering studies of the dynamical properties of solids nearly always require single crystal samples, and they are almost invariably flux-limited. In addition, in measurements with polarization analysis the available flux is reduced through selection of the particular neutron polarization, which puts even more stringent limits on the feasibility of a particular experiment. To date, these investigations have mostly been carried out on crystal spectrometers at high-flux reactors, which usually employ focusing Bragg optics to concentrate the neutron beam on a typically small sample. Construction at Oak Ridge of the high-luminosity spallation neutron source, which will provide intense pulsed neutron beams with time-averaged fluxes equal to those at medium-flux reactors, opens entirely new opportunities for single crystal neutron spectroscopy. Drawing upon experience acquired during decades of studies with both crystal and time-of-flight (TOF) spectrometers, the IDT has developed a conceptual design for a focused-beam, hybrid time-of-flight instrument with a crystal monochromator for the SNS called HYSPEC (an acronym for hybrid spectrometer). The proposed instrument has a potential to collect data more than an order of magnitude faster than existing steady-source spectrometers over a wide range of energy transfer ({h_bar}{omega}) and momentum transfer (Q) space, and will transform the way that data in elastic and inelastic single-crystal spectroscopy are collected. HYSPEC is optimized to provide the highest neutron flux on sample in the thermal and epithermal neutron energy ranges at a good-to-moderate energy resolution. By providing a flux on sample several times higher than other inelastic instruments currently planned for the SNS, the proposed instrument will indeed allow unique ground-breaking measurements, and will ultimately make polarized beam studies at a pulsed spallation source a realistic possibility.

  11. EIS-0247: Construction and Operation of the Spallation Neutron Source

    Broader source: Energy.gov [DOE]

    The United States needs a high-flux, short- pulsed neutron source to provide its scientific and industrial research communities with a much more intense source of pulsed neutrons for neutron...

  12. Neutronic Reactor Design to Reduce Neutron Loss

    DOE Patents [OSTI]

    Miles, F. T.

    1961-05-01

    A nuclear reactor construction is described in which an unmoderated layer of the fissionable material is inserted between the moderated portion of the reactor core and the core container steel wall. The wall is surrounded by successive layers of pure fertile material and moderator containing fertile material. The unmoderated layer of the fissionable material will insure that a greater portion of fast neutrons will pass through the steel wall than would thermal neutrons. Since the steel has a smaller capture cross section for the fast neutrons, greater nunnbers of neutrons will pass into the blanket, thereby increasing the over-all efficiency of the reactor. (AEC)

  13. NEUTRONIC REACTOR DESIGN TO REDUCE NEUTRON LOSS

    DOE Patents [OSTI]

    Mills, F.T.

    1961-05-01

    A nuclear reactor construction is described in which an unmoderated layer of the fissionable material is inserted between the moderated portion of the reactor core and the core container steel wall which is surrounded by successive layers of pure fertile material and fertile material having moderator. The unmoderated layer of the fissionable material will insure that a greater portion of fast neutrons will pass through the steel wall than would thermal neutrons. As the steel has a smaller capture cross-section for the fast neutrons, then greater numbers of the neutrons will pass into the blanket thereby increasing the over-all efficiency of the reactor.

  14. Intense fusion neutron sources

    SciTech Connect (OSTI)

    Kuteev, B. V.; Goncharov, P. R.; Sergeev, V. Yu.; Khripunov, V. I.

    2010-04-15

    The review describes physical principles underlying efficient production of free neutrons, up-to-date possibilities and prospects of creating fission and fusion neutron sources with intensities of 10{sup 15}-10{sup 21} neutrons/s, and schemes of production and application of neutrons in fusion-fission hybrid systems. The physical processes and parameters of high-temperature plasmas are considered at which optimal conditions for producing the largest number of fusion neutrons in systems with magnetic and inertial plasma confinement are achieved. The proposed plasma methods for neutron production are compared with other methods based on fusion reactions in nonplasma media, fission reactions, spallation, and muon catalysis. At present, intense neutron fluxes are mainly used in nanotechnology, biotechnology, material science, and military and fundamental research. In the near future (10-20 years), it will be possible to apply high-power neutron sources in fusion-fission hybrid systems for producing hydrogen, electric power, and technological heat, as well as for manufacturing synthetic nuclear fuel and closing the nuclear fuel cycle. Neutron sources with intensities approaching 10{sup 20} neutrons/s may radically change the structure of power industry and considerably influence the fundamental and applied science and innovation technologies. Along with utilizing the energy produced in fusion reactions, the achievement of such high neutron intensities may stimulate wide application of subcritical fast nuclear reactors controlled by neutron sources. Superpower neutron sources will allow one to solve many problems of neutron diagnostics, monitor nano-and biological objects, and carry out radiation testing and modification of volumetric properties of materials at the industrial level. Such sources will considerably (up to 100 times) improve the accuracy of neutron physics experiments and will provide a better understanding of the structure of matter, including that of the neutron itself.

  15. Neutron capture therapy with deep tissue penetration using capillary neutron focusing

    DOE Patents [OSTI]

    Peurrung, Anthony J.

    1997-01-01

    An improved method for delivering thermal neutrons to a subsurface cancer or tumor which has been first doped with a dopant having a high cross section for neutron capture. The improvement is the use of a guide tube in cooperation with a capillary neutron focusing apparatus, or neutron focusing lens, for directing neutrons to the tumor, and thereby avoiding damage to surrounding tissue.

  16. ATRC Neutron Detector Testing Quick Look Report

    SciTech Connect (OSTI)

    Troy C. Unruh; Benjamin M. Chase; Joy L. Rempe

    2013-08-01

    As part of the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program, a joint Idaho State University (ISU) / French Alternative Energies and Atomic Energy Commission (CEA) / Idaho National Laboratory (INL) project was initiated in FY-10 to investigate the feasibility of using neutron sensors to provide online measurements of the neutron flux and fission reaction rate in the ATR Critical Facility (ATRC). A second objective was to provide initial neutron spectrum and flux distribution information for physics modeling and code validation using neutron activation based techniques in ATRC as well as ATR during depressurized operations. Detailed activation spectrometry measurements were made in the flux traps and in selected fuel elements, along with standard fission rate distribution measurements at selected core locations. These measurements provide additional calibration data for the real-time sensors of interest as well as provide benchmark neutronics data that will be useful for the ATR Life Extension Program (LEP) Computational Methods and V&V Upgrade project. As part of this effort, techniques developed by Prof. George Imel will be applied by Idaho State University (ISU) for assessing the performance of various flux detectors to develop detailed procedures for initial and follow-on calibrations of these sensors. In addition to comparing data obtained from each type of detector, calculations will be performed to assess the performance of and reduce uncertainties in flux detection sensors and compare data obtained from these sensors with existing integral methods employed at the ATRC. The neutron detectors required for this project were provided to team participants at no cost. Activation detectors (foils and wires) from an existing, well-characterized INL inventory were employed. Furthermore, as part of an on-going ATR NSUF international cooperation, the CEA sent INL three miniature fission chambers (one for detecting fast flux and two for detecting thermal flux) with associated electronics for assessment. In addition, Prof. Imel, ISU, has access to an inventory of Self-Powered Neutron Detectors (SPNDs) with a range of response times as well as Back-to-Back (BTB) fission chambers from prior research he conducted at the Transient REActor Test Facility (TREAT) facility and Neutron RADiography (NRAD) reactors. Finally, SPNDs from the National Atomic Energy Commission of Argentina (CNEA) were provided in connection with the INL effort to upgrade ATR computational methods and V&V protocols that are underway as part of the ATR LEP. Work during fiscal year 2010 (FY10) focussed on design and construction of Experiment Guide Tubes (EGTs) for positioning the flux detectors in the ATRC N-16 locations as well as obtaining ATRC staff concurrence for the detector evaluations. Initial evaluations with CEA researchers were also started in FY10 but were cut short due to reactor reliability issues. Reactor availability issues caused experimental work to be delayed during FY11/12. In FY13, work resumed; and evaluations were completed. The objective of this "Quick Look" report is to summarize experimental activities performed from April 4, 2013 through May 16, 2013.

  17. Nickel-based gadolinium alloy for neutron adsorption application...

    Office of Scientific and Technical Information (OSTI)

    ... Sponsoring Org: USDOE Country of Publication: United States Language: English Subject: 11 ... NEUTRONS; NUCLEAR FUELS; SPENT FUELS; TESTING; THERMAL NEUTRONS; YUCCA MOUNTAIN Word ...

  18. High conduction neutron absorber to simulate fast reactor environment in an existing test reactor

    SciTech Connect (OSTI)

    Donna Post Guillen; Larry R. Greenwood; James R. Parry

    2014-06-22

    A new metal matrix composite material has been developed to serve as a thermal neutron absorber for testing fast reactor fuels and materials in an existing pressurized water reactor. The performance of this material was evaluated by placing neutron fluence monitors within shrouded and unshrouded holders and irradiating for up to four cycles. The monitor wires were analyzed by gamma and X-ray spectrometry to determine the activities of the activation products. Adjusted neutron fluences were calculated and grouped into three bins—thermal, epithermal, and fast—to evaluate the spectral shift created by the new material. A comparison of shrouded and unshrouded fluence monitors shows a thermal fluence decrease of ~11 % for the shielded monitors. Radioisotope activity and mass for each of the major activation products is given to provide insight into the evolution of thermal absorption cross-section during irradiation. The thermal neutron absorption capability of the composite material appears to diminish at total neutron fluence levels of ~8 × 1025 n/m2. Calculated values for dpa in excess of 2.0 were obtained for two common structural materials (iron and nickel) of interest for future fast flux experiments.

  19. 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.

  20. Radiation Damage Study in Natural Zircon Using Neutrons Irradiation

    SciTech Connect (OSTI)

    Lwin, Maung Tin Moe; Amin, Yusoff Mohd.; Kassim, Hasan Abu; Mohamed, Abdul Aziz; Karim, Julia Abdul

    2011-03-30

    Changes of atomic displacements in crystalline structure of natural zircon (ZrSiO{sub 4}) can be studied by using neutron irradiation on the surface of zircon and compared the data from XRD measurements before and after irradiation. The results of neutron irradiation on natural zircon using Pneumatic Transfer System (PTS) at PUSPATI TRIGA Research Reactor in the Malaysian Nuclear Agency are discussed in this work. The reactor produces maximum thermal power output of 1 MWatt and the neutron flux of up to 1x10{sup 13} ncm{sup -2}s{sup -1}. From serial decay processes of uranium and thorium radionuclides in zircon crystalline structure, the emission of alpha particles can produce damage in terms of atomic displacements in zircon. Hence, zircon has been extensively studied as a possible candidate for immobilization of fission products and actinides.

  1. Gamma and neutron detection modeling in the nuclear detection figure of merit (NDFOM) portal

    SciTech Connect (OSTI)

    Stroud, Phillip D [Los Alamos National Laboratory; Saeger, Kevin J [Los Alamos National Laboratory

    2009-01-01

    The Nuclear Detection Figure Of Merit (NDFOM) portal is a database of objects and algorithms for evaluating the performance of radiation detectors to detect nuclear material. This paper describes the algorithms used to model the physics and mathematics of radiation detection. As a first-principles end-to-end analysis system, it starts with the representation of the gamma and neutron spectral fluxes, which are computed with the particle and radiation transport code MCNPX. The gamma spectra emitted by uranium, plutonium, and several other materials of interest are described. The impact of shielding and other intervening material is computed by the method of build-up factors. The interaction of radiation with the detector material is computed by a detector response function approach. The construction of detector response function matrices based on MCNPX simulation runs is described in detail. Neutron fluxes are represented in a three group formulation to treat differences in detector sensitivities to thermal, epithermal, and fast neutrons.

  2. Boron nitride solid state neutron detector

    DOE Patents [OSTI]

    Doty, F. Patrick

    2004-04-27

    The present invention describes an apparatus useful for detecting neutrons, and particularly for detecting thermal neutrons, while remaining insensitive to gamma radiation. Neutrons are detected by direct measurement of current pulses produced by an interaction of the neutrons with hexagonal pyrolytic boron nitride.

  3. Coupled 3D-neutronics / thermal-hydraulics analysis of an unprotected loss-of-flow accident for a 3600 MWth SFR core

    SciTech Connect (OSTI)

    Sun, K.; Chenu, A.; Mikityuk, K.; Krepel, J.; Chawla, R.

    2012-07-01

    The core behaviour of a large (3600 MWth) sodium-cooled fast reactor (SFR) is investigated in this paper with the use of a coupled TRACE/PARCS model. The SFR neutron spectrum is characterized by several performance advantages, but also leads to one dominating neutronics drawback - a positive sodium void reactivity. This implies a positive reactivity effect when sodium coolant is removed from the core. In order to evaluate such feedback in terms of the dynamics, a representative unprotected loss-of-flow (ULOF) transient, i.e. flow run-down without SCRAM in which sodium boiling occurs, is analyzed. Although analysis of a single transient cannot allow general conclusions to be drawn, it does allow better understanding of the underlying physics and can lead to proposals for improving the core response during such an accident. The starting point of this study is the reference core design considered in the framework of the Collaborative Project on the European Sodium Fast Reactor (CP-ESFR). To reduce the void effect, the core has been modified by introducing an upper sodium plenum (along with a boron layer) and by reducing the core height-to-diameter ratio. For the ULOF considered, a sharp increase in core power results in melting of the fuel in the case of the reference core. In the modified core, a large dryout leads to melting of the clad. It seems that, for the hypothetical event considered, fuel failure cannot be avoided with just improvement of the neutronics design; therefore, thermal-hydraulics optimization has been considered. An innovative assembly design is proposed to prevent sodium vapour blocking the fuel channel. This results in preventing a downward propagation of the sodium boiling to the core center, thus limiting it to the upper region. Such a void map introduces a negative coolant density reactivity feedback, which dominates the total reactivity change. As a result, the power level and the fuel temperature are effectively reduced, and a large dryout is prevented. As a next step, the classical sodium plenum is replaced by a fission gas plenum (with lower sodium fraction), thus improving flow stability. Stable boiling at a steady power level is achieved in this final configuration. (authors)

  4. FABRICATION OF NEUTRON SOURCES

    DOE Patents [OSTI]

    Birden, J.H.

    1959-04-21

    A method is presented for preparing a neutron source from polonium-210 and substances, such as beryllium and boron, characterized by emission of neutrons upon exposure to alpha particles from the polonium. According to the invention, a source is prepared by placing powdered beryllium and a platinum foil electroplated with polonium-2;.0 in a beryllium container. The container is sealed and then heated by induction to a temperature of 450 to 1100 deg C to volatilize the polonium off the foil into the powder. The heating step is terminated upon detection of a maximum in the neutron flux level.

  5. 14 MeV neutron activation analysis of geological and lunar samples

    SciTech Connect (OSTI)

    Laul, J.C.; Wogman, N.A.

    1981-04-01

    14 MeV neutron activation analysis (NAA) is ideal for accurately determining Oxygen and Silicon contents in geological and lunar materials. It is fast, nondestructive, economical, and can be used on a routine basis in a laboratory. Although 14 MeV NAA is particularly suited to light elements, its use has been extended to measure other elements as well such as Aluminum, Magnesium, Iron, Calcium, Titanium, Strontium, Nickel, Yttrium, Zirconium, Niobium and Cerium. Thus, the use of 14 MeV neutrons is of considerable importance in NAA. The disadvantages of the method are that interference reactions are common because of high neutron energy; the flux is nonuniform in longer irradiation due to depletion of the target in the neutron generator. Overall, 14 MeV NAA is ideal for short irradiations and when supplemented with thermal NAA provides the maximum elemental information in small aliquants of geological and lunar materials.

  6. Measurement of cross sections for inelastic cold-neutron scattering in metals and polymers by the method of (n, {gamma}) analysis

    SciTech Connect (OSTI)

    Arzumanov, S. S.; Bondarenko, L. N.; Geltenbort, P.; Morozov, V. I.; Panin, Yu. N.; Chernyavsky, S. M.

    2008-11-15

    The results obtained by measuring the cross sections for the inelastic scattering of very cold neutrons for a number of metals and polymers by the method of a neutron-irradiation analysis are presented. The method is based on simultaneously measuring events of inelastic scattering and neutron capture in the sample under investigation via recording gamma radiation with a semiconductor germanium detector. Neutron capture by a nucleus of the sample is accompanied by the prompt radiation of gamma rays having a known spectrum. Upon inelastic scattering, a neutron acquires thermal energy. Upon leaving the sample, this neutron is absorbed in a special converter that contains the isotope {sup 10}B. The capture of the neutron by a {sup 10}B nucleus is followed by the emission of a 477-keV gamma ray. The probabilities of capture and inelastic scattering are proportional to the respective neutron-interaction cross sections, and the ratio of the recorded detector counts corresponding to events of the two types does not depend on the spectrum of the incident flux of very cold neutrons or on the trajectory of neutron motion in the sample. The sought inelastic-scattering cross section at a fixed sample temperature is calculated by using this ratio and the known cross section for neutron capture by the sample isotope having a known gamma-radiation spectrum.

  7. 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 effective tool for achieving the configuration and process control believed to be necessary for scientific instrument systems.

  8. 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.

  9. 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)

  10. Rapid cooling and structure of neutron stars

    SciTech Connect (OSTI)

    Van Riper, K.A.; Lattimer, J.M.

    1992-07-01

    This report discusses the following topics on neutron stars: direct URCA neutrino emission; thermal evolution models; analytic model for diffusion through the crust; and core superfluidity. (LSP).

  11. Rapid cooling and structure of neutron stars

    SciTech Connect (OSTI)

    Van Riper, K.A. ); Lattimer, J.M. . Dept. of Earth and Space Sciences)

    1992-01-01

    This report discusses the following topics on neutron stars: direct URCA neutrino emission; thermal evolution models; analytic model for diffusion through the crust; and core superfluidity. (LSP).

  12. Cylindrical neutron generator

    DOE Patents [OSTI]

    Leung, Ka-Ngo

    2005-06-14

    A cylindrical neutron generator is formed with a coaxial RF-driven plasma ion source and target. A deuterium (or deuterium and tritium) plasma is produced by RF excitation in a cylindrical plasma ion generator using an RF antenna. A cylindrical neutron generating target is coaxial with the ion generator, separated by plasma and extraction electrodes which contain many slots. The plasma generator emanates ions radially over 360.degree. and the cylindrical target is thus irradiated by ions over its entire circumference. The plasma generator and target may be as long as desired. The plasma generator may be in the center and the neutron target on the outside, or the plasma generator may be on the outside and the target on the inside. In a nested configuration, several concentric targets and plasma generating regions are nested to increase the neutron flux.

  13. Cylindrical neutron generator

    DOE Patents [OSTI]

    Leung, Ka-Ngo

    2008-04-22

    A cylindrical neutron generator is formed with a coaxial RF-driven plasma ion source and target. A deuterium (or deuterium and tritium) plasma is produced by RF excitation in a cylindrical plasma ion generator using an RF antenna. A cylindrical neutron generating target is coaxial with the ion generator, separated by plasma and extraction electrodes which contain many slots. The plasma generator emanates ions radially over 360.degree. and the cylindrical target is thus irradiated by ions over its entire circumference. The plasma generator and target may be as long as desired. The plasma generator may be in the center and the neutron target on the outside, or the plasma generator may be on the outside and the target on the inside. In a nested configuration, several concentric targets and plasma generating regions are nested to increase the neutron flux.

  14. Cylindrical neutron generator

    DOE Patents [OSTI]

    Leung, Ka-Ngo

    2009-12-29

    A cylindrical neutron generator is formed with a coaxial RF-driven plasma ion source and target. A deuterium (or deuterium and tritium) plasma is produced by RF excitation in a cylindrical plasma ion generator using an RF antenna. A cylindrical neutron generating target is coaxial with the ion generator, separated by plasma and extraction electrodes which contain many slots. The plasma generator emanates ions radially over 360.degree. and the cylindrical target is thus irradiated by ions over its entire circumference. The plasma generator and target may be as long as desired. The plasma generator may be in the center and the neutron target on the outside, or the plasma generator may be on the outside and the target on the inside. In a nested configuration, several concentric targets and plasma generating regions are nested to increase the neutron flux.

  15. NEUTRON MEASURING METHOD AND APPARATUS

    DOE Patents [OSTI]

    Seaborg, G.T.; Friedlander, G.; Gofman, J.W.

    1958-07-29

    A fast neutron fission detecting apparatus is described consisting of a source of fast neutrons, an ion chamber containing air, two electrodes within the ion chamber in confronting spaced relationship, a high voltage potential placed across the electrodes, a shield placed about the source, and a suitable pulse annplifier and recording system in the electrode circuit to record the impulse due to fissions in a sannple material. The sample material is coated onto the active surface of the disc electrode and shielding means of a material having high neutron capture capabilities for thermal neutrons are provided in the vicinity of the electrodes and about the ion chamber so as to absorb slow neutrons of thermal energy to effectively prevent their diffusing back to the sample and causing an error in the measurement of fast neutron fissions.

  16. The advanced neutron source research and development plan

    SciTech Connect (OSTI)

    Selby, D.L.

    1995-08-01

    The Advanced Neutron Source (ANS) is being designed as a user-oriented neutron research laboratory centered around the most intense continuous beams of thermal and subthermal neutrons in the world (an order of magnitude more intense than beams available from the most advanced existing reactors). The ANS will be built around a new research reactor of 330-MW fission power, producing an unprecedented peak thermal flux of >7 {center_dot} 10{sup 19} {center_dot} m{sup -2} {center_dot} s{sup -1}. Primarily a research facility, the ANS will accommodate more than 1000 academic, industrial, and government researchers each year. They will conduct basic research in all branches of science as well as applied research leading to better understanding of new materials, including high temperature super conductors, plastics, and thin films. Some 48 neutron beam stations will be set up in the ANS beam rooms and the neutron guide hall for neutron scattering and for fundamental and nuclear physics research. There also will be extensive facilities for materials irradiation, isotope production, and analytical chemistry. The top level work breakdown structure (WBS) for the project. As noted in this figure, one component of the project is a research and development (R&D) program (WBS 1.1). This program interfaces with all of the other project level two WBS activities. Because one of the project guidelines is to meet minimum performance goals without relying on new inventions, this R&D activity is not intended to produce new concepts to allow the project to meet minimum performance goals. Instead, the R&D program will focus on the four objectives described.

  17. Neutron capture therapies

    DOE Patents [OSTI]

    Yanch, Jacquelyn C.; Shefer, Ruth E.; Klinkowstein, Robert E.

    1999-01-01

    In one embodiment there is provided an application of the .sup.10 B(n,.alpha.).sup.7 Li nuclear reaction or other neutron capture reactions for the treatment of rheumatoid arthritis. This application, called Boron Neutron Capture Synovectomy (BNCS), requires substantially altered demands on neutron beam design than for instance treatment of deep seated tumors. Considerations for neutron beam design for the treatment of arthritic joints via BNCS are provided for, and comparisons with the design requirements for Boron Neutron Capture Therapy (BNCT) of tumors are made. In addition, exemplary moderator/reflector assemblies are provided which produce intense, high-quality neutron beams based on (p,n) accelerator-based reactions. In another embodiment there is provided the use of deuteron-based charged particle reactions to be used as sources for epithermal or thermal neutron beams for neutron capture therapies. Many d,n reactions (e.g. using deuterium, tritium or beryllium targets) are very prolific at relatively low deuteron energies.

  18. Maintenance neutron coincidence counter manual

    SciTech Connect (OSTI)

    Krick, M.S.; Polk, P.J.; Atencio, J.D.

    1989-09-01

    A compact thermal-neutron coincidence counter has been constructed specifically for use by the International Atomic Energy Agency as a reference neutron detector for maintenance activities. The counter is designed for use only with {sup 252}Cf sources in SR-CF-100 capsules. This manual describes the detector's mechanical and electrical components and its operating characteristics. 2 refs., 8 figs.

  19. Fission fragment driven neutron source

    DOE Patents [OSTI]

    Miller, Lowell G.; Young, Robert C.; Brugger, Robert M.

    1976-01-01

    Fissionable uranium formed into a foil is bombarded with thermal neutrons in the presence of deuterium-tritium gas. The resulting fission fragments impart energy to accelerate deuterium and tritium particles which in turn provide approximately 14 MeV neutrons by the reactions t(d,n).sup.4 He and d(t,n).sup.4 He.

  20. Neutron capture therapy with deep tissue penetration using capillary neutron focusing

    DOE Patents [OSTI]

    Peurrung, A.J.

    1997-08-19

    An improved method is disclosed for delivering thermal neutrons to a subsurface cancer or tumor which has been first doped with a dopant having a high cross section for neutron capture. The improvement is the use of a guide tube in cooperation with a capillary neutron focusing apparatus, or neutron focusing lens, for directing neutrons to the tumor, and thereby avoiding damage to surrounding tissue. 1 fig.

  1. APT target/blanket design and thermal hydraulics

    SciTech Connect (OSTI)

    Cappiello, M.; Pitcher, E.; Pasamehmetoglu, K.

    1999-04-01

    The Accelerator Production of Tritium (APT) Target/Blanket (T/B) system is comprised of an assembly of tritium producing modules supported by control, heat removal, shielding and retargeting systems. The T/B assembly produces tritium using a high-energy proton beam, a tungsten/lead spallation neutron source and {sup 3}He gas as the tritium producing feedstock. For the nominal production mode, protons are accelerated to an energy of 1030 MeV at a current of 100 mA and are directed onto the T/B assembly. The protons are expanded using a raster/expansion system to illuminate a 0.19m by 1.9m beam spot on the front face of a centrally located tungsten neutron source. A surrounding lead blanket produces additional neutrons from scattered high-energy particles. The tungsten neutron source consists of nested, Inconel-718 clad tungsten cylinders assembled in horizontal Inconel-718 tubes. Each tube contains up to 6 cylinders with annular flow channel gaps of 0.102 cm. These horizontal tubes are manifolded into larger diameter vertical inlet and outlet pipes, which provide coolant. The horizontal and vertical tubes make up a structure similar to that of rungs on a ladder. The entire tungsten neutron source consists of 11 such ladders separated into two modules, one containing five ladders and the other six. Ladders are separated by a 0.3 m void region to increase nucleon leakage. The peak thermal-hydraulic conditions in the tungsten neutron source occur in the second ladder from the front. Because tungsten neutron source design has a significant number of parallel flow channels, the limiting thermal-hydraulic parameter is the onset of significant void (OSV) rather than critical heat flux (CHF). A blanket region surrounds the tungsten neutron source. The lateral blanket region is approximately 120 cm thick and 400 cm high. Blanket material consists of lead, {sup 3}He gas, aluminum, and light-water coolant. The blanket region is subdivided into rows based on the local power density in the lead. The rows immediately surrounding the tungsten neutron source receive the highest particle fluxes, and therefore experience the highest power densities. Moving outward from the tungsten neutron source, the lead power density drops steeply. To accommodate this variation, the amount of lead and coolant is tailored to the power density to maximize tritium production.

  2. 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

  3. 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.

  4. Development of pulsed neutron uranium logging instrument

    SciTech Connect (OSTI)

    Wang, Xin-guang; Liu, Dan; Zhang, Feng

    2015-03-15

    This article introduces a development of pulsed neutron uranium logging instrument. By analyzing the temporal distribution of epithermal neutrons generated from the thermal fission of {sup 235}U, we propose a new method with a uranium-bearing index to calculate the uranium content in the formation. An instrument employing a D-T neutron generator and two epithermal neutron detectors has been developed. The logging response is studied using Monte Carlo simulation and experiments in calibration wells. The simulation and experimental results show that the uranium-bearing index is linearly correlated with the uranium content, and the porosity and thermal neutron lifetime of the formation can be acquired simultaneously.

  5. Portable Neutron Sensors for Emergency Response Operations

    SciTech Connect (OSTI)

    ,

    2012-06-24

    This article presents the experimental work performed in the area of neutron detector development at the Remote Sensing LaboratoryAndrews Operations (RSL-AO) sponsored by the U.S. Department of Energy, National Nuclear Security Administration (NNSA) in the last four years. During the 1950s neutron detectors were developed mostly to characterize nuclear reactors where the neutron flux is high. Due to the indirect nature of neutron detection via interaction with other particles, neutron counting and neutron energy measurements have never been as precise as gamma-ray counting measurements and gamma-ray spectroscopy. This indirect nature is intrinsic to all neutron measurement endeavors (except perhaps for neutron spin-related experiments, viz. neutron spin-echo measurements where one obtains ?eV energy resolution). In emergency response situations generally the count rates are low, and neutrons may be scattered around in inhomogeneous intervening materials. It is also true that neutron sensors are most efficient for the lowest energy neutrons, so it is not as easy to detect and count energetic neutrons. Most of the emergency response neutron detectors are offshoots of nuclear device diagnostics tools and special nuclear materials characterization equipment, because that is what is available commercially. These instruments mostly are laboratory equipment, and not field-deployable gear suited for mobile teams. Our goal is to design and prototype field-deployable, ruggedized, lightweight, efficient neutron detectors.

  6. Shifting scintillator neutron detector

    DOE Patents [OSTI]

    Clonts, Lloyd G; Cooper, Ronald G; Crow, Jr., Morris Lowell; Hannah, Bruce W; Hodges, Jason P; Richards, John D; Riedel, Richard A

    2014-03-04

    Provided are sensors and methods for detecting thermal neutrons. Provided is an apparatus having a scintillator for absorbing a neutron, the scintillator having a back side for discharging a scintillation light of a first wavelength in response to the absorbed neutron, an array of wavelength-shifting fibers proximate to the back side of the scintillator for shifting the scintillation light of the first wavelength to light of a second wavelength, the wavelength-shifting fibers being disposed in a two-dimensional pattern and defining a plurality of scattering plane pixels where the wavelength-shifting fibers overlap, a plurality of photomultiplier tubes, in coded optical communication with the wavelength-shifting fibers, for converting the light of the second wavelength to an electronic signal, and a processor for processing the electronic signal to identify one of the plurality of scattering plane pixels as indicative of a position within the scintillator where the neutron was absorbed.

  7. Energy dependence of the /sup 238/U thermal capture cross section. [25 to 450/sup 0/C

    SciTech Connect (OSTI)

    Baumann, N.P.; Owais, M.

    1980-01-01

    Integral activation measurements supported the thermal neutron energy dependence of /sup 238/U assumed in the ENDF/B-IV evaluation. The activation measurements were conducted in a thermally insulated graphite block at the side of the SP Reactor. The block was thermally heated to temperatures up to 450/sup 0/C. In addition to heating, gasolinium filters were used to tailor the neutron spectra incident on the foils. The metallic foils consisted of copper and depleted uranium. Copper served as the 1/v reference. Activation ratios of /sup 238/U to /sup 63/Cu in the tailored spectrum were compared with corresponding ratios in a well thermalized flux at room temperature. The difference in this ratio is strongly dependent on the energy dependence of the /sup 238/U cross section. 8 figures, 1 table.

  8. NEUTRON ABSORPTION AND SHIELDING DEVICE

    DOE Patents [OSTI]

    Axelrad, I.R.

    1960-06-21

    A neutron absorption and shielding device is described which is adapted for mounting in a radiation shielding wall surrounding a radioactive area through which instrumentation leads and the like may safely pass without permitting gamma or neutron radiation to pass to the exterior. The shielding device comprises a container having at least one nonrectilinear tube or passageway means extending therethrough, which is adapted to contain instrumentation leads or the like, a layer of a substance capable of absorbing gamma rays, and a solid resinous composition adapted to attenuate fast-moving neutrons and capture slow- moving or thermal neutrons.

  9. Neutron Thermal Cross Sections, Westcott Factors, Resonance Integrals, Maxwellian Averaged Cross Sections and Astrophysical Reaction Rates Calculated from the ENDF/B-VII.1, JEFF-3.1.2, JENDL-4.0, ROSFOND-2010, CENDL-3.1 and EAF-2010 Evaluated Data Libraries

    SciTech Connect (OSTI)

    Pritychenko, B.; Mughabghab, S.F.

    2012-12-15

    We present calculations of neutron thermal cross sections, Westcott factors, resonance integrals, Maxwellian-averaged cross sections and astrophysical reaction rates for 843 ENDF materials using data from the major evaluated nuclear libraries and European activation file. Extensive analysis of newly-evaluated neutron reaction cross sections, neutron covariances, and improvements in data processing techniques motivated us to calculate nuclear industry and neutron physics quantities, produce s-process Maxwellian-averaged cross sections and astrophysical reaction rates, systematically calculate uncertainties, and provide additional insights on currently available neutron-induced reaction data. Nuclear reaction calculations are discussed and new results are presented. Due to space limitations, the present paper contains only calculated Maxwellian-averaged cross sections and their uncertainties. The complete data sets for all results are published in the Brookhaven National Laboratory report.

  10. Manufacturing techniques studies of ceramics by neutron and ?-ray radiography

    SciTech Connect (OSTI)

    Latini, R. M.; Bellido, A. V. B.; Souza, M. I. S.; Almeida, G. L.

    2014-11-11

    In this study, the aim was to evaluate capabilities and constraints of radiographic imagery using thermal neutrons and gamma-rays as tools to identify the type of technique employed in ceramics manufacturing especially that used in prehistoric Brazilian pottery from Acre state. For this purpose, radiographic images of test objects made with clay of this region using both techniques - palette and rollers - have been acquired with a system comprised of a source of gamma-rays or thermal neutrons and a corresponding X-ray or neutron-sensitive Imaging Plate as detector. For the neutrongraphy samples were exposed to a thermal neutron flux of order of 10{sup 5}n.cm{sup ?2}.s{sup ?1} for 3 minutes at main port of Argonauta research reactor of the Instituto de Engenharia Nuclear - IEN/CNEN. The radiographic images using ?-rays from {sup 165}Dy (95 keV) and {sup 198}Au (412 keV) both produced at this reactor, have been acquired under an exposure time of a couple of hours. After acquisition, images have undergone a treatment to improve their quality through enhancement of their contrast, a procedure involving corrections of the beam divergence, sample shape and averaging of the attenuation map profile. Preliminary results show that difference between manufacturing techniques is better identified by radiography using low energy ?-rays from {sup 165}Dy rather than neutrongraphy or ?-rays from {sup 198}Au. Nevertheless, disregarding the kind of employed radiation, it should be stressed that feasibility to apply the technique is tightly tied to homogeneity of the clay itself and tempers due to their different attenuation.

  11. Variable control of neutron albedo in toroidal fusion devices

    DOE Patents [OSTI]

    Jassby, Daniel L.; Micklich, Bradley J.

    1986-01-01

    An arrangement is provided for controlling neutron albedo in toroidal fusion devices having inboard and outboard vacuum vessel walls for containment of the neutrons of a fusion plasma. Neutron albedo material is disposed immediately adjacent the inboard wall, and is movable, preferably in vertical directions, so as to be brought into and out of neutron modifying communication with the fusion neutrons. Neutron albedo material preferably comprises a liquid form, but may also take pebble, stringer and curtain-like forms. A neutron flux valve, rotatable about a vertical axis is also disclosed.

  12. Method for correcting for isotope burn-in effects in fission neutron dosimeters

    DOE Patents [OSTI]

    Gold, Raymond; McElroy, William N.

    1988-01-01

    A method is described for correcting for effect of isotope burn-in in fission neutron dosimeters. Two quantities are measured in order to quantify the "burn-in" contribution, namely P.sub.Z',A', the amount of (Z', A') isotope that is burned-in, and F.sub.Z', A', the fissions per unit volume produced in the (Z', A') isotope. To measure P.sub.Z', A', two solid state track recorder fission deposits are prepared from the very same material that comprises the fission neutron dosimeter, and the mass and mass density are measured. One of these deposits is exposed along with the fission neutron dosimeter, whereas the second deposit is subsequently used for observation of background. P.sub.Z', A' is then determined by conducting a second irradiation, wherein both the irradiated and unirradiated fission deposits are used in solid state track recorder dosimeters for observation of the absolute number of fissions per unit volume. The difference between the latter determines P.sub.Z', A' since the thermal neutron cross section is known. F.sub.Z', A' is obtained by using a fission neutron dosimeter for this specific isotope, which is exposed along with the original threshold fission neutron dosimeter to experience the same neutron flux-time history at the same location. In order to determine the fissions per unit volume produced in the isotope (Z', A') as it ingrows during the irradiation, B.sub.Z', A', from these observations, the neutron field must generally be either time independent or a separable function of time t and neutron energy E.

  13. FLUX SENSOR EVALUATIONS AT THE ATR CRITICAL FACILITY

    SciTech Connect (OSTI)

    Troy Unruh; Joy Rempe; David Nigg; George Imel; Jason Harris; Eric Bonebrake

    2010-11-01

    The Advanced Test Reactor (ATR) and the ATR Critical (ATRC) facilities lack real-time methods for detecting thermal neutron flux and fission reaction rates for irradiation capsules. Direct measurements of the actual power deposited into a test are now possible without resorting to complicated correction factors. In addition, it is possible to directly measure minor actinide fission reaction rates and to provide time-dependent monitoring of the fission reaction rate or fast/thermal flux during transient testing. A joint Idaho State University /Idaho National Laboratory ATR National Scientific User Facility (ATR NSUF) project was recently initiated to evaluate new real-time state-of-the-art in-pile flux detection sensors. Initially, the project is comparing the accuracy, response time, and long duration performance of French Atomic Energy Commission (CEA)-developed miniature fission chambers, specialized self-powered neutron detectors (SPNDs) by the Argentinean National Energy Commission (CNEA), specially developed commercial SPNDs, and back-to-back fission (BTB) chambers developed by Argonne National Laboratory (ANL). As discussed in this paper, specialized fixturing and software was developed by INL to facilitate these joint ISU/INL evaluations. Calculations were performed by ISU to assess the performance of and reduce uncertainties in flux detection sensors and compare data obtained from these sensors with existing integral methods employed at the ATRC. Ultimately, project results will be used to select the detector that can provide the best online regional ATRC power measurement. It is anticipated that project results may offer the potential to increase the ATRCs current power limit and its ability to perform low-level irradiation experiments. In addition, results from this effort will provide insights about the viability of using these detectors in the ATR. Hence, this effort complements current activities to improve ATR software tools, computational protocols and in-core instrumentation under the ATR Modeling, Simulation and V&V Upgrade initiative, as well as the work to replace nuclear instrumentation under the ATR Life Extension Project (LEP) and provide support to the ATR NSUF.

  14. Monte Carlo Hauser-Feshbach Calculations of Prompt Fission Neutrons...

    Office of Scientific and Technical Information (OSTI)

    Technical Report: Monte Carlo Hauser-Feshbach Calculations of Prompt Fission Neutrons and Gamma Rays: Application to Thermal Neutron-Induced Fission Reactions on U-235 and Pu-239 ...

  15. The temperature dependence of ultra-cold neutron wall losses...

    Office of Scientific and Technical Information (OSTI)

    Title: The temperature dependence of ultra-cold neutron wall losses in material bottles coated with deuterated polystryene Ultra-cold neutrons (UCN) from the LANSCE super-thermal ...

  16. Neutron skins and neutron stars

    SciTech Connect (OSTI)

    Piekarewicz, J.

    2013-11-07

    The neutron-skin thickness of heavy nuclei provides a fundamental link to the equation of state of neutron-rich matter, and hence to the properties of neutron stars. The Lead Radius Experiment ('PREX') at Jefferson Laboratory has recently provided the first model-independence evidence on the existence of a neutron-rich skin in {sup 208}Pb. In this contribution we examine how the increased accuracy in the determination of neutron skins expected from the commissioning of intense polarized electron beams may impact the physics of neutron stars.

  17. Neutron guide

    DOE Patents [OSTI]

    Greene, Geoffrey L.

    1999-01-01

    A neutron guide in which lengths of cylindrical glass tubing have rectangular glass plates properly dimensioned to allow insertion into the cylindrical glass tubing so that a sealed geometrically precise polygonal cross-section is formed in the cylindrical glass tubing. The neutron guide provides easier alignment between adjacent sections than do the neutron guides of the prior art.

  18. Spallation Neutron Source (SNS) | U.S. DOE Office of Science (SC)

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

    Neutron Scattering Facilities » Spallation Neutron Source (SNS) 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 Spallation Neutron Source (SNS) Print Text Size: A A A

  19. Facility for fast neutron irradiation tests of electronics at the ISIS spallation neutron source

    SciTech Connect (OSTI)

    Andreani, C.; Pietropaolo, A.; Salsano, A.; Gorini, G.; Tardocchi, M.; Paccagnella, A.; Gerardin, S.; Frost, C. D.; Ansell, S.; Platt, S. P.

    2008-03-17

    The VESUVIO beam line at the ISIS spallation neutron source was set up for neutron irradiation tests in the neutron energy range above 10 MeV. The neutron flux and energy spectrum were shown, in benchmark activation measurements, to provide a neutron spectrum similar to the ambient one at sea level, but with an enhancement in intensity of a factor of 10{sup 7}. Such conditions are suitable for accelerated testing of electronic components, as was demonstrated here by measurements of soft error rates in recent technology field programable gate arrays.

  20. Neutronic reactor construction

    DOE Patents [OSTI]

    Huston, Norman E.

    1976-07-06

    1. A neutronic reactor comprising a moderator including horizontal layers formed of horizontal rows of graphite blocks, alternate layers of blocks having the rows extending in one direction, the remaining alternate layers having the rows extending transversely to the said one direction, alternate rows of blocks in one set of alternate layers having longitudinal ducts, the moderator further including slotted graphite tubes positioned in the ducts, the reactor further comprising an aluminum coolant tube positioned within the slotted tube in spaced relation thereto, bodies of thermal-neutron-fissionable material, and jackets enclosing the bodies and being formed of a corrosion-resistant material having a low neutron-capture cross section, the bodies and jackets being positioned within the coolant tube so that the jackets are spaced from the coolant tube.

  1. Neutron coincidence detectors employing heterogeneous materials

    DOE Patents [OSTI]

    Czirr, J. Bartley (Mapleton, UT); Jensen, Gary L. (Orem, UT)

    1993-07-27

    A neutron detector relies upon optical separation of different scintillators to measure the total energy and/or number of neutrons from a neutron source. In pulse mode embodiments of the invention, neutrons are detected in a first detector which surrounds the neutron source and in a second detector surrounding the first detector. An electronic circuit insures that only events are measured which correspond to neutrons first detected in the first detector followed by subsequent detection in the second detector. In spectrometer embodiments of the invention, neutrons are thermalized in the second detector which is formed by a scintillator-moderator and neutron energy is measured from the summed signals from the first and second detectors.

  2. Diagnostic of fusion neutrons on JET tokamak using diamond detector

    SciTech Connect (OSTI)

    Nemtsev, G.; Amosov, V.; Marchenko, N.; Meshchaninov, S.; Rodionov, R.; Popovichev, S.; Collaboration: JET EFDA Conbributors

    2014-08-21

    In 2011-2012, an experimental campaign with a significant yield of fusion neutrons was carried out on the JET tokamak. During this campaign the facility was equipped with two diamond detectors based on natural and artificial CVD diamond. These detectors were designed and manufactured in State Research Center of Russian Federation TRINITI. The detectors measure the flux of fast neutrons with energies above 0.2 MeV. They have been installed in the torus hall and the distance from the center of plasma was about 3 m. For some of the JET pulses in this experiment, the neutron flux density corresponded to the operational conditions in collimator channels of ITER Vertical Neutron Camera. The main objective of diamond monitors was the measurement of total fast neutron flux at the detector location and the estimation of the JET total neutron yield. The detectors operate as threshold counters. Additionally a spectrometric measurement channel has been configured that allowed us to distinguish various energy components of the neutron spectrum. In this paper we describe the neutron signal measuring and calibration procedure of the diamond detector. Fluxes of DD and DT neutrons at the detector location were measured. It is shown that the signals of total neutron yield measured by the diamond detector correlate with signals measured by the main JET neutron diagnostic based on fission chambers with high accuracy. This experiment can be considered as a successful test of diamond detectors in ITER-like conditions.

  3. COMPOSITE NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Menke, J.R.

    1963-06-11

    This patent relates to a reactor having a core which comprises an inner active region and an outer active region, each region separately having a k effective less than one and a k infinity greater than one. The inner and outer regions in combination have a k effective at least equal to one and each region contributes substantially to the k effective of the reactor core. The inner region has a low moderator to fuel ratio such that the majority of fissions occurring therein are induced by neutrons having energies greater than thermal. The outer region has a high moderator to fuel ratio such that the majority of fissions occurring therein are induced by thermal neutrons. (AEC)

  4. Hybrid superconducting neutron detectors

    SciTech Connect (OSTI)

    Merlo, V.; Lucci, M.; Ottaviani, I.; Salvato, M.; Cirillo, M.; Scherillo, A.; Celentano, G.; Pietropaolo, A.

    2015-03-16

    A neutron detection concept is presented that is based on superconductive niobium (Nb) strips coated by a boron (B) layer. The working principle of the detector relies on the nuclear reaction, {sup 10}B + n → α + {sup 7}Li, with α and Li ions generating a hot spot on the current-biased Nb strip which in turn induces a superconducting-normal state transition. The latter is recognized as a voltage signal which is the evidence of the incident neutron. The above described detection principle has been experimentally assessed and verified by irradiating the samples with a pulsed neutron beam at the ISIS spallation neutron source (UK). It is found that the boron coated superconducting strips, kept at a temperature T = 8 K and current-biased below the critical current I{sub c}, are driven into the normal state upon thermal neutron irradiation. As a result of the transition, voltage pulses in excess of 40 mV are measured while the bias current can be properly modulated to bring the strip back to the superconducting state, thus resetting the detector. Measurements on the counting rate of the device are presented and the basic physical features of the detector are discussed.

  5. Fast flux locked loop

    DOE Patents [OSTI]

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

    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.

  6. Ultracold neutrons

    SciTech Connect (OSTI)

    Saunders, Alexander

    2015-06-22

    This series of slides describes ultracold neutrons (UCN) and their properties, various UCN sources, and an overview of UCN-based experiments. Numerous diagrams and photographs are included.

  7. Thermo-mechanical and neutron lifetime modeling and design of Be pebbles in the neutron multiplier for the LIFE engine

    SciTech Connect (OSTI)

    DeMange, P; Marian, J; de Caro, M S; Caro, A

    2009-03-16

    Concept designs for the laser-initiated fusion/fission engine (LIFE) include a neutron multiplication blanket containing Be pebbles flowing in a molten salt coolant. These pebbles must be designed to withstand the extreme irradiation and temperature conditions in the blanket to enable a safe and cost-effective operation of LIFE. In this work, we develop design criteria for spherical Be pebbles on the basis of their thermomechanical behavior under continued neutron exposure. We consider the effects of high fluence/fast flux on the elastic, thermal and mechanical properties of nuclear-grade Be. Our results suggest a maximum pebble diameter of 30 mm to avoid tensile failure, coated with an anti-corrosive, high-strength metallic shell to avoid failure by pebble contact. Moreover, we find that the operation temperature must always be kept above 450 C to enable creep to relax the stresses induced by swelling, which we estimate to be at least 16 months if uncoated and up to six years when coated. We identify the sources of uncertainty on the properties used and discuss the advantages of new intermetallic beryllides and their use in LIFE's neutron multiplier. To establish Be-pebble lifetimes with improved confidence, reliable experiments to measure irradiation creep must be performed.

  8. 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.

  9. Neutron tubes

    DOE Patents [OSTI]

    Leung, Ka-Ngo; Lou, Tak Pui; Reijonen, Jani

    2008-03-11

    A neutron tube or generator is based on a RF driven plasma ion source having a quartz or other chamber surrounded by an external RF antenna. A deuterium or mixed deuterium/tritium (or even just a tritium) plasma is generated in the chamber and D or D/T (or T) ions are extracted from the plasma. A neutron generating target is positioned so that the ion beam is incident thereon and loads the target. Incident ions cause D-D or D-T (or T-T) reactions which generate neutrons. Various embodiments differ primarily in size of the chamber and position and shape of the neutron generating target. Some neutron generators are small enough for implantation in the body. The target may be at the end of a catheter-like drift tube. The target may have a tapered or conical surface to increase target surface area.

  10. High-Heat Flux Testing of Irradiated Tungsten based Materials for Fusion Applications using Infrared Plasma Arc Lamps

    SciTech Connect (OSTI)

    Sabau, Adrian S; Ohriner, Evan Keith; Kiggans Jr, James O; Schaich, Charles Ross; Ueda, Yoshio; Harper, David C; Katoh, Yutai; Snead, Lance Lewis; Byun, Thak Sang

    2014-01-01

    Testing of advanced materials and component mock-ups under prototypical fusion high-heat flux conditions, while historically a mainstay of fusion research has proved challenging, especially for irradiated materials. A new high-heat flux testing facility based on water-wall Plasma Arc Lamps (PALs) is now being used for materials and small component testing. Two PAL systems, utilizing a 12,000 C plasma arc contained in a quartz tube cooled by a spiral water flow over the inside tube surface, are currently in use. The first PAL system provides a maximum incident heat flux of 4.2 MW/m2 over an area of 9x12 cm2. The second PAL available at ORNL provides a maximum incident heat flux of 27 MW/m2 over an area of 1x10 cm2. The absorbed heat fluxes into a tungsten target for the two PALs are approximately 1.97 and 12.7 MW/m2, respectively. This paper will present the overall design of the new PAL facilities as well as the design and implementation of the Irradiated Material Target Station (IMTS). The IMTS is primarily designed for testing the effects of heat flux or thermal cycling on material coupons of interested, such as those for plasma facing components. Moreover, IMTS designs are underway to extend the testing of small mock-ups for assessing the combined heating and thermomechanical effects of cooled, irradiated components. For the testing of material coupons , the specimens are placed in a shallow recess within the molybdenum holder that is attached to a water-cooled copper alloy rod. As the measurement of the specimen temperature for PAL is historically challenging since traditional approaches of temperature measurement cannot be employed due to the infrared heating and proximity of the PAL reflector to the specimen that does not allow a direct line of site, experiments for temperature calibration are presented. Finally, results for the high-heat flux testing of tungsten-based materials using the PAL are presented. As a demonstration of the system, results will be shown of thermal fatigue and high-heat flux testing of tungsten coupon specimens that were neutron irradiated in the HFIR reactor to neutron dose consistent to ITER lifetime.

  11. High-Heat Flux Testing of Irradiated Tungsten based Materials for Fusion Applications using Infrared Plasma Arc Lamps

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

    Sabau, Adrian S; Ohriner, Evan Keith; Kiggans, Jr, James O; Schaich, Charles Ross; Ueda, Yoshio; Harper, David C; Katoh, Yutai; Snead, Lance Lewis; Byun, Thak Sang

    2014-01-01

    Testing of advanced materials and component mock-ups under prototypical fusion high-heat flux conditions, while historically a mainstay of fusion research has proved challenging, especially for irradiated materials. A new high-heat flux testing facility based on water-wall Plasma Arc Lamps (PALs) is now being used for materials and small component testing. Two PAL systems, utilizing a 12,000 C plasma arc contained in a quartz tube cooled by a spiral water flow over the inside tube surface, are currently in use. The first PAL system provides a maximum incident heat flux of 4.2 MW/m2 over an area of 9x12 cm2. The secondmore » PAL available at ORNL provides a maximum incident heat flux of 27 MW/m2 over an area of 1x10 cm2. The absorbed heat fluxes into a tungsten target for the two PALs are approximately 1.97 and 12.7 MW/m2, respectively. This paper will present the overall design of the new PAL facilities as well as the design and implementation of the Irradiated Material Target Station (IMTS). The IMTS is primarily designed for testing the effects of heat flux or thermal cycling on material coupons of interested, such as those for plasma facing components. Moreover, IMTS designs are underway to extend the testing of small mock-ups for assessing the combined heating and thermomechanical effects of cooled, irradiated components. For the testing of material coupons , the specimens are placed in a shallow recess within the molybdenum holder that is attached to a water-cooled copper alloy rod. As the measurement of the specimen temperature for PAL is historically challenging since traditional approaches of temperature measurement cannot be employed due to the infrared heating and proximity of the PAL reflector to the specimen that does not allow a direct line of site, experiments for temperature calibration are presented. Finally, results for the high-heat flux testing of tungsten-based materials using the PAL are presented. As a demonstration of the system, results will be shown of thermal fatigue and high-heat flux testing of tungsten coupon specimens that were neutron irradiated in the HFIR reactor to neutron dose consistent to ITER lifetime.« less

  12. 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.

  13. 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.

  14. Tracking heat flux sensors for concentrating solar applications

    DOE Patents [OSTI]

    Andraka, Charles E; Diver, Jr., Richard B

    2013-06-11

    Innovative tracking heat flux sensors located at or near the solar collector's focus for centering the concentrated image on a receiver assembly. With flux sensors mounted near a receiver's aperture, the flux gradient near the focus of a dish or trough collector can be used to precisely position the focused solar flux on the receiver. The heat flux sensors comprise two closely-coupled thermocouple junctions with opposing electrical polarity that are separated by a thermal resistor. This arrangement creates an electrical signal proportional to heat flux intensity, and largely independent of temperature. The sensors are thermally grounded to allow a temperature difference to develop across the thermal resistor, and are cooled by a heat sink to maintain an acceptable operating temperature.

  15. Plasma momentum meter for momentum flux measurements

    DOE Patents [OSTI]

    Zonca, Fulvio; Cohen, Samuel A.; Bennett, Timothy; Timberlake, John R.

    1993-01-01

    Invention comprises an instrument in which momentum flux onto a biasable target plate is transferred via a suspended quartz tube onto a sensitive force transducer--a capacitance-type pressure gauge. The transducer is protected from thermal damage, arcing and sputtering, and materials used in the target and pendulum are electrically insulating, rigid even at elevated temperatures, and have low thermal conductivity. The instrument enables measurement of small forces (10.sup.-5 to 10.sup.3 N) accompanied by high heat fluxes which are transmitted by energetic particles with 10's of eV of kinetic energy in a intense magnetic field and pulsed plasma environment.

  16. Neutron-driven gamma-ray laser

    DOE Patents [OSTI]

    Bowman, Charles D.

    1990-01-01

    A lasing cylinder emits laser radiation at a gamma-ray wavelength of 0.87 .ANG. when subjected to an intense neutron flux of about 400 eV neutrons. A 250 .ANG. thick layer of Be is provided between two layers of 100 .ANG. thick layer of .sup.57 Co and these layers are supported on a foil substrate. The coated foil is coiled to form the lasing cylinder. Under the neutron flux .sup.57 Co becomes .sup.58 Co by neutron absorption. The .sup.58 Co then decays to .sup.57 Fe by 1.6 MeV proton emission. .sup.57 Fe then transitions by mesne decay to a population inversion for lasing action at 14.4 keV. Recoil from the proton emission separates the .sup.57 Fe from the .sup.57 Co and into the Be, where Mossbauer emission occurs at a gamma-ray wavelength.

  17. Covariance Evaluation Methodology for Neutron Cross Sections

    SciTech Connect (OSTI)

    Herman,M.; Arcilla, R.; Mattoon, C.M.; Mughabghab, S.F.; Oblozinsky, P.; Pigni, M.; Pritychenko, b.; Songzoni, A.A.

    2008-09-01

    We present the NNDC-BNL methodology for estimating neutron cross section covariances in thermal, resolved resonance, unresolved resonance and fast neutron regions. The three key elements of the methodology are Atlas of Neutron Resonances, nuclear reaction code EMPIRE, and the Bayesian code implementing Kalman filter concept. The covariance data processing, visualization and distribution capabilities are integral components of the NNDC methodology. We illustrate its application on examples including relatively detailed evaluation of covariances for two individual nuclei and massive production of simple covariance estimates for 307 materials. Certain peculiarities regarding evaluation of covariances for resolved resonances and the consistency between resonance parameter uncertainties and thermal cross section uncertainties are also discussed.

  18. Construction of the WSU Epithermal Neutron Filter

    SciTech Connect (OSTI)

    Venhuizen, James Robert; Nigg, David Waler; Tripard, G.

    2002-09-01

    Moderating material has been installed in the original thermal-neutron filter region of the Washington State University (WSU) TRIGA™ type reactor to produce an epithermal-neutron beam. Attention has been focused upon the development of a convenient, local, epithermal-neutron beam facility at WSU for collaborative Idaho National Engineering and Environmental Laboratory (INEEL)/WSU boron neutron capture therapy (BNCT) preclinical research and boronated pharmaceutical screening in cell and animal models. The design of the new facility was performed in a collaborative effort1,2 of WSU and INEEL scientists. This paper summarizes the physical assembly of this filter.

  19. 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/.

  20. ATR LEU Monolithic Foil-Type Fuel with Integral Cladding Burnable Absorber Neutronics Performance Evaluation

    SciTech Connect (OSTI)

    Gray Chang

    2012-03-01

    The Advanced Test Reactor (ATR), currently operating in the United States, is used for material testing at very high neutron fluxes. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting HEU driven reactor cores to low-enriched uranium (LEU) cores. The burnable absorber - 10B, was added in the inner and outer plates to reduce the initial excess reactivity, and to improve the peak ratio of the inner/outer heat flux. The present work investigates the LEU Monolithic foil-type fuel with 10B Integral Cladding Burnable Absorber (ICBA) design and evaluates the subsequent neutronics operating effects of this proposed fuel designs. The proposed LEU fuel specification in this work is directly related to both the RERTR LEU Development Program and the Advanced Test Reactor (ATR) LEU Conversion Project at Idaho National Laboratory (INL).

  1. 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)

  2. Method and apparatus for detecting neutrons

    DOE Patents [OSTI]

    Perkins, R.W.; Reeder, P.L.; Wogman, N.A.; Warner, R.A.; Brite, D.W.; Richey, W.C.; Goldman, D.S.

    1997-10-21

    The instant invention is a method for making and using an apparatus for detecting neutrons. Scintillating optical fibers are fabricated by melting SiO{sub 2} with a thermal neutron capturing substance and a scintillating material in a reducing atmosphere. The melt is then drawn into fibers in an anoxic atmosphere. The fibers may then be coated and used directly in a neutron detection apparatus, or assembled into a geometrical array in a second, hydrogen-rich, scintillating material such as a polymer. Photons generated by interaction with thermal neutrons are trapped within the coated fibers and are directed to photoelectric converters. A measurable electronic signal is generated for each thermal neutron interaction within the fiber. These electronic signals are then manipulated, stored, and interpreted by normal methods to infer the quality and quantity of incident radiation. When the fibers are arranged in an array within a second scintillating material, photons generated by kinetic neutrons interacting with the second scintillating material and photons generated by thermal neutron capture within the fiber can both be directed to photoelectric converters. These electronic signals are then manipulated, stored, and interpreted by normal methods to infer the quality and quantity of incident radiation. 5 figs.

  3. Method and apparatus for detecting neutrons

    DOE Patents [OSTI]

    Perkins, Richard W.; Reeder, Paul L.; Wogman, Ned A.; Warner, Ray A.; Brite, Daniel W.; Richey, Wayne C.; Goldman, Don S.

    1997-01-01

    The instant invention is a method for making and using an apparatus for detecting neutrons. Scintillating optical fibers are fabricated by melting SiO.sub.2 with a thermal neutron capturing substance and a scintillating material in a reducing atmosphere. The melt is then drawn into fibers in an anoxic atmosphere. The fibers may then be coated and used directly in a neutron detection apparatus, or assembled into a geometrical array in a second, hydrogen-rich, scintillating material such as a polymer. Photons generated by interaction with thermal neutrons are trapped within the coated fibers and are directed to photoelectric converters. A measurable electronic signal is generated for each thermal neutron interaction within the fiber. These electronic signals are then manipulated, stored, and interpreted by normal methods to infer the quality and quantity of incident radiation. When the fibers are arranged in an array within a second scintillating material, photons generated by kinetic neutrons interacting with the second scintillating material and photons generated by thermal neutron capture within the fiber can both be directed to photoelectric converters. These electronic signals are then manipulated, stored, and interpreted by normal methods to infer the quality and quantity of incident radiation.

  4. Neutron range spectrometer

    DOE Patents [OSTI]

    Manglos, S.H.

    1988-03-10

    A neutron range spectrometer and method for determining the neutron energy spectrum of a neutron emitting source are disclosed. Neutrons from the source are colliminated along a collimation axis and a position sensitive neutron counter is disposed in the path of the collimated neutron beam. The counter determines positions along the collimation axis of interactions between the neutrons in the neutron beam and a neutron-absorbing material in the counter. From the interaction positions, a computer analyzes the data and determines the neutron energy spectrum of the neutron beam. The counter is preferably shielded and a suitable neutron-absorbing material is He-3. 1 fig.

  5. ARM - Measurement - Methane flux

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

    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...

  6. Characterization of a Thermo Scientific D711 D-T Neutron Generator Located in a Low-Scatter Facility

    SciTech Connect (OSTI)

    Hayes, John W.; Finn, Erin C.; Greenwood, Lawrence R.; Wittman, Richard S.

    2014-03-21

    Pacific Northwest National Laboratory (PNNL) purchased and installed a D711 D-T neutron generator (D-T) from Thermo Scientific in August 2011. The D-T nominally produces 14 MeV neutrons which are important for research in matters of national security. Fast neutrons provide the capability of harnessing threshold reactions for the production of rare isotopes, which are of interest to radiochemistry groups at PNNL concerned with validating radioanalytical techniques for the separation of these isotopes. Rare fission product isotopes from fast fission of 235U, 238U, and 239Pu are also desired to further develop these techniques. Experiments with 14 MeV neutrons are also of interest because nuclear data for fast fission has not been researched as extensively as it has been for thermal fission. Analyses of these applications require first that the source spectrum be well characterized. Neutron fluences in Fe, Ni, Al, In, and Au were measured in 21 locations near the generator head. STAYSL PNNL and MCNP codes were used to produce flux spectra based on experimental fluences.

  7. Neutron scattering study of underdoped Ba1-xKxFe₂As₂ (x=0.09 and 0.17) self-flux-grown single crystals and the universality of the tricritical point

    SciTech Connect (OSTI)

    Rotundu, C. R.; Tian, W.; Rule, K. C.; Forrest, T. R.; Zhao, J.; Zarestky, J. L.; Birgeneau, R. J.

    2012-04-04

    We present a combination of elastic neutron scattering measurements in zero and 14.5 T and magnetization measurements in zero and 14 T on underdoped superconducting Ba1-xKxFe₂As₂ (x=0.17), and the same measurements in zero field on a nonsuperconducting crystal with x=0.09. The data suggest that the underdoped materials may not be electronic phase separated but rather have slightly inhomogeneous potassium doping. The temperature dependence of the magnetic order parameter below the transition of the sample with x=0.09 is more gradual than that for the case of the undoped BaFe₂As₂, suggesting that this doping may be in the vicinity of a tricritical point. We advance therefore the hypothesis that the tricritical point is a common feature of all superconducting 122s. For the x=0.17 sample, while Tc is suppressed from ≈17 to ≈8 K by a magnetic field of 14 T, the intensity of the magnetic Bragg peaks (1 0 3) at 1.2 K is enhanced by 10%, showing competition of superconductivity and antiferromagnetism. The intensity of the magnetic Bragg peaks (1 0 3) in the (Tc, TN) temperature interval remain practically unchanged in 14.5 T within a 10% statistical error. The present results are discussed in the context of the existing literature.

  8. Neutron scattering study of underdoped Ba1-xKxFe₂As₂ (x=0.09 and 0.17) self-flux-grown single crystals and the universality of the tricritical point

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

    Rotundu, C. R.; Tian, W.; Rule, K. C.; Forrest, T. R.; Zhao, J.; Zarestky, J. L.; Birgeneau, R. J.

    2012-04-04

    We present a combination of elastic neutron scattering measurements in zero and 14.5 T and magnetization measurements in zero and 14 T on underdoped superconducting Ba1-xKxFe₂As₂ (x=0.17), and the same measurements in zero field on a nonsuperconducting crystal with x=0.09. The data suggest that the underdoped materials may not be electronic phase separated but rather have slightly inhomogeneous potassium doping. The temperature dependence of the magnetic order parameter below the transition of the sample with x=0.09 is more gradual than that for the case of the undoped BaFe₂As₂, suggesting that this doping may be in the vicinity of a tricriticalmore » point. We advance therefore the hypothesis that the tricritical point is a common feature of all superconducting 122s. For the x=0.17 sample, while Tc is suppressed from ≈17 to ≈8 K by a magnetic field of 14 T, the intensity of the magnetic Bragg peaks (1 0 3) at 1.2 K is enhanced by 10%, showing competition of superconductivity and antiferromagnetism. The intensity of the magnetic Bragg peaks (1 0 3) in the (Tc, TN) temperature interval remain practically unchanged in 14.5 T within a 10% statistical error. The present results are discussed in the context of the existing literature.« less

  9. Heat flux viscosity in collisional magnetized plasmas

    SciTech Connect (OSTI)

    Liu, C.; Fox, W.; Bhattacharjee, A.

    2015-05-15

    Momentum transport in collisional magnetized plasmas due to gradients in the heat flux, a “heat flux viscosity,” is demonstrated. Even though no net particle flux is associated with a heat flux, in a plasma there can still be momentum transport owing to the velocity dependence of the Coulomb collision frequency, analogous to the thermal force. This heat-flux viscosity may play an important role in numerous plasma environments, in particular, in strongly driven high-energy-density plasma, where strong heat flux can dominate over ordinary plasma flows. The heat flux viscosity can influence the dynamics of the magnetic field in plasmas through the generalized Ohm's law and may therefore play an important role as a dissipation mechanism allowing magnetic field line reconnection. The heat flux viscosity is calculated directly using the finite-difference method of Epperlein and Haines [Phys. Fluids 29, 1029 (1986)], which is shown to be more accurate than Braginskii's method [S. I. Braginskii, Rev. Plasma Phys. 1, 205 (1965)], and confirmed with one-dimensional collisional particle-in-cell simulations. The resulting transport coefficients are tabulated for ease of application.

  10. Method and apparatus for determination of temperature, neutron absorption cross section and neutron moderating power

    DOE Patents [OSTI]

    Vagelatos, Nicholas; Steinman, Donald K.; John, Joseph; Young, Jack C.

    1981-01-01

    A nuclear method and apparatus determines the temperature of a medium by injecting fast neutrons into the medium and detecting returning slow neutrons in three first energy ranges by producing three respective detection signals. The detection signals are combined to produce three derived indicia each systematically related to the population of slow neutrons returning from the medium in a respective one of three second energy ranges, specifically exclusively epithermal neutrons, exclusively substantially all thermal neutrons and exclusively a portion of the thermal neutron spectrum. The derived indicia are compared with calibration indicia similarly systematically related to the population of slow neutrons in the same three second energy ranges returning from similarly irradiated calibration media for which the relationships temperature, neutron absorption cross section and neutron moderating power to such calibration indicia are known. The comparison indicates the temperature at which the calibration indicia correspond to the derived indicia and consequently the temperature of the medium. The neutron absorption cross section and moderating power of the medium can be identified at the same time.

  11. Variable control of neutron albedo in toroidal fusion devices

    DOE Patents [OSTI]

    Jassby, D.L.; Micklich, B.J.

    1983-06-01

    This invention pertains to methods of controlling in the steady state, neutron albedo in toroidal fusion devices, and in particular, to methods of controlling the flux and energy distribution of collided neutrons which are incident on an outboard wall of a toroidal fusion device.

  12. Cadmium Depletion Impacts on Hardening Neutron6 Spectrum for Advanced Fuel Testing in ATR

    SciTech Connect (OSTI)

    Gray S. Chang

    2011-05-01

    For transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products effectively is in a fast neutron spectrum reactor. In the absence of a fast spectrum test reactor in the United States of America (USA), initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. A test region is achieved with a Cadmium (Cd) filter which can harden the neutron spectrum to a spectrum similar (although still somewhat softer) to that of the liquid metal fast breeder reactor (LMFBR). A fuel test loop with a Cd-filter has been installed within the East Flux Trap (EFT) of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). A detailed comparison analyses between the cadmium (Cd) filter hardened neutron spectrum in the ATR and the LMFBR fast neutron spectrum have been performed using MCWO. MCWO is a set of scripting tools that are used to couple the Monte Carlo transport code MCNP with the isotope depletion and buildup code ORIGEN-2.2. The MCWO-calculated results indicate that the Cd-filter can effectively flatten the Rim-Effect and reduce the linear heat rate (LHGR) to meet the advanced fuel testing project requirements at the beginning of irradiation (BOI). However, the filtering characteristics of Cd as a strong absorber quickly depletes over time, and the Cd-filter must be replaced for every two typical operating cycles within the EFT of the ATR. The designed Cd-filter can effectively depress the LHGR in experimental fuels and harden the neutron spectrum enough to adequately flatten the Rim Effect in the test region.

  13. Conducting Polymers for Neutron Detection

    SciTech Connect (OSTI)

    Kimblin, Clare; Miller, Kirk; Vogel, Bob; Quam, Bill; McHugh, Harry; Anthony, Glen; Mike, Grover

    2007-12-01

    Conjugated polymers have emerged as an attractive technology for large-area electronic applications. As organic semiconductors, they can be used to make large-area arrays of diodes or transistors using fabrication techniques developed for polymer coatings, such as spraying and screen-printing. We have demonstrated both neutron and alpha detection using diodes made from conjugated polymers and have done preliminary work to integrate a boron carbide layer into the conventional polymer device structure to capture thermal neutrons. The polymer devices appear to be insensitive to gamma rays, due to their small physical thickness and low atomic number.

  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. 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 ...

  17. Neutron Generators for Spent Fuel Assay

    SciTech Connect (OSTI)

    Ludewigt, Bernhard A

    2010-12-30

    The Next Generation Safeguards Initiative (NGSI) of the U.S. DOE has initiated a multi-lab/university collaboration to quantify the plutonium (Pu) mass in, and detect the diversion of pins from, spent nuclear fuel (SNF) assemblies with non-destructive assay (NDA). The 14 NDA techniques being studied include several that require an external neutron source: Delayed Neutrons (DN), Differential Die-Away (DDA), Delayed Gammas (DG), and Lead Slowing-Down Spectroscopy (LSDS). This report provides a survey of currently available neutron sources and their underlying technology that may be suitable for NDA of SNF assemblies. The neutron sources considered here fall into two broad categories. The term 'neutron generator' is commonly used for sealed devices that operate at relatively low acceleration voltages of less than 150 kV. Systems that employ an acceleration structure to produce ion beam energies from hundreds of keV to several MeV, and that are pumped down to vacuum during operation, rather than being sealed units, are usually referred to as 'accelerator-driven neutron sources.' Currently available neutron sources and future options are evaluated within the parameter space of the neutron generator/source requirements as currently understood and summarized in section 2. Applicable neutron source technologies are described in section 3. Commercially available neutron generators and other source options that could be made available in the near future with some further development and customization are discussed in sections 4 and 5, respectively. The pros and cons of the various options and possible ways forward are discussed in section 6. Selection of the best approach must take a number of parameters into account including cost, size, lifetime, and power consumption, as well as neutron flux, neutron energy spectrum, and pulse structure that satisfy the requirements of the NDA instrument to be built.

  18. NEUTRONIC REACTOR CORE

    DOE Patents [OSTI]

    Thomson, W.B.; Corbin, A. Jr.

    1961-07-18

    An improved core for a gas-cooled power reactor which admits gas coolant at high temperatures while affording strong integral supporting structure and efficient moderation of neutrons is described. The multiplicities of fuel elements constituting the critical amassment of fissionable material are supported and confined by a matrix of metallic structure which is interspersed therebetween. Thermal insulation is interposed between substantially all of the metallic matrix and the fuel elements; the insulation then defines the principal conduit system for conducting the coolant gas in heat-transfer relationship with the fuel elements. The metallic matrix itseif comprises a system of ducts through which an externally-cooled hydrogeneous liquid, such as water, is circulated to serve as the principal neutron moderant for the core and conjointly as the principal coolant for the insulated metallic structure. In this way, use of substantially neutron transparent metals, such as aluminum, becomes possible for the supporting structure, despite the high temperatures of the proximate gas. The Aircraft Nuclear Propulsion program's "R-1" reactor design is a preferred embodiment.

  19. 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.

  20. 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. Consequently, ANL and ILL have collaborated to investigate alternative designs. A promising candidate design has been selected and studied, increasing the total amount of fuel without changing the external plate dimensions by relocating the burnable poison. In this way, changes required in the fuel element are reasonably small. With this new design, neutronics analyses have shown that performance could be maintained at a high level: 2 day decrease of cycle length (to 47.5 days at 58.3 MW) and 1-2% decrease of brightness in the cold and hot sources in comparison to the current typical operation. In addition, studies have shown that the thermal-hydraulic and shutdown margins for the proposed LEU design would satisfy technical specifications.

  1. Materials Selection for the HFIR Cold Neutron Source

    SciTech Connect (OSTI)

    Farrell, K.

    2001-08-24

    In year 2002 the High Flux Isotope Reactor (HFIR) will be fitted with a source of cold neutrons to upgrade and expand its existing neutron scattering facilities. The in-reactor components of the new source consist of a moderator vessel containing supercritical hydrogen gas moderator at a temperature of 20K and pressure of 15 bar, and a surrounding vacuum vessel. They will be installed in an enlarged beam tube located at the site of the present horizontal beam tube, HB-4; which terminates within the reactor's beryllium reflector. These components must withstand exceptional service conditions. This report describes the reasons and factors underlying the choice of 6061-T6 aluminum alloy for construction of the in-reactor components. The overwhelming considerations are the need to minimize generation of nuclear heat and to remove that heat through the flowing moderator, and to achieve a minimum service life of about 8 years coincident with the replacement schedule for the beryllium reflector. 6061-T6 aluminum alloy offers the best combination of low nuclear heating, high thermal conductivity, good fabricability, compatibility with hydrogen, superior cryogenic properties, and a well-established history of satisfactory performance in nuclear environments. These features are documented herein. An assessment is given of the expected performance of each component of the cold source.

  2. PHISICS multi-group transport neutronic capabilities for RELAP5

    SciTech Connect (OSTI)

    Epiney, A.; Rabiti, C.; Alfonsi, A.; Wang, Y.; Cogliati, J.; Strydom, G.

    2012-07-01

    PHISICS is a neutronic code system currently under development at INL. Its goal is to provide state of the art simulation capability to reactor designers. This paper reports on the effort of coupling this package to the thermal hydraulic system code RELAP5. This will enable full prismatic core and system modeling and the possibility to model coupled (thermal-hydraulics and neutronics) problems with more options for 3D neutron kinetics, compared to the existing diffusion theory neutron kinetics module in RELAP5 (NESTLE). The paper describes the capabilities of the coupling and illustrates them with a set of sample problems. (authors)

  3. Modeling Advanced Neutron Source reactor station blackout accident using RELAP5

    SciTech Connect (OSTI)

    Chen, N.C.J. (Oak Ridge National Lab., TN (USA)); Fletcher, C.D. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1990-01-01

    The Advanced Neutron Source (ANS) system model using RELAP5 has been developed to perform loss-of-coolant accident (LOCA) and non-LOCA transients as safety-related input for early design considerations. The transients studies include LOCA, station blackout, and reactivity insertion accidents. The small-, medium-, and large-break LOCA results were presented and documented. This paper will focus on the station blackout scenario. The station blackout analyses have concentrated on thermal-hydraulic system response with and without accumulators. Five transient calculations were performed to characterize system performance using various numbers and sizes of accumulators at several key sites. The main findings will be discussed with recommendations for conceptual design considerations. ANS is a state-of-the-art research reactor to be built and operated at high heat flux, high mass flux, and high coolant subcooling. To accommodate these features, three ANS-specific changes were made in the RELAP5 code by adding: the Petukhov heat transfer correlation for single-phase forced convection in the thin coolant channel; the Gambill additive method with the Weatherhead wall superheat for the critical heat flux; and the Griffith drift flux model for the interfacial drag in the slug flow regime. 7 refs., 6 figs., 1 tab.

  4. SU-E-T-557: Measuring Neutron Activation of Cardiac Devices Irradiated During Proton Therapy Using Indium Foils

    SciTech Connect (OSTI)

    Avery, S; Christodouleas, J; Delaney, K; Diffenderfer, E; Brown, K

    2014-06-01

    Purpose: Measuring Neutron Activation of Cardiac devices Irradiated during Proton Therapy using Indium Foils Methods: The foils had dimensions of 25mm x 25mm x 1mm. After being activated, the foils were placed in a Canberra Industries well chamber utilizing a NaI(Tl) scintillation detector. The resulting gamma spectrum was acquired and analyzed using Genie 2000 spectroscopy software. One activation foil was placed over the upper, left chest of RANDO where a pacemaker would be. The rest of the foils were placed over the midline of the patient at different distances, providing a spatial distribution over the phantom. Using lasers and BBs to align the patient, 200 MU square fields were delivered to various treatment sites: the brain, the pancreas, and the prostate. Each field was shot at least a day apart, giving more than enough time for activity of the foil to decay (t1=2 = 54.12 min). Results: The net counts (minus background) of the three aforementioned peaks were used for our measurements. These counts were adjusted to account for detector efficiency, relative photon yields from decay, and the natural abundance of 115-In. The average neutron flux for the closed multi-leaf collimator irradiation was measured to be 1.62 x 106 - 0.18 x 106 cm2 s-1. An order of magnitude estimate of the flux for neutrons up to 1 keV from Diffenderfer et al. gives 3 x 106 cm2 s-1 which does agree on the order of magnitude. Conclusion: Lower energy neutrons have higher interaction cross-sections and are more likely to damage pacemakers. The thermal/slow neutron component may be enough to estimate the overall risk. The true test of the applicability of activation foils is whether or not measurements are capable of predicting cardiac device malfunction. For that, additional studies are needed to provide clinical evidence one way or the other.

  5. Neutron reflecting supermirror structure

    DOE Patents [OSTI]

    Wood, James L. (Drayton Plains, MI)

    1992-01-01

    An improved neutron reflecting supermirror structure comprising a plurality of stacked sets of bilayers of neutron reflecting materials. The improved neutron reflecting supermirror structure is adapted to provide extremely good performance at high incidence angles, i.e. up to four time the critical angle of standard neutron mirror structures. The reflection of neutrons striking the supermirror structure at a high critical angle provides enhanced neutron throughput, and hence more efficient and economical use of neutron sources.

  6. Neutron reflecting supermirror structure

    DOE Patents [OSTI]

    Wood, J.L.

    1992-12-01

    An improved neutron reflecting supermirror structure comprising a plurality of stacked sets of bilayers of neutron reflecting materials. The improved neutron reflecting supermirror structure is adapted to provide extremely good performance at high incidence angles, i.e. up to four time the critical angle of standard neutron mirror structures. The reflection of neutrons striking the supermirror structure at a high critical angle provides enhanced neutron throughput, and hence more efficient and economical use of neutron sources. 2 figs.

  7. Asymptotic, multigroup flux reconstruction and consistent discontinuity factors

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

    Trahan, Travis J.; Larsen, Edward W.

    2015-05-12

    Recent theoretical work has led to an asymptotically derived expression for reconstructing the neutron flux from lattice functions and multigroup diffusion solutions. The leading-order asymptotic term is the standard expression for flux reconstruction, i.e., it is the product of a shape function, obtained through a lattice calculation, and the multigroup diffusion solution. The first-order asymptotic correction term is significant only where the gradient of the diffusion solution is not small. Inclusion of this first-order correction term can significantly improve the accuracy of the reconstructed flux. One may define discontinuity factors (DFs) to make certain angular moments of the reconstructed fluxmore » continuous across interfaces between assemblies in 1-D. Indeed, the standard assembly discontinuity factors make the zeroth moment (scalar flux) of the reconstructed flux continuous. The inclusion of the correction term in the flux reconstruction provides an additional degree of freedom that can be used to make two angular moments of the reconstructed flux continuous across interfaces by using current DFs in addition to flux DFs. Thus, numerical results demonstrate that using flux and current DFs together can be more accurate than using only flux DFs, and that making the second angular moment continuous can be more accurate than making the zeroth moment continuous.« less

  8. 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.

  9. Neutron Scattering Facilities | U.S. DOE Office of Science (SC)

    Office of Science (SC) Website

    Facilities 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 User Facilities Neutron Scattering Facilities Print Text Size: A A A FeedbackShare Page This activity supports the operation of two neutron scattering

  10. Safety control circuit for a neutronic reactor

    DOE Patents [OSTI]

    Ellsworth, Howard C. (Richland, WA)

    2004-04-27

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

  11. 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.

  12. Thermal Analysis of a Uranium Silicide Miniplate Irradiation Experiment

    SciTech Connect (OSTI)

    Donna Post Guillen

    2009-09-01

    This paper outlines the thermal analysis for the irradiation of high density uranium-silicide (U3Si2 dispersed in an aluminum matrix and clad in aluminum) booster fuel for a Boosted Fast Flux Loop designed to provide fast neutron flux test capability in the ATR. The purpose of this experiment (designated as Gas Test Loop-1 [GTL-1]) is two-fold: (1) to assess the adequacy of the U3Si2/Al dispersion fuel and the aluminum alloy 6061 cladding, and (2) to verify stability of the fuel cladding boehmite pre-treatment at nominal power levels in the 430 to 615 W/cm2 (2.63 to 3.76 Btu/s•in2) range. The GTL-1 experiment relies on a difficult balance between achieving a high heat flux, yet keeping fuel centerline temperature below a specified maximum value throughout an entire operating cycle of the reactor. A detailed finite element model was constructed to calculate temperatures and heat flux levels and to reveal which experiment parameters place constraints on reactor operations. Analyses were performed to determine the bounding lobe power level at which the experiment could be safely irradiated, yet still provide meaningful data under nominal operating conditions. Then, simulations were conducted for nominal and bounding lobe power levels under steady-state and transient conditions with the experiment in the reactor. Reactivity changes due to a loss of commercial power with pump coast-down to emergency flow or a standard in-pile tube pump discharge break were evaluated. The time after shutdown for which the experiment can be adequately cooled by natural convection cooling was determined using a system thermal hydraulic model. An analysis was performed to establish the required in-reactor cooling time prior to removal of the experiment from the reactor. The inclusion of machining tolerances in the numerical model has a large effect on heat transfer.

  13. Non-Destructive Spent Fuel Characterization with Semi-Conducting Gallium Arsinde Neutron Imaging Arrays

    SciTech Connect (OSTI)

    Douglas S. McGregor; Holly K. Gersch; Jeffrey D. Sanders; John C. Lee; Mark D. Hammig; Michael R. Hartman; Yong Hong Yang; Raymond T. Klann; Brian Van Der Elzen; John T. Lindsay; Philip A. Simpson

    2002-01-30

    High resistivity bulk grown GaAs has been used to produce thermal neutron imaging devices for use in neutron radiography and characterizing burnup in spent fuel. The basic scheme utilizes a portable Sb/Be source for monoenergetic (24 keV) neutron radiation source coupled to an Fe filter with a radiation hard B-coated pixellated GaAs detector array as the primary neutron detector. The coated neutron detectors have been tested for efficiency and radiation hardness in order to determine their fitness for the harsh environments imposed by spent fuel. Theoretical and experimental results are presented, showing detector radiation hardness, expected detection efficiency and the spatial resolution from such a scheme. A variety of advanced neutron detector designs have been explored, with experimental results achieving 13% thermal neutron detection efficiency while projecting the possibility of over 30% thermal neutron detection efficiency.

  14. Identifying and bounding uncertainties in nuclear reactor thermal power calculations

    SciTech Connect (OSTI)

    Phillips, J.; Hauser, E.; Estrada, H.

    2012-07-01

    Determination of the thermal power generated in the reactor core of a nuclear power plant is a critical element in the safe and economic operation of the plant. Direct measurement of the reactor core thermal power is made using neutron flux instrumentation; however, this instrumentation requires frequent calibration due to changes in the measured flux caused by fuel burn-up, flux pattern changes, and instrumentation drift. To calibrate the nuclear instruments, steam plant calorimetry, a process of performing a heat balance around the nuclear steam supply system, is used. There are four basic elements involved in the calculation of thermal power based on steam plant calorimetry: The mass flow of the feedwater from the power conversion system, the specific enthalpy of that feedwater, the specific enthalpy of the steam delivered to the power conversion system, and other cycle gains and losses. Of these elements, the accuracy of the feedwater mass flow and the feedwater enthalpy, as determined from its temperature and pressure, are typically the largest contributors to the calorimetric calculation uncertainty. Historically, plants have been required to include a margin of 2% in the calculation of the reactor thermal power for the licensed maximum plant output to account for instrumentation uncertainty. The margin is intended to ensure a cushion between operating power and the power for which safety analyses are performed. Use of approved chordal ultrasonic transit-time technology to make the feedwater flow and temperature measurements (in place of traditional differential-pressure- based instruments and resistance temperature detectors [RTDs]) allows for nuclear plant thermal power calculations accurate to 0.3%-0.4% of plant rated power. This improvement in measurement accuracy has allowed many plant operators in the U.S. and around the world to increase plant power output through Measurement Uncertainty Recapture (MUR) up-rates of up to 1.7% of rated power, while also decreasing the probability of significant over-power events. This paper will examine the basic elements involved in calculation of thermal power using ultrasonic transit-time technology and will discuss the criteria for bounding uncertainties associated with each element in order to achieve reactor thermal power calculations to within 0.3% to 0.4%. (authors)

  15. Self-regulating neutron coincidence counter

    DOE Patents [OSTI]

    Baron, N.

    1980-06-16

    A device for accurately measuring the mass of /sup 240/Pu and /sup 239/Pu in a sample having arbitrary moderation and mixed with various contaminants. The device utilizes a thermal neutron well counter which has two concentric rings of neutron detectors separated by a moderating material surrounding the well. Neutron spectroscopic information derived by the two rings of detectors is used to measure the quantity of /sup 239/Pu and /sup 240/Pu in device which corrects for background radiation, deadtime losses of the detector and electronics and various other constants of the system.

  16. CRC handbook of fast neutron generators

    SciTech Connect (OSTI)

    Csikai, G.J.

    1987-01-01

    This handbook reviews those problems and methods of science and technology where the neutrons produced in the /sup 3/H/d, n//sup 4/He and /sup 2/H/d, N//sup 3/He reactions play the main role. It also discusses possible applications of these small generators as thermal neutron sources, addresses the small accelerators as charged particle and X-ray sources, enables suitable topics to be selected for education and training and provides a wide range of experiments with the detection of neutrons and charged particles, including the study of shielding and the generator technology itself.

  17. 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.

  18. On the Modeling of Local Neutronically-Coupled Flow-Induced Oscillations in Advanced Boiling Water Reactors

    SciTech Connect (OSTI)

    Aniel-Buchheit, Sylvie; Podowski, Michael Z.

    2006-07-01

    The purpose of this paper is to discuss the development in progress of a complete space- and time-dependent model of the coupled neutron kinetic and reactor thermal-hydraulics. The neutron kinetics model is based on two-group diffusion equations with Doppler and void reactivity feedback effects. This model is coupled with the model of two-phase flow and heat transfer in parallel coolant channels. The modeling concepts considered for this purpose include one-dimensional drift flux and two-fluid models, as well a CFD model implemented in the NPHASE advanced computational multiphase fluid dynamics (CMFD) computer code. Two methods of solution for the overall model are proposed. One is based on direct numerical integration of the spatially-discretized governing equations. The other approach is based on a quasi-analytical modal approach to the neutronics model, in which a complete set of eigenvectors is found for step-wise temporal changes of the cross-sections of core materials (fuel and coolant/moderator). The issues investigated in the paper include details of model formulation, as well as the results of calculations for neutronically-coupled density-wave oscillations. (authors)

  19. Neutron Imaging Calibration to Measure Void Fraction

    SciTech Connect (OSTI)

    Geoghegan, Patrick J; Bilheux, Hassina Z; Sharma, Vishaldeep; Fricke, Brian A

    2015-01-01

    Void fraction is an intuitive parameter that describes the fraction of vapor in a two-phase flow. It appears as a key variable in most heat transfer and pressure drop correlations used to design evaporating and condensing heat exchangers, as well as determining charge inventory in refrigeration systems. Void fraction measurement is not straightforward, however, and assumptions on the invasiveness of the measuring technique must be made. Neutron radiography or neutron imaging has the potential to be a truly non-invasive void fraction measuring technique but has until recently only offered qualitative descriptions of two-phase flow, in terms of flow maldistributions, for example. This paper describes the calibration approach necessary to employ neutron imaging to measure steady-state void fraction. Experiments were conducted at the High Flux Isotope Reactor (HFIR) Cold Guide 1D neutron imaging facility at Oak Ridge National Laboratory (ORNL), Oak Ridge, TN, USA.

  20. Enhanced reaction rates in NDP analysis with neutron scattering

    SciTech Connect (OSTI)

    Downing, R. Gregory

    2014-04-15

    Neutron depth profiling (NDP) makes accessible quantitative information on a few isotopic concentration profiles ranging from the surface into the sample a few micrometers. Because the candidate analytes for NDP are few, there is little interference encountered. Furthermore, neutrons have no charge so mixed chemical states in the sample are of no direct concern. There are a few nuclides that exhibit large probabilities for neutron scattering. The effect of neutron scattering on NDP measurements has not previously been evaluated as a basis for either enhancing the reaction rates or as a source of measurement error. Hydrogen is a common element exhibiting large neutron scattering probability found in or around sample volumes being analyzed by NDP. A systematic study was conducted to determine the degree of signal change when neutron scattering occurs during analysis. The relative signal perturbation was evaluated for materials of varied neutron scattering probability, concentration, total mass, and geometry. Signal enhancements up to 50% are observed when the hydrogen density is high and in close proximity to the region of analysis with neutron beams of sub thermal energies. Greater signal enhancements for the same neutron number density are reported for thermal neutron beams. Even adhesive tape used to position the sample produces a measureable signal enhancement. Because of the shallow volume, negligible distortion of the NDP measured profile shape is encountered from neutron scattering.

  1. Real-Time Active Cosmic Neutron Background Reduction Methods

    SciTech Connect (OSTI)

    Mukhopadhyay, Sanjoy; Maurer, Richard; Wolff, Ronald; Mitchell, Stephen; Guss, Paul

    2013-09-01

    Neutron counting using large arrays of pressurized 3He proportional counters from an aerial system or in a maritime environment suffers from the background counts from the primary cosmic neutrons and secondary neutrons caused by cosmic ray?induced mechanisms like spallation and charge-exchange reaction. This paper reports the work performed at the Remote Sensing LaboratoryAndrews (RSL-A) and results obtained when using two different methods to reduce the cosmic neutron background in real time. Both methods used shielding materials with a high concentration (up to 30% by weight) of neutron-absorbing materials, such as natural boron, to remove the low-energy neutron flux from the cosmic background as the first step of the background reduction process. Our first method was to design, prototype, and test an up-looking plastic scintillator (BC-400, manufactured by Saint Gobain Corporation) to tag the cosmic neutrons and then create a logic pulse of a fixed time duration (~120 ?s) to block the data taken by the neutron counter (pressurized 3He tubes running in a proportional counter mode). The second method examined the time correlation between the arrival of two successive neutron signals to the counting array and calculated the excess of variance (Feynman variance Y2F)1 in the neutron count distribution from Poisson distribution. The dilution of this variance from cosmic background values ideally would signal the presence of man-made neutrons.2 The first method has been technically successful in tagging the neutrons in the cosmic-ray flux and preventing them from being counted in the 3He tube array by electronic vetofield measurement work shows the efficiency of the electronic veto counter to be about 87%. The second method has successfully derived an empirical relationship between the percentile non-cosmic component in a neutron flux and the Y2F of the measured neutron count distribution. By using shielding materials alone, approximately 55% of the neutron flux from man-made sources like 252Cf or Am-Be was removed.

  2. 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.

  3. NEUTRON SOURCE

    DOE Patents [OSTI]

    Reardon, W.A.; Lennox, D.H.; Nobles, R.G.

    1959-01-13

    A neutron source of the antimony--beryllium type is presented. The source is comprised of a solid mass of beryllium having a cylindrical recess extending therein and a cylinder containing antimony-124 slidably disposed within the cylindrical recess. The antimony cylinder is encased in aluminum. A berylliunn plug is removably inserted in the open end of the cylindrical recess to completely enclose the antimony cylinder in bsryllium. The plug and antimony cylinder are each provided with a stud on their upper ends to facilitate handling remotely.

  4. NEUTRON COUNTER

    DOE Patents [OSTI]

    Curtis, C.D.; Carlson, R.L.; Tubinis, M.P.

    1958-07-29

    An ionization chamber instrument is described for cylindrical electrodes with an ionizing gag filling the channber. The inner electrode is held in place by a hermetic insulating seal at one end of the outer electrode, the other end of the outer electrode being closed by a gas filling tube. The outer surface of the inner electrode is coated with an active material which is responsive to neutron bombardment, such as uranium235 or boron-10, to produce ionizing radiations in the gas. The transverse cross sectional area of the inner electrode is small in relation to that of the channber whereby substantially all of the radiations are directed toward the outer electrode.

  5. The synchronous active neutron detection assay system

    SciTech Connect (OSTI)

    Pickrell, M.M.; Kendall, P.K.

    1994-08-01

    We have begun to develop a novel technique for active neutron assay of fissile material in spent nuclear fuel. This approach will exploit a 14-MeV neutron generator developed by Schlumberger. The technique, termed synchronous active neutron detection (SAND), follows a method used routinely in other branches of physics to detect very small signals in presence of large backgrounds. Synchronous detection instruments are widely available commercially and are termed ``lock-in`` amplifiers. We have implemented a digital lock-in amplifier in conjunction with the Schlumberger neutron generator to explore the possibility of synchronous detection with active neutrons. The Schlumberger system can operate at up to a 50% duty factor, in effect, a square wave of neutron yield. Results are preliminary but promising. The system is capable of resolving the fissile material contained in a small fraction of the fuel rods in a cold fuel assembly; it also appears resilient to background neutron interference. The interrogating neutrons appear to be non-thermal and penetrating. Work remains to fully explore relevant physics and optimize instrument design.

  6. 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...

  7. Advancing Materials Science using Neutrons at Oak Ridge National Laboratory

    ScienceCinema (OSTI)

    Carpenter, John

    2014-06-03

    Jack Carpenter, pioneer of accelerator-based pulsed spallation neutron sources, talks about neutron science at Oak Ridge National Laboratory (ORNL) and a need for a second target station at the Spallation Neutron Source (SNS). ORNL is the Department of Energy's largest multiprogram science and energy laboratory, and is home to two scientific user facilities serving the neutron science research community: the High Flux Isotope Reactor (HFIR) and SNS. HFIR and SNS provide researchers with unmatched capabilities for understanding the structure and properties of materials, macromolecular and biological systems, and the fundamental physics of the neutron. Neutrons provide a window through which to view materials at a microscopic level that allow researchers to develop better materials and better products. Neutrons enable us to understand materials we use in everyday life. Carpenter explains the need for another station to produce long wavelength neutrons, or cold neutrons, to answer questions that are addressed only with cold neutrons. The second target station is optimized for that purpose. Modern technology depends more and more upon intimate atomic knowledge of materials, and neutrons are an ideal probe.

  8. Irradiation effect on deuterium behaviour in low-dose HFIR neutron-irradiated tungsten

    SciTech Connect (OSTI)

    Shimada, Masashi; Cao, G.; Otsuka, T.; Hara, M.; Kobayashi, M.; Oya, Y.; Hatano, Y.

    2014-12-01

    Tungsten samples were irradiated by neutrons in the High Flux Isotope Reactor, Oak Ridge National Laboratory at reactor coolant temperatures of 50-70C to low displacement damage of 0.025 and 0.3 dpa under the framework of the US-Japan TITAN program (2007-2013). After cooling down, the HFIR neutron-irradiated tungsten samples were exposed to deuterium plasmas in the Tritium Plasma Experiment, Idaho National Laboratory at 100, 200 and 500 C twice at the ion fluence of 510? m? to reach a total ion fluence of 110? m? in order to investigate the near surface deuterium retention and saturation via nuclear reaction analysis. Final thermal desorption spectroscopy was performed to elucidate irradiation effect on total deuterium retention. Nuclear reaction analysis results showed that the maximum near surface (<5 m depth) deuterium concentration increased from 0.5 at % D/W in 0.025 dpa samples to 0.8 at. % D/W in 0.3 dpa samples. The large discrepancy between the total retention via thermal desorption spectroscopy and the near surface retention via nuclear reaction analysis indicated the deuterium was migrated and trapped in bulk (at least 50 m depth for 0.025 dpa and 35 m depth for 0.025 dpa) at 500 C case even in the relatively low ion fluence of 10? m?.

  9. 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.

  10. Ground water and snow sensor based on directional detection of cosmogenic neutrons.

    SciTech Connect (OSTI)

    Cooper, Robert Lee; Marleau, Peter; Griffin, Patrick J.

    2011-06-01

    A fast neutron detector is being developed to measure the cosmic ray neutron flux in order to measure soil moisture. Soil that is saturated with water has an enhanced ability to moderate fast neutrons, removing them from the backscatter spectrum. The detector is a two-element, liquid scintillator detector. The choice of liquid scintillator allows rejection of gamma background contamination from the desired neutron signal. This enhances the ability to reconstruct the energy and direction of a coincident neutron event. The ability to image on an event-by-event basis allows the detector to selectively scan the neutron flux as a function of distance from the detector. Calibrations, simulations, and optimization have been completed to understand the detector response to neutron sources at variable distances and directions. This has been applied to laboratory background measurements in preparation for outdoor field tests.

  11. High Brightness Neutron Source for Radiography

    SciTech Connect (OSTI)

    Cremer, J. T.; Piestrup, Melvin, A.; Gary, Charles, K.; Harris, Jack, L. Williams, David, J.; Jones, Glenn, E.; Vainionpaa, J. , H.; Fuller, Michael, J.; Rothbart, George, H.; Kwan, J., W.; Ludewigt, B., A.; Gough, R.., A..; Reijonen, Jani; Leung, Ka-Ngo

    2008-12-08

    This research and development program was designed to improve nondestructive evaluation of large mechanical objects by providing both fast and thermal neutron sources for radiography. Neutron radiography permits inspection inside objects that x-rays cannot penetrate and permits imaging of corrosion and cracks in low-density materials. Discovering of fatigue cracks and corrosion in piping without the necessity of insulation removal is possible. Neutron radiography sources can provide for the nondestructive testing interests of commercial and military aircraft, public utilities and petrochemical organizations. Three neutron prototype neutron generators were designed and fabricated based on original research done at the Lawrence Berkeley National Laboratory (LBNL). The research and development of these generators was successfully continued by LBNL and Adelphi Technology Inc. under this STTR. The original design goals of high neutron yield and generator robustness have been achieved, using new technology developed under this grant. In one prototype generator, the fast neutron yield and brightness was roughly 10 times larger than previously marketed neutron generators using the same deuterium-deuterium reaction. In another generator, we integrate a moderator with a fast neutron source, resulting in a high brightness thermal neutron generator. The moderator acts as both conventional moderator and mechanical and electrical support structure for the generator and effectively mimics a nuclear reactor. In addition to the new prototype generators, an entirely new plasma ion source for neutron production was developed. First developed by LBNL, this source uses a spiral antenna to more efficiently couple the RF radiation into the plasma, reducing the required gas pressure so that the generator head can be completely sealed, permitting the possible use of tritium gas. This also permits the generator to use the deuterium-tritium reaction to produce 14-MeV neutrons with increases of yield of two orders of magnitude. The first fast neutron radiographic images were obtained using neutron cameras and a new fast neutron generator. These early images demonstrated the feasibility of using fast neutrons for imaging and penetrating thick objects of high density and imaging. Fast neutrons can be used to image low atomic number materials (e.g. plastics, explosives, lubricants and ceramics) that are shielded by high density materials (e.g. lead, tungsten and uranium). Fast neutron radiography could be used as a means to screen weapons for flaws and chemical stability. X-ray radiography can not easily do this. Fast neutron imaging is technically difficult and, consequently, a completely undeveloped market. Two of the generators were designed to have small source size and high brightness, ideal for fast-neutron imaging. With these generators we successfully used two fast neutron cameras: one developed by us, and another developed by a collaborator, Commonwealth Scientific and Industrial Research Organization, CSIRO. We have successfully used these cameras to obtain low resolution images of various objects such as pipe fittings filled with water and other mechanical objects. Higher resolution and contrast images are expected by decreasing the source size and increasing generator yield.

  12. Composite neutron absorbing coatings for nuclear criticality control

    DOE Patents [OSTI]

    Wright, Richard N.; Swank, W. David; Mizia, Ronald E.

    2005-07-19

    Thermal neutron absorbing composite coating materials and methods of applying such coating materials to spent nuclear fuel storage systems are provided. A composite neutron absorbing coating applied to a substrate surface includes a neutron absorbing layer overlying at least a portion of the substrate surface, and a corrosion resistant top coat layer overlying at least a portion of the neutron absorbing layer. An optional bond coat layer can be formed on the substrate surface prior to forming the neutron absorbing layer. The neutron absorbing layer can include a neutron absorbing material, such as gadolinium oxide or gadolinium phosphate, dispersed in a metal alloy matrix. The coating layers may be formed by a plasma spray process or a high velocity oxygen fuel process.

  13. Accelerator-based neutron source for boron neutron capture therapy (BNCT) and method

    DOE Patents [OSTI]

    Yoon, Woo Y. (Idaho Falls, ID); Jones, James L. (Idaho Falls, ID); Nigg, David W. (Idaho Falls, ID); Harker, Yale D. (Idaho Falls, ID)

    1999-01-01

    A source for boron neutron capture therapy (BNCT) comprises a body of photoneutron emitter that includes heavy water and is closely surrounded in heat-imparting relationship by target material; one or more electron linear accelerators for supplying electron radiation having energy of substantially 2 to 10 MeV and for impinging such radiation on the target material, whereby photoneutrons are produced and heat is absorbed from the target material by the body of photoneutron emitter. The heavy water is circulated through a cooling arrangement to remove heat. A tank, desirably cylindrical or spherical, contains the heavy water, and a desired number of the electron accelerators circumferentially surround the tank and the target material as preferably made up of thin plates of metallic tungsten. Neutrons generated within the tank are passed through a surrounding region containing neutron filtering and moderating materials and through neutron delimiting structure to produce a beam or beams of epithermal neutrons normally having a minimum flux intensity level of 1.0.times.10.sup.9 neutrons per square centimeter per second. Such beam or beams of epithermal neutrons are passed through gamma ray attenuating material to provide the required epithermal neutrons for BNCT use.

  14. Accelerator-based neutron source for boron neutron capture therapy (BNCT) and method

    DOE Patents [OSTI]

    Yoon, W.Y.; Jones, J.L.; Nigg, D.W.; Harker, Y.D.

    1999-05-11

    A source for boron neutron capture therapy (BNCT) comprises a body of photoneutron emitter that includes heavy water and is closely surrounded in heat-imparting relationship by target material; one or more electron linear accelerators for supplying electron radiation having energy of substantially 2 to 10 MeV and for impinging such radiation on the target material, whereby photoneutrons are produced and heat is absorbed from the target material by the body of photoneutron emitter. The heavy water is circulated through a cooling arrangement to remove heat. A tank, desirably cylindrical or spherical, contains the heavy water, and a desired number of the electron accelerators circumferentially surround the tank and the target material as preferably made up of thin plates of metallic tungsten. Neutrons generated within the tank are passed through a surrounding region containing neutron filtering and moderating materials and through neutron delimiting structure to produce a beam or beams of epithermal neutrons normally having a minimum flux intensity level of 1.0{times}10{sup 9} neutrons per square centimeter per second. Such beam or beams of epithermal neutrons are passed through gamma ray attenuating material to provide the required epithermal neutrons for BNCT use. 3 figs.

  15. Sensitivity of the interpretation of the experimental ion thermal

    Office of Scientific and Technical Information (OSTI)

    diffusivity to the determination of the ion conductive heat flux (Journal Article) | SciTech Connect Sensitivity of the interpretation of the experimental ion thermal diffusivity to the determination of the ion conductive heat flux Citation Details In-Document Search Title: Sensitivity of the interpretation of the experimental ion thermal diffusivity to the determination of the ion conductive heat flux A moments equation formalism for the interpretation of the experimental ion thermal

  16. Thermal Hydraulic Characteristics of Fuel Defects in Plate Type...

    Office of Scientific and Technical Information (OSTI)

    Title: Thermal Hydraulic Characteristics of Fuel Defects in Plate Type Nuclear Research Reactors Turbulent flow coupled with heat transfer is investigated for a High Flux Isotope ...

  17. High Flux Isotope Reactor | Neutron Science at ORNL

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

    ORNL achieves milestone with plutonium-238 sample December 22, 2015 Jaime Fernandez-Baca and Mark Lumsden have been elected fellows of the American Physical Society. Two ORNL...

  18. First Evaluation of the Biologic Effectiveness Factors of Boron Neutron Capture Therapy (BNCT) in a Human Colon Carcinoma Cell Line

    SciTech Connect (OSTI)

    Dagrosa, Maria Alejandra; Crivello, Martin; Perona, Marina; Thorp, Silvia; Santa Cruz, Gustavo Alberto; Pozzi, Emiliano; Casal, Mariana; Thomasz, Lisa; Cabrini, Romulo; Kahl, Steven; Juvenal, Guillermo Juan; Pisarev, Mario Alberto

    2011-01-01

    Purpose: DNA lesions produced by boron neutron capture therapy (BNCT) and those produced by gamma radiation in a colon carcinoma cell line were analyzed. We have also derived the relative biologic effectiveness factor (RBE) of the neutron beam of the RA-3- Argentine nuclear reactor, and the compound biologic effectiveness (CBE) values for p-boronophenylalanine ({sup 10}BPA) and for 2,4-bis ({alpha},{beta}-dihydroxyethyl)-deutero-porphyrin IX ({sup 10}BOPP). Methods and Materials: Exponentially growing human colon carcinoma cells (ARO81-1) were distributed into the following groups: (1) BPA (10 ppm {sup 10}B) + neutrons, (2) BOPP (10 ppm {sup 10}B) + neutrons, (3) neutrons alone, and (4) gamma rays ({sup 60}Co source at 1 Gy/min dose-rate). Different irradiation times were used to obtain total absorbed doses between 0.3 and 5 Gy ({+-}10%) (thermal neutrons flux = 7.5 10{sup 9} n/cm{sup 2} sec). Results: The frequency of micronucleated binucleated cells and the number of micronuclei per micronucleated binucleated cells showed a dose-dependent increase until approximately 2 Gy. The response to gamma rays was significantly lower than the response to the other treatments (p < 0.05). The irradiations with neutrons alone and neutrons + BOPP showed curves that did not differ significantly from, and showed less DNA damage than, irradiation with neutrons + BPA. A decrease in the surviving fraction measured by 3-(4,5-dimetiltiazol-2-il)-2,5-difeniltetrazolium bromide (MTT) assay as a function of the absorbed dose was observed for all the treatments. The RBE and CBE factors calculated from cytokinesis block micronucleus (CBMN) and MTT assays were, respectively, the following: beam RBE: 4.4 {+-} 1.1 and 2.4 {+-} 0.6; CBE for BOPP: 8.0 {+-} 2.2 and 2.0 {+-} 1; CBE for BPA: 19.6 {+-} 3.7 and 3.5 {+-} 1.3. Conclusions: BNCT and gamma irradiations showed different genotoxic patterns. To our knowledge, these values represent the first experimental ones obtained for the RA-3 in a biologic model and could be useful for future experimental studies for the application of BNCT to colon carcinoma.

  19. 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.

  20. Neutron reflecting supermirror structure

    DOE Patents [OSTI]

    Wood, James L. (Drayton Plains, MI)

    1992-01-01

    An improved neutron reflecting supermirror structure comprising a plurality of stacked sets of bilayers of neutron reflecting materials. The improved neutron reflecting supermirror structure is adapted to provide extremely good performance at high incidence angles, i.e. up to four time the critical angle of standard neutron mirror structures. The reflection of neutrons striking the supermirror structure at a high critical angle provides enhanced neutron throughput, and hence more efficient and economical use of neutron sources. One layer of each set of bilayers consist of titanium, and the second layer of each set of bilayers consist of an alloy of nickel with carbon interstitially present in the nickel alloy.

  1. DIFFERENTIAL SPECTRUM OF NEUTRONS AT CHACALTAYA-BOLIVIA

    SciTech Connect (OSTI)

    Mayta, R.; Zanini, A.; Ticona, R.; Velarde, A.

    2009-04-30

    We describe the Neutron Spectrometer Experiment installed at Chacaltaya Cosmic Rays Observatory (68 deg. O, 16.2 deg. S), located in Bolivia, at 5230 m.a.s.l. This experimental system is constituted by passive detectors which register the flux of neutrons, in an energy range of 10 KeV-20 MeV. Using the unfolding code BUNTO a peak around 1 MeV of the characteristic spectrum of neutrons was obtained. Experimental values, observed during April of 2008, are compared with similar ones carried out in 1997 at the same place, in order to look for eventual changes due to local atmosphere. A similar experiment was also carried up at the Laboratory of Testa Grigia-Italy (45.56 deg. N, 7.42 deg. E,. 3480 m.a.l.s). Data of both stations allow us to compare the spectra in order to explain the difference of neutron flux of these two stations.

  2. Clifford G. Shull, Neutron Diffraction, Hydrogen Atoms, and Neutron...

    Office of Scientific and Technical Information (OSTI)

    ... The Diffraction of Neutrons by Crystalline Powders; DOE Technical Report; 1948 Neutron Diffraction Studies, DOE Technical Report, 1948 Laue Photography of Neutron Diffraction; ...

  3. Neutron capture cross section standards for BNL 325, Fourth Edition

    SciTech Connect (OSTI)

    Holden, N.E.

    1981-01-01

    This report evaluates the experimental data and recommends values for the thermal neutron cross sections and resonance integrals for the neutron capture reactions: /sup 55/Mn(n,..gamma..), /sup 59/Co(n,..gamma..) and /sup 197/Au(n,..gamma..). The failure of lithium and boron as standards due to the natural variation of the absorption cross sections of these elements is discussed. The Westcott convention, which describes the neutron spectrum as a thermal Maxwellian distribution with an epithermal component, is also discussed.

  4. Development of A Self Biased High Efficiency Solid-State Neutron Detector

    Office of Scientific and Technical Information (OSTI)

    for MPACT Applications (Technical Report) | SciTech Connect Development of A Self Biased High Efficiency Solid-State Neutron Detector for MPACT Applications Citation Details In-Document Search Title: Development of A Self Biased High Efficiency Solid-State Neutron Detector for MPACT Applications Neutron detection is an important aspect of materials protection, accounting, and control for transmutation (MPACT). Currently He-3 filled thermal neutron detectors are utilized in many applications;

  5. 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.

  6. Neutron range spectrometer

    DOE Patents [OSTI]

    Manglos, Stephen H.

    1989-06-06

    A neutron range spectrometer and method for determining the neutron energy spectrum of a neutron emitting source are disclosed. Neutrons from the source are collimnated along a collimation axis and a position sensitive neutron counter is disposed in the path of the collimated neutron beam. The counter determines positions along the collimation axis of interactions between the neutrons in the neutron beam and a neutron-absorbing material in the counter. From the interaction positions, a computer analyzes the data and determines the neutron energy spectrum of the neutron beam. The counter is preferably shielded and a suitable neutron-absorbing material is He-3. The computer solves the following equation in the analysis: ##EQU1## where: N(x).DELTA.x=the number of neutron interactions measured between a position x and x+.DELTA.x, A.sub.i (E.sub.i).DELTA.E.sub.i =the number of incident neutrons with energy between E.sub.i and E.sub.i +.DELTA.E.sub.i, and C=C(E.sub.i)=N .sigma.(E.sub.i) where N=the number density of absorbing atoms in the position sensitive counter means and .sigma. (E.sub.i)=the average cross section of the absorbing interaction between E.sub.i and E.sub.i +.DELTA.E.sub.i.

  7. Cumulative fission yields of short-lived isotopes under natural-abundance-boron-carbide-moderated neutron spectrum

    SciTech Connect (OSTI)

    Finn, Erin C.; Metz, Lori A.; Greenwood, Lawrence R.; Pierson, Bruce; Wittman, Richard S.; Friese, Judah I.; Kephart, Rosara F.

    2015-04-09

    The availability of gamma spectroscopy data on samples containing mixed fission products at short times after irradiation is limited. Due to this limitation, data interpretation methods for gamma spectra of mixed fission product samples, where the individual fission products have not been chemically isolated from interferences, are not well-developed. The limitation is particularly pronounced for fast pooled neutron spectra because of the lack of available fast reactors in the United States. Samples containing the actinide isotopes 233, 235, 238U, 237Np, and 239Pu individually were subjected to a 2$ pulse in the Washington State University 1 MW TRIGA reactor. To achieve a fission-energy neutron spectrum, the spectrum was tailored using a natural abundance boron carbide capsule to absorb neutrons in the thermal and epithermal region of the spectrum. Our tailored neutron spectrum is unique to the WSU reactor facility, consisting of a soft fission spectrum that contains some measurable flux in the resonance region. This results in a neutron spectrum at greater than 0.1 keV with an average energy of 70 keV, similar to fast reactor spectra and approaching that of 235U fission. Unique fission product gamma spectra were collected from 4 minutes to 1 week after fission using single-crystal high purity germanium detectors. Cumulative fission product yields measured in the current work generally agree with published fast pooled fission product yield values from ENDF/B-VII, though a bias was noted for 239Pu. The present work contributes to the compilation of energy-resolved fission product yield nuclear data for nuclear forensic purposes.

  8. Neutron streak camera

    DOE Patents [OSTI]

    Wang, C.L.

    1981-05-14

    Apparatus for improved sensitivity and time resolution of a neutron measurement. The detector is provided with an electrode assembly having a neutron sensitive cathode which emits relatively low energy secondary electrons. The neutron sensitive cathode has a large surface area which provides increased sensitivity by intercepting a greater number of neutrons. The cathode is also curved to compensate for differences in transit time of the neutrons emanating from the point source. The slower speeds of the secondary electrons emitted from a certain portion of the cathode are matched to the transit times of the neutrons impinging thereupon.

  9. Neutron streak camera

    DOE Patents [OSTI]

    Wang, C.L.

    1983-09-13

    Disclosed is an apparatus for improved sensitivity and time resolution of a neutron measurement. The detector is provided with an electrode assembly having a neutron sensitive cathode which emits relatively low energy secondary electrons. The neutron sensitive cathode has a large surface area which provides increased sensitivity by intercepting a greater number of neutrons. The cathode is also curved to compensate for differences in transit time of the neutrons emanating from the point source. The slower speeds of the secondary electrons emitted from a certain portion of the cathode are matched to the transit times of the neutrons impinging thereupon. 4 figs.

  10. Neutron streak camera

    DOE Patents [OSTI]

    Wang, Ching L.

    1983-09-13

    Apparatus for improved sensitivity and time resolution of a neutron measurement. The detector is provided with an electrode assembly having a neutron sensitive cathode which emits relatively low energy secondary electrons. The neutron sensitive cathode has a large surface area which provides increased sensitivity by intercepting a greater number of neutrons. The cathode is also curved to compensate for differences in transit time of the neutrons emanating from the point source. The slower speeds of the secondary electrons emitted from a certain portion of the cathode are matched to the transit times of the neutrons impinging thereupon.

  11. Layered semiconductor neutron detectors

    DOE Patents [OSTI]

    Mao, Samuel S; Perry, Dale L

    2013-12-10

    Room temperature operating solid state hand held neutron detectors integrate one or more relatively thin layers of a high neutron interaction cross-section element or materials with semiconductor detectors. The high neutron interaction cross-section element (e.g., Gd, B or Li) or materials comprising at least one high neutron interaction cross-section element can be in the form of unstructured layers or micro- or nano-structured arrays. Such architecture provides high efficiency neutron detector devices by capturing substantially more carriers produced from high energy .alpha.-particles or .gamma.-photons generated by neutron interaction.

  12. Organic metal neutron detector

    DOE Patents [OSTI]

    Butler, M.A.; Ginley, D.S.

    1984-11-21

    A device for detection of neutrons comprises: as an active neutron sensing element, a conductive organic polymer having an electrical conductivity and a cross-section for said neutrons whereby a detectable change in said conductivity is caused by impingement of said neutrons on the conductive organic polymer which is responsive to a property of said polymer which is altered by impingement of said neutrons on the polymer; and means for associating a change in said alterable property with the presence of neutrons at the location of said device.

  13. Porcelain enamel neutron absorbing material

    DOE Patents [OSTI]

    Iverson, D.C.

    1987-11-20

    A porcelain enamel composition as a neutron absorbing material can be prepared of a major proportion by weight of a cadmium compound and a minor proportion of compound of boron, lithium and silicon. These compounds in the form of a porcelain enamel coating or layer on several alloys has been found to be particularly effective in enhancing the nuclear safety of equipment for use in the processing and storage of fissile material. The composition of the porcelain enamel coating can be tailored to match the coefficient of thermal expansion of the equipment to be coated and excellent coating adhesion can be achieved. 2 figs.

  14. Porcelain enamel neutron absorbing material

    DOE Patents [OSTI]

    Iverson, Daniel C.

    1990-01-01

    A porcelain enamel composition as a neutron absorbing material can be prepared of a major proportion by weight of a cadmium compound and a minor proportion of compounds of boron, lithium and silicon. These compounds in the form of a porcelain enamel coating or layer on several alloys has been found to be particularly effective in enhancing the nuclear safety of equipment for use in the processing and storage of fissile material. The composition of the porcelain enamel coating can be tailored to match the coefficient of thermal expansion of the equipment to be coated and excellent coating adhesion can be achieved.

  15. 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.

  16. Measuring and monitoring KIPT Neutron Source Facility Reactivity

    SciTech Connect (OSTI)

    Cao, Yan; Gohar, Yousry; Zhong, Zhaopeng

    2015-08-01

    Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on developing and constructing a neutron source facility at Kharkov, Ukraine. The facility consists of an accelerator-driven subcritical system. The accelerator has a 100 kW electron beam using 100 MeV electrons. The subcritical assembly has keff less than 0.98. To ensure the safe operation of this neutron source facility, the reactivity of the subcritical core has to be accurately determined and continuously monitored. A technique which combines the area-ratio method and the flux-to-current ratio method is purposed to determine the reactivity of the KIPT subcritical assembly at various conditions. In particular, the area-ratio method can determine the absolute reactivity of the subcritical assembly in units of dollars by performing pulsed-neutron experiments. It provides reference reactivities for the flux-to-current ratio method to track and monitor the reactivity deviations from the reference state while the facility is at other operation modes. Monte Carlo simulations are performed to simulate both methods using the numerical model of the KIPT subcritical assembly. It is found that the reactivities obtained from both the area-ratio method and the flux-to-current ratio method are spatially dependent on the neutron detector locations and types. Numerical simulations also suggest optimal neutron detector locations to minimize the spatial effects in the flux-to-current ratio method. The spatial correction factors are calculated using Monte Carlo methods for both measuring methods at the selected neutron detector locations. Monte Carlo simulations are also performed to verify the accuracy of the flux-to-current ratio method in monitoring the reactivity swing during a fuel burnup cycle.

  17. A TARGETED SEARCH FOR POINT SOURCES OF EeV NEUTRONS

    SciTech Connect (OSTI)

    Aab, A.; Abreu, P.; Andringa, S.; Aglietta, M.; Ahlers, M.; Ahn, E. J.; Al Samarai, I.; Albuquerque, I. F. M.; Allekotte, I.; Allen, J.; Allison, P.; Almela, A.; Castillo, J. Alvarez; Alvarez-Muñiz, J.; Batista, R. Alves; Ambrosio, M.; Aramo, C.; Aminaei, A.; Anchordoqui, L.; Arqueros, F.; Collaboration: Pierre Auger Collaboration101; and others

    2014-07-10

    A flux of neutrons from an astrophysical source in the Galaxy can be detected in the Pierre Auger Observatory as an excess of cosmic-ray air showers arriving from the direction of the source. To avoid the statistical penalty for making many trials, classes of objects are tested in combinations as nine ''target sets'', in addition to the search for a neutron flux from the Galactic center or from the Galactic plane. Within a target set, each candidate source is weighted in proportion to its electromagnetic flux, its exposure to the Auger Observatory, and its flux attenuation factor due to neutron decay. These searches do not find evidence for a neutron flux from any class of candidate sources. Tabulated results give the combined p-value for each class, with and without the weights, and also the flux upper limit for the most significant candidate source within each class. These limits on fluxes of neutrons significantly constrain models of EeV proton emission from non-transient discrete sources in the Galaxy.

  18. Neutrons for Catalysis: A Workshop on Neutron Scattering Techniques for Studies in Catalysis

    SciTech Connect (OSTI)

    Overbury, Steven {Steve} H; Coates, Leighton; Herwig, Kenneth W; Kidder, Michelle

    2011-10-01

    This report summarizes the Workshop on Neutron Scattering Techniques for Studies in Catalysis, held at the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory (ORNL) on September 16 and 17, 2010. The goal of the Workshop was to bring experts in heterogeneous catalysis and biocatalysis together with neutron scattering experimenters to identify ways to attack new problems, especially Grand Challenge problems in catalysis, using neutron scattering. The Workshop locale was motivated by the neutron capabilities at ORNL, including the High Flux Isotope Reactor (HFIR) and the new and developing instrumentation at the SNS. Approximately 90 researchers met for 1 1/2 days with oral presentations and breakout sessions. Oral presentations were divided into five topical sessions aimed at a discussion of Grand Challenge problems in catalysis, dynamics studies, structure characterization, biocatalysis, and computational methods. Eleven internationally known invited experts spoke in these sessions. The Workshop was intended both to educate catalyst experts about the methods and possibilities of neutron methods and to educate the neutron community about the methods and scientific challenges in catalysis. Above all, it was intended to inspire new research ideas among the attendees. All attendees were asked to participate in one or more of three breakout sessions to share ideas and propose new experiments that could be performed using the ORNL neutron facilities. The Workshop was expected to lead to proposals for beam time at either the HFIR or the SNS; therefore, it was expected that each breakout session would identify a few experiments or proof-of-principle experiments and a leader who would pursue a proposal after the Workshop. Also, a refereed review article will be submitted to a prominent journal to present research and ideas illustrating the benefits and possibilities of neutron methods for catalysis research.

  19. Dose equivalent neutron dosimeter

    DOE Patents [OSTI]

    Griffith, Richard V.; Hankins, Dale E.; Tomasino, Luigi; Gomaa, Mohamed A. M.

    1983-01-01

    A neutron dosimeter is disclosed which provides a single measurements indicating the amount of potential biological damage resulting from the neutron exposure of the wearer, for a wide range of neutron energies. The dosimeter includes a detecting sheet of track etch detecting material such as a carbonate plastic, for detecting higher energy neutrons, and a radiator layer containing conversion material such as .sup.6 Li and .sup.10 B lying adjacent to the detecting sheet for converting moderate energy neutrons to alpha particles that produce tracks in the adjacent detecting sheet. The density of conversion material in the radiator layer is of an amount which is chosen so that the density of tracks produced in the detecting sheet is proportional to the biological damage done by neutrons, regardless of whether the tracks are produced as the result of moderate energy neutrons striking the radiator layer or as the result of higher energy neutrons striking the sheet of track etch material.

  20. Pulsed-neutron monochromator

    DOE Patents [OSTI]

    Mook, H.A. Jr.

    1984-01-01

    In one aspect, the invention is an improved pulsed-neutron monochromator of the vibrated-crystal type. The monochromator is designed to provide neutron pulses which are characterized both by short duration and high density. A row of neutron-reflecting crystals is disposed in a neutron beam to reflect neutrons onto a common target. The crystals in the row define progressively larger neutron-scattering angles and are vibrated sequentially in descending order with respect to the size of their scattering angles, thus generating neutron pulses which arrive simultaneously at the target. Transducers are coupled to one end of the crystals to vibrate them in an essentially non-resonant mode. The transducers propagate transverse waves in the crystal which progress longitudinally therein. The waves are absorbed at the undriven ends of the crystals by damping material mounted thereon. In another aspect, the invention is a method for generating neutron pulses characterized by high intensity and short duration.

  1. ULTRASONIC NEUTRON DOSIMETER

    DOE Patents [OSTI]

    Truell, R.; de Klerk, J.; Levy, P.W.

    1960-02-23

    A neutron dosimeter is described which utilizes ultrasonic waves in the megacycle region for determination of the extent of neutron damage in a borosilicate glass through ultrasonic wave velocity and attenuation measurements before and after damage.

  2. Ultrafast neutron detector

    DOE Patents [OSTI]

    Wang, C.L.

    1985-06-19

    A neutron detector of very high temporal resolution is described. It may be used to measure distributions of neutrons produced by fusion reactions that persist for times as short as about 50 picoseconds.

  3. Pulsed-neutron monochromator

    DOE Patents [OSTI]

    Mook, Jr., Herbert A. (Oak Ridge, TN)

    1985-01-01

    In one aspect, the invention is an improved pulsed-neutron monochromator of the vibrated-crystal type. The monochromator is designed to provide neutron pulses which are characterized both by short duration and high density. A row of neutron-reflecting crystals is disposed in a neutron beam to reflect neutrons onto a common target. The crystals in the row define progressively larger neutron-scattering angles and are vibrated sequentially in descending order with respect to the size of their scattering angles, thus generating neutron pulses which arrive simultaneously at the target. Transducers are coupled to one end of the crystals to vibrate them in an essentially non-resonant mode. The transducers propagate transverse waves in the crystal which progress longitudinally therein. The wave are absorbed at the undriven ends of the crystals by damping material mounted thereon. In another aspect, the invention is a method for generating neutron pulses characterized by high intensity and short duration.

  4. Neutron dose equivalent meter

    DOE Patents [OSTI]

    Olsher, Richard H. (Los Alamos, NM); Hsu, Hsiao-Hua (Los Alamos, NM); Casson, William H. (Los Alamos, NM); Vasilik, Dennis G. (Los Alamos, NM); Kleck, Jeffrey H. (Menlo Park, CA); Beverding, Anthony (Foster City, CA)

    1996-01-01

    A neutron dose equivalent detector for measuring neutron dose capable of accurately responding to neutron energies according to published fluence to dose curves. The neutron dose equivalent meter has an inner sphere of polyethylene, with a middle shell overlying the inner sphere, the middle shell comprising RTV.RTM. silicone (organosiloxane) loaded with boron. An outer shell overlies the middle shell and comprises polyethylene loaded with tungsten. The neutron dose equivalent meter defines a channel through the outer shell, the middle shell, and the inner sphere for accepting a neutron counter tube. The outer shell is loaded with tungsten to provide neutron generation, increasing the neutron dose equivalent meter's response sensitivity above 8 MeV.

  5. 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 ...

  6. 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 ...

  7. 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 ...

  8. Measurement of uranium and plutonium in solid waste by passive photon or neutron counting and isotopic neutron source interrogation

    SciTech Connect (OSTI)

    Crane, T.W.

    1980-03-01

    A summary of the status and applicability of nondestructive assay (NDA) techniques for the measurement of uranium and plutonium in 55-gal barrels of solid waste is reported. The NDA techniques reviewed include passive gamma-ray and x-ray counting with scintillator, solid state, and proportional gas photon detectors, passive neutron counting, and active neutron interrogation with neutron and gamma-ray counting. The active neutron interrogation methods are limited to those employing isotopic neutron sources. Three generic neutron sources (alpha-n, photoneutron, and /sup 252/Cf) are considered. The neutron detectors reviewed for both prompt and delayed fission neutron detection with the above sources include thermal (/sup 3/He, /sup 10/BF/sub 3/) and recoil (/sup 4/He, CH/sub 4/) proportional gas detectors and liquid and plastic scintillator detectors. The instrument found to be best suited for low-level measurements (< 10 nCi/g) is the /sup 252/Cf Shuffler. The measurement technique consists of passive neutron counting followed by cyclic activation using a /sup 252/Cf source and delayed neutron counting with the source withdrawn. It is recommended that a waste assay station composed of a /sup 252/Cf Shuffler, a gamma-ray scanner, and a screening station be tested and evaluated at a nuclear waste site. 34 figures, 15 tables.

  9. Advanced neutron absorber materials

    DOE Patents [OSTI]

    Branagan, Daniel J.; Smolik, Galen R.

    2000-01-01

    A neutron absorbing material and method utilizing rare earth elements such as gadolinium, europium and samarium to form metallic glasses and/or noble base nano/microcrystalline materials, the neutron absorbing material having a combination of superior neutron capture cross sections coupled with enhanced resistance to corrosion, oxidation and leaching.

  10. Arsenic activation neutron detector

    DOE Patents [OSTI]

    Jacobs, E.L.

    1980-01-28

    A detector of bursts of neutrons from a deuterium-deuteron reaction includes a quantity of arsenic adjacent a gamma detector such as a scintillator and photomultiplier tube. The arsenic is activated by the 2.5-MeV neutrons to release gamma radiation which is detected to give a quantitative representation of detected neutrons.

  11. 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.

  12. 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 beams 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 beams 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 models 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.

  13. VORTEX CREEP AGAINST TOROIDAL FLUX LINES, CRUSTAL ENTRAINMENT, AND PULSAR GLITCHES

    SciTech Connect (OSTI)

    Ggercino?lu, Erbil; Alpar, M. Ali E-mail: alpar@sabanciuniv.edu

    2014-06-10

    A region of toroidally oriented quantized flux lines must exist in the proton superconductor in the core of the neutron star. This region will be a site of vortex pinning and creep. Entrainment of the neutron superfluid with the crustal lattice leads to a requirement of superfluid moment of inertia associated with vortex creep in excess of the available crustal moment of inertia. This will bring about constraints on the equation of state. The toroidal flux region provides the moment of inertia necessary to complement the crust superfluid with postglitch relaxation behavior fitting the observations.

  14. Method and apparatus for determining vertical heat flux of geothermal field

    DOE Patents [OSTI]

    Poppendiek, Heinz F.

    1982-01-01

    A method and apparatus for determining vertical heat flux of a geothermal field, and mapping the entire field, is based upon an elongated heat-flux transducer (10) comprised of a length of tubing (12) of relatively low thermal conductivity with a thermopile (20) inside for measuring the thermal gradient between the ends of the transducer after it has been positioned in a borehole for a period sufficient for the tube to reach thermal equilibrium. The transducer is thermally coupled to the surrounding earth by a fluid annulus, preferably water or mud. A second transducer comprised of a length of tubing of relatively high thermal conductivity is used for a second thermal gradient measurement. The ratio of the first measurement to the second is then used to determine the earth's thermal conductivity, k.sub..infin., from a precalculated graph, and using the value of thermal conductivity thus determined, then determining the vertical earth temperature gradient, b, from predetermined steady state heat balance equations which relate the undisturbed vertical earth temperature distributions at some distance from the borehole and earth thermal conductivity to the temperature gradients in the transducers and their thermal conductivity. The product of the earth's thermal conductivity, k.sub..infin., and the earth's undisturbed vertical temperature gradient, b, then determines the earth's vertical heat flux. The process can be repeated many times for boreholes of a geothermal field to map vertical heat flux.

  15. Personnel neutron dose assessment upgrade: Volume 1, Personnel neutron dosimetry assessment: (Final report)

    SciTech Connect (OSTI)

    Hadlock, D.E.; Brackenbush, L.W.; Griffith, R.V.; Hankins, D.E.; Parkhurst, M.A.; Stroud, C.M.; Faust, L.G.; Vallario, E.J.

    1988-07-01

    This report provides guidance on the characteristics, use, and calibration criteria for personnel neutron dosimeters. The report is applicable for neutrons with energies ranging from thermal to less than 20 MeV. Background for general neutron dosimetry requirements is provided, as is relevant federal regulations and other standards. The characteristics of personnel neutron dosimeters are discussed, with particular attention paid to passive neutron dosimetry systems. Two of the systems discussed are used at DOE and DOE-contractor facilities (nuclear track emulsion and thermoluminescent-albedo) and another (the combination TLD/TED) was recently developed. Topics discussed in the field applications of these dosimeters include their theory of operation, their processing, readout, and interpretation, and their advantages and disadvantages for field use. The procedures required for occupational neutron dosimetry are discussed, including radiation monitoring and the wearing of dosimeters, their exchange periods, dose equivalent evaluations, and the documenting of neutron exposures. The coverage of dosimeter testing, maintenance, and calibration includes guidance on the selection of calibration sources, the effects of irradiation geometries, lower limits of detectability, fading, frequency of calibration, spectrometry, and quality control. 49 refs., 6 figs., 8 tabs.

  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. Semiconductor neutron detector

    DOE Patents [OSTI]

    Ianakiev, Kiril D.; Littlewood, Peter B.; Blagoev, Krastan B.; Swinhoe, Martyn T.; Smith, James L.; Sullivan, Clair J.; Alexandrov, Boian S.; Lashley, Jason Charles

    2011-03-08

    A neutron detector has a compound of lithium in a single crystal form as a neutron sensor element. The lithium compound, containing improved charge transport properties, is either lithium niobate or lithium tantalate. The sensor element is in direct contact with a monitor that detects an electric current. A signal proportional to the electric current is produced and is calibrated to indicate the neutrons sensed. The neutron detector is particularly useful for detecting neutrons in a radiation environment. Such radiation environment may, e.g. include gamma radiation and noise.

  18. Neutron scatter camera

    DOE Patents [OSTI]

    Mascarenhas, Nicholas; Marleau, Peter; Brennan, James S.; Krenz, Kevin D.

    2010-06-22

    An instrument that will directly image the fast fission neutrons from a special nuclear material source has been described. This instrument can improve the signal to background compared to non imaging neutron detection techniques by a factor given by ratio of the angular resolution window to 4.pi.. In addition to being a neutron imager, this instrument will also be an excellent neutron spectrometer, and will be able to differentiate between different types of neutron sources (e.g. fission, alpha-n, cosmic ray, and D-D or D-T fusion). Moreover, the instrument is able to pinpoint the source location.

  19. High energy neutron dosimeter

    DOE Patents [OSTI]

    Rai, K.S.F.

    1994-01-11

    A device for measuring dose equivalents in neutron radiation fields is described. The device includes nested symmetrical hemispheres (forming spheres) of different neutron moderating materials that allow the measurement of dose equivalents from 0.025 eV to past 1 GeV. The layers of moderating material surround a spherical neutron counter. The neutron counter is connected by an electrical cable to an electrical sensing means which interprets the signal from the neutron counter in the center of the moderating spheres. The spherical shape of the device allows for accurate measurement of dose equivalents regardless of its positioning. 2 figures.

  20. PERSONNEL NEUTRON DOSIMETER

    DOE Patents [OSTI]

    Fitzgerald, J.J.; Detwiler, C.G. Jr.

    1960-05-24

    A description is given of a personnel neutron dosimeter capable of indicating the complete spectrum of the neutron dose received as well as the dose for each neutron energy range therein. The device consists of three sets of indium foils supported in an aluminum case. The first set consists of three foils of indium, the second set consists of a similar set of indium foils sandwiched between layers of cadmium, whereas the third set is similar to the second set but is sandwiched between layers of polyethylene. By analysis of all the foils the neutron spectrum and the total dose from neutrons of all energy levels can be ascertained.

  1. High energy neutron dosimeter

    DOE Patents [OSTI]

    Sun, Rai Ko S.F. (Albany, CA)

    1994-01-01

    A device for measuring dose equivalents in neutron radiation fields. The device includes nested symmetrical hemispheres (forming spheres) of different neutron moderating materials that allow the measurement of dose equivalents from 0.025 eV to past 1 GeV. The layers of moderating material surround a spherical neutron counter. The neutron counter is connected by an electrical cable to an electrical sensing means which interprets the signal from the neutron counter in the center of the moderating spheres. The spherical shape of the device allows for accurate measurement of dose equivalents regardless of its positioning.

  2. 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

    2014-09-01

    Flux detector evaluations were completed as part of a joint Idaho State University (ISU) / Idaho National Laboratory (INL) / French Atomic Energy commission (CEA) ATR National Scientific User Facility (ATR NSUF) project to compare the accuracy, response time, and long duration performance of several flux detectors. Special fixturing developed by INL allows real-time flux detectors to be inserted into various ATRC core positions and perform lobe power measurements, axial flux profile measurements, and detector cross-calibrations. Detectors initially evaluated in this program include the French Atomic Energy Commission (CEA)-developed miniature fission chambers; specialized self-powered neutron detectors (SPNDs) developed by the Argentinean National Energy Commission (CNEA); specially developed commercial SPNDs from Argonne National Laboratory. As shown in this article, data obtained from this program provides important insights related to flux detector accuracy and resolution for subsequent ATR and CEA experiments and flux data required for bench-marking models in the ATR V&V Upgrade Initiative.

  3. The distribution of ion orbit loss fluxes of ions and energy from the plasma edge across the last closed flux surface into the scrape-off layer

    SciTech Connect (OSTI)

    Stacey, Weston M.; Schumann, Matthew T.

    2015-04-15

    A more detailed calculation strategy for the evaluation of ion orbit loss of thermalized plasma ions in the edge of tokamaks is presented. In both this and previous papers, the direct loss of particles from internal flux surfaces is calculated from the conservation of canonical angular momentum, energy, and magnetic moment. The previous result that almost all of the ion energy and particle fluxes crossing the last closed flux surface are in the form of ion orbit fluxes is confirmed, and the new result that the distributions of these fluxes crossing the last closed flux surface into the scrape-off layer are very strongly peaked about the outboard midplane is demonstrated. Previous results of a preferential loss of counter current particles leading to a co-current intrinsic rotation peaking just inside of the last closed flux surface are confirmed. Various physical details are discussed.

  4. High Dose Neutron Irradiation of Hi-Nicalon Type S Silicon Carbide Composites, Part 2. Mechanical and Physical Properties

    SciTech Connect (OSTI)

    Katoh, Yutai; Nozawa, Takashi; Shih, Chunghao Phillip; Ozawa, Kazumi; Koyanagi, Takaaki; Porter, Wallace D; Snead, Lance Lewis

    2015-01-07

    Nuclear-grade silicon carbide (SiC) composite material was examined for mechanical and thermophysical properties following high-dose neutron irradiation in the High Flux Isotope Reactor at a temperature range of 573–1073 K. Likewise, the material was chemical vapor-infiltrated SiC-matrix composite with a two-dimensional satin weave Hi-Nicalon Type S SiC fiber reinforcement and a multilayered pyrocarbon/SiC interphase. Moderate (1073 K) to very severe (573 K) degradation in mechanical properties was found after irradiation to >70 dpa, whereas no evidence was found for progressive evolution in swelling and thermal conductivity. The swelling was found to recover upon annealing beyond the irradiation temperature, indicating the irradiation temperature, but only to a limited extent. Moreover, the observed strength degradation is attributed primarily to fiber damage for all irradiation temperatures, particularly a combination of severe fiber degradation and likely interphase damage at relatively low irradiation temperatures.

  5. NEUTRON DENSITY CONTROL IN A NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Young, G.J.

    1959-06-30

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

  6. Spatial corrections for pulsed-neutron reactivity measurements.

    SciTech Connect (OSTI)

    Cao, Y.; Lee, J.; Nuclear Engineering Division; Univ. of Michigan

    2010-07-01

    For pulsed-neutron experiments performed in a subcritical reactor, the reactivity obtained from the area-ratio method is sensitive to detector positions. The spatial effects are induced by the presence of both the prompt neutron harmonics and the delayed neutron harmonics in the reactor. The traditional kinetics distortion factor is only limited to correcting the spatial effects caused by the fundamental prompt-{alpha} mode. In this paper, we derive spatial correction factors fp and fd to account for spatial effects induced by the prompt neutron harmonics and the delayed neutron harmonics, respectively. Our numerical simulations with the FX2-TH time-dependent multigroup diffusion code indicate that the high-order prompt neutron harmonics lead to significant spatial effects and cannot be neglected in calculating the spatial correction factors. The prompt spatial correction factor fp can be simply determined by the ratio of the normalized detector responses corresponding to the fundamental k-mode and the prompt neutron flux integrated over the pulse period. Thus, it is convenient to calculate and provides physically intuitive explanations on the spatial dependence of reactivity measured in the MUSE-4 experiments: overestimation of the subcriticality in regions close to the external neutron source and underestimation of the subcriticality away from the source but within the fuel region.

  7. Final Report for the “WSU Neutron Capture Therapy Facility Support”

    SciTech Connect (OSTI)

    Gerald E. Tripard; Keith G. Fox

    2006-08-24

    The objective for the cooperative research program for which this report has been written was to provide separate NCT facility user support for the students, faculty and scientists who would be doing the U.S. Department of Energy Office (DOE) of Science supported advanced radiotargeted research at the WSU 1 megawatt TRIGA reactor. The participants were the Idaho National laboratory (INL, P.I., Dave Nigg), the Veterinary Medical Research Center of Washington State University (WSU, Janean Fidel and Patrick Gavin), and the Washington State University Nuclear Radiation Center (WSU, P.I., Gerald Tripard). A significant number of DOE supported modifications were made to the WSU reactor in order to create an epithermal neutron beam while at the same time maintaining the other activities of the 1 MW reactor. These modifications were: (1) Removal of the old thermal column. (2) Construction and insertion of a new epithermal filter, collimator and shield. (3) Construction of a shielded room that could accommodate the very high radiation field created by an intense neutron beam. (4) Removal of the previous reactor core fuel cluster arrangement. (5) Design and loading of the new reactor core fuel cluster arrangement in order to optimize the neutron flux entering the epithermal neutron filter. (6) The integration of the shielded rooms interlocks and radiological controls into the SCRAM chain and operating electronics of the reactor. (7) Construction of a motorized mechanism for moving and remotely controlling the position of the entire reactor bridge. (8) The integration of the reactor bridge control electronics into the SCRAM chain and operating electronics of the reactor. (9) The design, construction and attachment to the support structure of the reactor of an irradiation box that could be inserted into position next to the face of the reactor. (Necessitated by the previously mentioned core rearrangement). All of the above modifications were successfully completed and tested. The resulting epithermal beam of 1 x 10{sup 9} n/sec-cm{sup 2} was measured by Idaho National Laboratory with assistance from WSU's Neutron Activation Analysis Group. The beam is as good as our initial proposals for the project had predicted. In addition to all of the design, construction and insertion of the hardware, shielding, electronics and radiation monitoring systems there was considerable manpower and effort put into changes in the Technical Specifications of the reactor and implementing procedures for use of the new facility. This staff involvement is one of the reasons we requested special facility support from the DOE. Once the facility was competed and all of the recalibrations and measurements made to characterize the differences between this reactor core and the previous core we began to assist INL in making their beam measurements with foils and phantoms. Although we proposed support for only one additional staff position to support this new NCT facility the staff support provided by the WSU Nuclear Radiation Center was greater than had been anticipated by our initial proposal. INL was also assisted in the testing of a heavy water (deuterated water) bladder that can be inserted into the collimator in order to produce an intense, external thermal neutron beam. The external epithermal and/or thermal neutron beam capability remains available for use, if funding becomes available for future research projects.

  8. NEUTRON REACTOR HAVING A Xe$sup 135$ SHIELD

    DOE Patents [OSTI]

    Stanton, H.E.

    1957-10-29

    Shielding for reactors of the type in which the fuel is a chain reacting liquid composition comprised essentially of a slurry of fissionable and fertile material suspended in a liquid moderator is discussed. The neutron reflector comprises a tank containing heavy water surrounding the reactor, a shield tank surrounding the reflector, a gamma ray shield surrounding said shield tank, and a means for conveying gaseous fission products, particularly Xe/sup 135/, from the reactor chamber to the shield tank, thereby serving as a neutron shield by capturing the thermalized neutrons that leak outwardly from the shield tank.

  9. Thermal Sciences

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

    Thermal Sciences NETL's Thermal Sciences competency provides the scientific, engineering, and technology development community with innovative and efficient approaches to measure, harness, and convert thermal energy. Research includes sensors, advanced energy concepts, and thermodynamic optimization, specifically: Sensors and Diagnostics Advanced sensor and diagnostic technology to develop and evaluate advanced methods for non-intrusive measurement and measurement in extreme environments.

  10. Lujan Neutron Scattering Center

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

    responds to radiological incident August 27, 2012 The Laboratory is investigating the inadvertent spread of Technetium 99 by employees and contractors at the Lujan Neutron Scattering Center August 27, 2012-The Laboratory is investigating the inadvertent spread of Technetium 99 by employees and contractors at the Lujan Neutron Scattering Center at the Los Alamos Neutron Science Center (LANSCE), a multidisciplinary accelerator facility used for both civilian and national security research. The

  11. ARM - Measurement - Soil moisture flux

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

    instrument for a list of all available measurements, including those recorded for diagnostic or quality assurance purposes. ARM Instruments CO2FLX : Carbon Dioxide Flux...

  12. Ion Beam Analysis of Targets Used in Controlatron Neutron Generators

    SciTech Connect (OSTI)

    Banks, James C.; Doyle, Barney L.; Walla, Lisa A.; Walsh, David S.

    2009-03-10

    Controlatron neutron generators are used for testing neutron detection systems at Sandia National Laboratories. To provide for increased tube lifetimes for the moderate neutron flux output of these generators, metal hydride (ZrT{sub 2}) target fabrication processes have been developed. To provide for manufacturing quality control of these targets, ion beam analysis techniques are used to determine film composition. The load ratios (i.e. T/Zr concentration ratios) of ZrT{sub 2} Controlatron neutron generator targets have been successfully measured by simultaneously acquiring RBS and ERD data using a He{sup ++} beam energy of 10 MeV. Several targets were measured and the film thicknesses obtained from RBS measurements agreed within {+-}2% with Dektak profilometer measurements. The target fabrication process and ion beam analysis techniques will be presented.

  13. Fast Neutron Detection Evaluation

    SciTech Connect (OSTI)

    McKigney, Edward A.; Stange, Sy

    2014-03-17

    These slides present a summary of previous work, conclusions, and anticipated schedule for the conclusion of our fast neutron detection evaluation.

  14. Neutron detection apparatus

    DOE Patents [OSTI]

    Kopp, Manfred K.; Valentine, Kenneth H.

    1983-01-01

    An atomic fission counting apparatus used for neutron detection is provided with spirally curved electrode plates uniformly spaced apart in a circular array and coated with fissile material.

  15. Magnetization of neutron matter

    SciTech Connect (OSTI)

    Bigdeli, M.

    2011-09-21

    In this paper, we compute magnetization of neutron matter at strong magnetic field using the lowest order constrained variational (LOCV) technique.

  16. LOS ALAMOS NEUTRON SCIENCE CENTER CONTRIBUTIONS TO THE DEVELOPMENT OF FUTURE POWER REACTORS

    SciTech Connect (OSTI)

    GAVRON, VICTOR I.; HILL, TONY S.; PITCHER, ERIC J.; TOVESSON, FREDERIK K.

    2007-01-09

    The Los Alamos Neutron Science Center (LANSCE) is a large spallation neutron complex centered around an 800 MeV high-currently proton accelerator. Existing facilities include a highly-moderated neutron facility (Lujan Center) where neutrons between thermal and keV energies are produced, and the Weapons Neutron Research Center (WNR), where a bare spallation target produces neutrons between 0.1 and several hundred MeV.The LANSCE facility offers a unique capability to provide high precision nuclear data over a large energy region, including that for fast reactor systems. In an ongoing experimental program the fission and capture cross sections are being measured for a number of minor actinides relevant for Generation-IV reactors and transmutation technology. Fission experiments makes use of both the highly moderated spallation neutron spectrum at the Lujan Center, and the unmoderated high energy spectrum at WNR. By combininb measurements at these two facilities the differential fission cross section is measured relative to the {sup 235}U(n,f) standard from subthermal energies up to about 200 MeV. An elaborate data acquisition system is designed to deal with all the different types of background present when spanning 10 energy decades. The first isotope to be measured was {sup 237}Np, and the results were used to improve the current ENDF/B-VII evaluation. Partial results have also been obtained for {sup 240}Pu and {sup 242}Pu, and the final results are expected shortly. Capture cross sections are measured at LANSCE using the Detector for Advanced Neutron Capture Experiments (DANCE). This unique instrument is highly efficient in detecting radiative capture events, and can thus handle radioactive samples of half-lives as low as 100 years. A number of capture cross sections important to fast reaction applications have been measured with DANCE. The first measurement was on {sup 237}Np(n,{gamma}), and the results have been submitted for publication. Other capture measurements in progress include {sup 240}Pu and {sup 242}Pu. The United States recently announced the Global Nuclear Energy Partnership (GNEP), with the goal of closing the commercial nuclear fuel cycle while minimizing proliferation risk. GNEP achieves these goals using fast-spectrum nuclear reactors powered by new transmutation fuels that contain significant quantities of minor actinides. The proposed Materials Test Station (MTS) will provide the GNEP with a cost-effective means of obtaining domestic fast-spectrum irradiations of advanced transmutation fuel forms and structural materials, which is an important step in the fuels qualification process. The MTS will be located at the LANSCE, and will be driven by a 1.08-MW proton beam. Th epeak neutron flux in the irradiation region is 1.67 x 10{sup 15} n/cm{sup 2}/s, and the energy spectrum is similar to that of a fast reactor, with the addition of a high-energy tail. The facility is expected to operate at least 4,400 hours per year. Fuel burnup rates will exceed 4% per year, and the radiation damage rate in iron will be 18 dpa (displacements per atom) per year. The construction cost is estimated to be $73M (including 25% contingency), with annual operating costs in the range of $6M to $10M. Appropriately funded, the MTS could begin operation in 2010.

  17. Neutron reactions in accreting neutron stars: a new pathway to...

    Office of Scientific and Technical Information (OSTI)

    a new pathway to efficient crust heating Citation Details In-Document Search Title: Neutron reactions in accreting neutron stars: a new pathway to efficient crust heating You ...

  18. Irradiation effect on deuterium behaviour in low-dose HFIR neutron-irradiated tungsten

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

    Shimada, Masashi; Cao, G.; Otsuka, T.; Hara, M.; Kobayashi, M.; Oya, Y.; Hatano, Y.

    2014-12-01

    Tungsten samples were irradiated by neutrons in the High Flux Isotope Reactor, Oak Ridge National Laboratory at reactor coolant temperatures of 50-70°C to low displacement damage of 0.025 and 0.3 dpa under the framework of the US-Japan TITAN program (2007-2013). After cooling down, the HFIR neutron-irradiated tungsten samples were exposed to deuterium plasmas in the Tritium Plasma Experiment, Idaho National Laboratory at 100, 200 and 500 °C twice at the ion fluence of 5×10²⁵ m⁻² to reach a total ion fluence of 1×10²⁶ m⁻² in order to investigate the near surface deuterium retention and saturation via nuclear reaction analysis. Finalmore » thermal desorption spectroscopy was performed to elucidate irradiation effect on total deuterium retention. Nuclear reaction analysis results showed that the maximum near surface (<5 µm depth) deuterium concentration increased from 0.5 at % D/W in 0.025 dpa samples to 0.8 at. % D/W in 0.3 dpa samples. The large discrepancy between the total retention via thermal desorption spectroscopy and the near surface retention via nuclear reaction analysis indicated the deuterium was migrated and trapped in bulk (at least 50 µm depth for 0.025 dpa and 35 µm depth for 0.025 dpa) at 500 °C case even in the relatively low ion fluence of 10²⁶ m⁻².« less

  19. Neutron and Nuclear Science News

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

    News Recent news and events related to neutron and nuclear science at LANSCE. Neutron and Nuclear Science News Links Neutron and Nuclear Science News Media Links Profiles Events at LANSCE

  20. Neutronic fuel element fabrication

    DOE Patents [OSTI]

    Korton, George

    2004-02-24

    This disclosure describes a method for metallurgically bonding a complete leak-tight enclosure to a matrix-type fuel element penetrated longitudinally by a multiplicity of coolant channels. Coolant tubes containing solid filler pins are disposed in the coolant channels. A leak-tight metal enclosure is then formed about the entire assembly of fuel matrix, coolant tubes and pins. The completely enclosed and sealed assembly is exposed to a high temperature and pressure gas environment to effect a metallurgical bond between all contacting surfaces therein. The ends of the assembly are then machined away to expose the pin ends which are chemically leached from the coolant tubes to leave the coolant tubes with internal coolant passageways. The invention described herein was made in the course of, or under, a contract with the U.S. Atomic Energy Commission. It relates generally to fuel elements for neutronic reactors and more particularly to a method for providing a leak-tight metal enclosure for a high-performance matrix-type fuel element penetrated longitudinally by a multiplicity of coolant tubes. The planned utilization of nuclear energy in high-performance, compact-propulsion and mobile power-generation systems has necessitated the development of fuel elements capable of operating at high power densities. High power densities in turn require fuel elements having high thermal conductivities and good fuel retention capabilities at high temperatures. A metal clad fuel element containing a ceramic phase of fuel intimately mixed with and bonded to a continuous refractory metal matrix has been found to satisfy the above requirements. Metal coolant tubes penetrate the matrix to afford internal cooling to the fuel element while providing positive fuel retention and containment of fission products generated within the fuel matrix. Metal header plates are bonded to the coolant tubes at each end of the fuel element and a metal cladding or can completes the fuel-matrix enclosure by encompassing the sides of the fuel element between the header plates.

  1. Final Report on Actinide Glass Scintillators for Fast Neutron Detection

    SciTech Connect (OSTI)

    Bliss, Mary; Stave, Jean A.

    2012-10-01

    This is the final report of an experimental investigation of actinide glass scintillators for fast-neutron detection. It covers work performed during FY2012. This supplements a previous report, PNNL-20854 “Initial Characterization of Thorium-loaded Glasses for Fast Neutron Detection” (October 2011). The work in FY2012 was done with funding remaining from FY2011. As noted in PNNL-20854, the glasses tested prior to July 2011 were erroneously identified as scintillators. The decision was then made to start from “scratch” with a literature survey and some test melts with a non-radioactive glass composition that could later be fabricated with select actinides, most likely thorium. The normal stand-in for thorium in radioactive waste glasses is cerium in the same oxidation state. Since cerium in the 3+ state is used as the light emitter in many scintillating glasses, the next most common substitute was used: hafnium. Three hafnium glasses were melted. Two melts were colored amber and a third was clear. It barely scintillated when exposed to alpha particles. The uses and applications for a scintillating fast neutron detector are important enough that the search for such a material should not be totally abandoned. This current effort focused on actinides that have very high neutron capture energy releases but low neutron capture cross sections. This results in very long counting times and poor signal to noise when working with sealed sources. These materials are best for high flux applications and access to neutron generators or reactors would enable better test scenarios. The total energy of the neutron capture reaction is not the only factor to focus on in isotope selection. Many neutron capture reactions result in energetic gamma rays that require large volumes or high densities to detect. If the scintillator is to separate neutrons from gamma rays, the capture reactions should produce heavy particles and few gamma rays. This would improve the detection of a signal for fast neutron capture.

  2. The temperature dependence of ultra-cold neutron wall losses in material

    Office of Scientific and Technical Information (OSTI)

    bottles coated with deuterated polystryene (Conference) | SciTech Connect The temperature dependence of ultra-cold neutron wall losses in material bottles coated with deuterated polystryene Citation Details In-Document Search Title: The temperature dependence of ultra-cold neutron wall losses in material bottles coated with deuterated polystryene Ultra-cold neutrons (UCN) from the LANSCE super-thermal deuterium source were used to fill an acrylic bottle coated with deuterated polystyrene.

  3. Schoenborn wins Bau Neutron Award

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

    of LANL's Bioenergy and Biome Sciences group, to receive the 2016 Bau Neutron Diffraction Award. The award recognizes exceptional research achievement in neutron...

  4. Neutron and Nuclear Science News

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

    News Recent news and events related to neutron and nuclear science at LANSCE. Neutron and Nuclear Science News Nuclear and Materials Science Research at LANSCE Nuclear science...

  5. 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.

  6. Neutron monitoring systems including gamma thermometers and methods of calibrating nuclear instruments using gamma thermometers

    DOE Patents [OSTI]

    Moen, Stephan Craig; Meyers, Craig Glenn; Petzen, John Alexander; Foard, Adam Muhling

    2012-08-07

    A method of calibrating a nuclear instrument using a gamma thermometer may include: measuring, in the instrument, local neutron flux; generating, from the instrument, a first signal proportional to the neutron flux; measuring, in the gamma thermometer, local gamma flux; generating, from the gamma thermometer, a second signal proportional to the gamma flux; compensating the second signal; and calibrating a gain of the instrument based on the compensated second signal. Compensating the second signal may include: calculating selected yield fractions for specific groups of delayed gamma sources; calculating time constants for the specific groups; calculating a third signal that corresponds to delayed local gamma flux based on the selected yield fractions and time constants; and calculating the compensated second signal by subtracting the third signal from the second signal. The specific groups may have decay time constants greater than 5.times.10.sup.-1 seconds and less than 5.times.10.sup.5 seconds.

  7. 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

  8. 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.

  9. Compact neutron generator

    DOE Patents [OSTI]

    Leung, Ka-Ngo; Lou, Tak Pui

    2005-03-22

    A compact neutron generator has at its outer circumference a toroidal shaped plasma chamber in which a tritium (or other) plasma is generated. A RF antenna is wrapped around the plasma chamber. A plurality of tritium ion beamlets are extracted through spaced extraction apertures of a plasma electrode on the inner surface of the toroidal plasma chamber and directed inwardly toward the center of neutron generator. The beamlets pass through spaced acceleration and focusing electrodes to a neutron generating target at the center of neutron generator. The target is typically made of titanium tubing. Water is flowed through the tubing for cooling. The beam can be pulsed rapidly to achieve ultrashort neutron bursts. The target may be moved rapidly up and down so that the average power deposited on the surface of the target may be kept at a reasonable level. The neutron generator can produce fast neutrons from a T-T reaction which can be used for luggage and cargo interrogation applications. A luggage or cargo inspection system has a pulsed T-T neutron generator or source at the center, surrounded by associated gamma detectors and other components for identifying explosives or other contraband.

  10. Single-Volume Neutron Scatter Camera for High-Efficiency Neutron Imaging and Source Characterization. Year 2 of 3 Summary

    SciTech Connect (OSTI)

    Brubaker, Erik

    2015-10-01

    The neutron scatter camera (NSC), an imaging spectrometer for fission energy neutrons, is an established and proven detector for nuclear security applications such as weak source detection of special nuclear material (SNM), arms control treaty verification, and emergency response. Relative to competing technologies such as coded aperture imaging, time-encoded imaging, neutron time projection chamber, and various thermal neutron imagers, the NSC provides excellent event-by-event directional information for signal/background discrimination, reasonable imaging resolution, and good energy resolution. Its primary drawback is very low detection efficiency due to the requirement for neutron elastic scatters in two detector cells. We will develop a singlevolume double-scatter neutron imager, in which both neutron scatters can occur in the same large active volume. If successful, the efficiency will be dramatically increased over the current NSC cell-based geometry. If the detection efficiency approaches that of e.g. coded aperture imaging, the other inherent advantages of double-scatter imaging would make it the most attractive fast neutron detector for a wide range of security applications.

  11. Pocked surface neutron detector

    DOE Patents [OSTI]

    McGregor, Douglas; Klann, Raymond

    2003-04-08

    The detection efficiency, or sensitivity, of a neutron detector material such as of Si, SiC, amorphous Si, GaAs, or diamond is substantially increased by forming one or more cavities, or holes, in its surface. A neutron reactive material such as of elemental, or any compound of, .sup.10 B, .sup.6 Li, .sup.6 LiF, U, or Gd is deposited on the surface of the detector material so as to be disposed within the cavities therein. The portions of the neutron reactive material extending into the detector material substantially increase the probability of an energetic neutron reaction product in the form of a charged particle being directed into and detected by the neutron detector material.

  12. NEUTRON SHIELDING STRUCTURE

    DOE Patents [OSTI]

    Mattingly, J.T.

    1962-09-25

    A lightweight neutron shielding structure comprises a honeycomb core which is filled with a neutron absorbing powder. The honeycomb core is faced with parallel planar facing sheets to form a lightweight rigid unit. Suitable absorber powders are selected from among the following: B, B/sub 4/C, B/sub 2/O/ sub 3/, CaB/sub 6/, Li/sub 2/CO3, LiOH, LiBO/sub 2/, Li/s ub 2/O. The facing sheets are constructed of a neutron moderating material, so that fast neutrons will be moderated while traversing the facing sheets, and ultimately be absorbed by the absorber powder in the honeycomb. Beryllium is a preferred moderator material for use in the facing sheets. The advantage of the structure is that it combines the rigidity and light weight of a honeycomb construction with the neutron absorption properties of boron and lithium. (AEC)

  13. Pulsed neutron detector

    DOE Patents [OSTI]

    Robertson, deceased, J. Craig; Rowland, Mark S.

    1989-03-21

    A pulsed neutron detector and system for detecting low intensity fast neutron pulses has a body of beryllium adjacent a body of hydrogenous material the latter of which acts as a beta particle detector, scintillator, and moderator. The fast neutrons (defined as having En>1.5 MeV) react in the beryllium and the hydrogenous material to produce larger numbers of slow neutrons than would be generated in the beryllium itself and which in the beryllium generate hellium-6 which decays and yields beta particles. The beta particles reach the hydrogenous material which scintillates to yield light of intensity related to the number of fast neutrons. A photomultiplier adjacent the hydrogenous material (scintillator) senses the light emission from the scintillator. Utilization means, such as a summing device, sums the pulses from the photo-multiplier for monitoring or other purposes.

  14. RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - models and correlations

    SciTech Connect (OSTI)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

    1998-03-01

    This document describes the major modifications and improvements made to the modeling of the RAMONA-3B/MOD0 code since 1981, when the code description and assessment report was completed. The new version of the code is RAMONA-4B. RAMONA-4B is a systems transient code for application to different versions of Boiling Water Reactors (BWR) such as the current BWR, the Advanced Boiling Water Reactor (ABWR), and the Simplified Boiling Water Reactor (SBWR). This code uses a three-dimensional neutron kinetics model coupled with a multichannel, non-equilibrium, drift-flux, two-phase flow formulation of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients and instability issues. Chapter 1 is an overview of the code`s capabilities and limitations; Chapter 2 discusses the neutron kinetics modeling and the implementation of reactivity edits. Chapter 3 is an overview of the heat conduction calculations. Chapter 4 presents modifications to the thermal-hydraulics model of the vessel, recirculation loop, steam separators, boron transport, and SBWR specific components. Chapter 5 describes modeling of the plant control and safety systems. Chapter 6 presents and modeling of Balance of Plant (BOP). Chapter 7 describes the mechanistic containment model in the code. The content of this report is complementary to the RAMONA-3B code description and assessment document. 53 refs., 81 figs., 13 tabs.

  15. Design of a differential radiometer for atmospheric radiative flux measurements

    SciTech Connect (OSTI)

    LaDelfe, P.C.; Weber, P.G.; Rodriguez, C.W.

    1994-11-01

    The Hemispherical Optimized NEt Radiometer (HONER) is an instrument under development at the Los Alamos National Laboratory for deployment on an unmanned aerospace vehicle as part of the Atmospheric Radiation Measurements (ARM/UAV) program. HONER is a differential radiometer which will measure the difference between the total upwelling and downwelling fluxes and is intended to provide a means of measuring the atmospheric radiative flux divergence. Unlike existing instruments which measure the upwelling and downwelling fluxes separately, HONER will achieve an optical difference by chopping the two fluxes alternately onto a common pyroelectric detector. HONER will provide data resolved into two spectral bands; one covering the solar dominated region from less than 0.4 micrometer to approximately 4.5 micrometers and the other covering the region from approximately 4.5 micrometers to greater than 50 micrometers, dominated by thermal radiation. The means of separating the spectral regions guarantees seamless summation to calculate the total flux. The fields-of-view are near-hemispherical, upward and downward. The instrument can be converted, in flight, from the differential mode to absolute mode, measuring the upwelling and downwelling fluxes separately and simultaneously. The instrument also features continuous calibration from on-board sources. We will describe the design and operation of the sensor head and the on-board reference sources as well as the means of deployment.

  16. Application of neutron-activation analysis to geological materials

    SciTech Connect (OSTI)

    Laul, J.C.; Wogman, N.A.

    1980-12-01

    Neutron activation analysis (NAA) is an extremely sensitive, selective, and precise method, which yields a wealth of elemental information from even a small-sized sample. By varying neutron fluxes, irradiation times, decay and counting intervals in instrumental NAA, it is possible to accurately determine about 35 elements in a geological aliquot. When INAA is coupled with coincidence-noncoincidence Ge(Li)-Na(Tl) counting, it enhances the sensitivities of various elements by order of magnitude. The attractive features of INAA are that it is fast, nondestructive and economical.

  17. Solid state neutron detector and method for use

    DOE Patents [OSTI]

    Doty, F. Patrick; Zwieback, Ilya; Ruderman, Warren

    2002-01-01

    Crystals of lithium tetraborate or alpha-barium borate had been found to be neutron detecting materials. The crystals are prepared using known crystal growing techniques, wherein the process does not include the common practice of using a fluxing agent, such as sodium oxide or sodium fluoride, to reduce the melting temperature of the crystalline compound. Crystals prepared by this method can be sliced into thin single or polycrystalline wafers, or ground to a powder and prepared as a sintered compact or a print paste, and then configured with appropriate electronic hardware, in order to function as neutron detectors.

  18. A DIRECT MEASUREMENT OF THE HEAT RELEASE IN THE OUTER CRUST OF THE TRANSIENTLY ACCRETING NEUTRON STAR XTE J1709-267

    SciTech Connect (OSTI)

    Degenaar, N.; Miller, J. M.; Wijnands, R.

    2013-04-20

    The heating and cooling of transiently accreting neutron stars provides a powerful probe of the structure and composition of their crust. Observations of superbursts and cooling of accretion-heated neutron stars require more heat release than is accounted for in current models. Obtaining firm constraints on the depth and magnitude of this extra heat is challenging and therefore its origin remains uncertain. We report on Swift and XMM-Newton observations of the transient neutron star low-mass X-ray binary XTE J1709-267, which were made in 2012 September-October when it transitioned to quiescence after a {approx_equal}10 week long accretion outburst. The source is detected with XMM-Newton at a 0.5-10 keV luminosity of L{sub X} {approx_equal} 2 Multiplication-Sign 10{sup 34}(D/8.5 kpc){sup 2} erg s{sup -1}. The X-ray spectrum consists of a thermal component that fits to a neutron star atmosphere model and a non-thermal emission tail, each of which contribute {approx_equal}50% to the total flux. The neutron star temperature decreases from {approx_equal}158 to {approx_equal}152 eV during the {approx_equal}8 hr long observation. This can be interpreted as cooling of a crustal layer located at a column density of y {approx_equal} 5 Multiplication-Sign 10{sup 12} g cm{sup -2} ({approx_equal}50 m inside the neutron star), which is just below the ignition depth of superbursts. The required heat generation in the layers on top would be {approx_equal}0.06-0.13 MeV per accreted nucleon. The magnitude and depth rule out electron captures and nuclear fusion reactions as the heat source, but it may be accounted for by chemical separation of light and heavy nuclei. Low-level accretion offers an alternative explanation for the observed variability.

  19. How to produce a reactor neutron spectrum using a proton accelerator

    SciTech Connect (OSTI)

    Burns, Kimberly A.; Wootan, David W.; Gates, Robert O.; Schmitt, Bruce E.; Asner, David M.

    2015-01-01

    A method for reproducing the neutron energy spectrum present in the core of an operating nuclear reactor using an engineered target in an accelerator proton beam is proposed. The protons interact with a target to create neutrons through various (p,n) type reactions. Spectral tailoring of the emitted neutrons can be used to modify the energy of the generated neutron spectrum to represent various reactor spectra. Through the use of moderators and reflectors, the neutron spectrum can be modified to reproduce many different spectra of interest including spectra in small thermal test reactors, large pressurized water reactors, and fast reactors. The particular application of this methodology is the design of an experimental approach for using an accelerator to measure the betas produced during fission to be used to reduce uncertainties in the interpretation of reactor antineutrino measurements. This approach involves using a proton accelerator to produce a neutron field representative of a power reactor, and using this neutron field to irradiate fission foils of the primary isotopes contributing to fission in the reactor, creating unstable, neutron rich fission products that subsequently beta decay and emit electron antineutrinos. A major advantage of an accelerator neutron source over a neutron beam from a thermal reactor is that the fast neutrons can be slowed down or tailored to approximate various power reactor spectra. An accelerator based neutron source that can be tailored to match various reactor neutron spectra provides an advantage for control in studying how changes in the neutron spectra affect parameters such as the resulting fission product beta spectrum.

  20. Neutron and Nuclear Science Publications

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

    Publications Recent publications related to neutron and nuclear science at LANSCE. Neutron and Nuclear Science Publications Chi-Nu Publications DANCE Publications GEANIE Publications ICE House and ICE II Publications (n,z) Publications Neutron Radiography Publications SPIDER Publications Target 2 Publications TPC Publications Links Publications/Media Neutron and Nuclear Science News Profiles Events at LANSCE LANL Research Libary

  1. Traveling-wave device with mass flux suppression

    DOE Patents [OSTI]

    Swift, Gregory W.; Backhaus, Scott N.; Gardner, David L.

    2000-01-01

    A traveling-wave device is provided with the conventional moving pistons eliminated. Acoustic energy circulates in a direction through a fluid within a torus. A side branch may be connected to the torus for transferring acoustic energy into or out of the torus. A regenerator is located in the torus with a first heat exchanger located on a first side of the regenerator downstream of the regenerator relative to the direction of the circulating acoustic energy; and a second heat exchanger located on an upstream side of the regenerator. The improvement is a mass flux suppressor located in the torus to minimize time-averaged mass flux of the fluid. In one embodiment, the device further includes a thermal buffer column in the torus to thermally isolate the heat exchanger that is at the operating temperature of the device.

  2. SmAHTR-CTC Neutronic Design

    SciTech Connect (OSTI)

    Ilas, Dan; Holcomb, David Eugene; Gehin, Jess C

    2014-01-01

    Building on prior experience for the 2010 initial SmAHTR neutronic design and on 2012 neutronic design for the advanced high temperature reactor (AHTR), this paper presents the main results of the neutronic design effort for the newly re-purposed SmAHTR-CTC reactor concept. The results are obtained based on full-core simulations performed with SCALE6.1. The dimensionality of the SmAHTR design space is reduced by using constraints originating in material fabricability, fuel licensing, molten salt chemistry, thermal-hydraulic and mechanical considerations. The new design represents in many regards a substantial improvement from the neutronic performance standpoint over the 2010 SmAHTR concept. Among other results, it is shown that fuel cycle length of over 2 years or discharged fuel burnup of 40GWd/MTHM are possible with a low, 8% fuel enrichment in a once-through fuel cycle, while 8-year once-through fuel cycle lengths are possible at higher fuel enrichments.

  3. Vaporization, dispersion, and radiant fluxes from LPG spills. Final report

    SciTech Connect (OSTI)

    Not Available

    1981-12-01

    Both burning and non-burning spills of LPG (primarily propane) were studied. Vaporization rates for propane spills on soil, concrete, insulating concrete, asphalt, sod, wood, and polymer foams were measured. Thermal conductivity, heat transfer coefficients, and steady state vaporization rates were determined. Vapor concentrations were measured downwind of open propane pools 25, 100, 400, and 1600 ft/sup 2/ in area. A Gaussian dispersion model modified for area sources provided a good correlation of measured concentrations. Emitted and incident radiant fluxes from propane fires were measured. Simplified flame radiation models were adequate for predicting radiant fluxes; the maximum effective flux emitted at the flame surface was about 50,000 Btu/h-ft/sup 2/. A few tests in which propane was sprayed into the air showed that at moderately high spray rates all the propane flashed to vapor or atomized; no liquid collected on the ground.

  4. Neutron and gamma irradiation damage to organic materials.

    SciTech Connect (OSTI)

    White, Gregory Von, II; Bernstein, Robert

    2012-04-01

    This document discusses open literature reports which investigate the damage effects of neutron and gamma irradiation on polymers and/or epoxies - damage refers to reduced physical chemical, and electrical properties. Based on the literature, correlations are made for an SNL developed epoxy (Epon 828-1031/DDS) with an expected total fast-neutron fluence of {approx}10{sup 12} n/cm{sup 2} and a {gamma} dosage of {approx}500 Gy received over {approx}30 years at < 200 C. In short, there are no gamma and neutron irradiation concerns for Epon 828-1031/DDS. To enhance the fidelity of our hypotheses, in regards to radiation damage, we propose future work consisting of simultaneous thermal/irradiation (neutron and gamma) experiments that will help elucidate any damage concerns at these specified environmental conditions.

  5. Neutronics analyses in support of rotating target developments at SNS

    SciTech Connect (OSTI)

    Gallmeier, Franz X.

    2010-03-08

    A second target station (STS) for Spallation Neutron Souce (SNS) very likely being operated in long-pulse mode is in the early design phase, will complement the ORNL neutron sources, which presently consist of a short-pulse spallation source and the HFIR research reactor. As an alternative to the stationary liquid metal target, a rotating target is being considered. Neutronics studies in support of a 3MW power 20 Hz repetition rate rotating target feasibility study funded through the laboratory LDRD program, was extended towards a 1.5 MW STS design. The scope of work included in-operation heat deposition rates in target structures for thermal and structural analyses, target radionuclide inventory for decay heat and safety analyses, lifetime estimations due to radiation-driven material damage of target and moderator components, moderator neutron performance and moderator cryogenic heatloads.

  6. Neutron crystallography aids drug design

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

    Neutron crystallography aids drug design Neutron crystallography aids drug design Researchers have used neutron crystallography for the first time to determine the structure of a clinical drug in complex with its human target enzyme. October 9, 2012 Rich Marquez The Protein Crystallography Station at the Los Alamos Neutron Science Center, where groundbreaking work in new drug-design methods is underway using neutron diffraction techniques. Contact Nancy Ambrosiano Communications Office (505)

  7. Schoenborn wins Bau Neutron Award

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

    Schoenborn wins Bau Neutron Award Schoenborn wins Bau Neutron Award The American Crystallographic Association (ACA) has selected retired Laboratory Senior Fellow Benno Schoenborn to receive the 2016 Bau Neutron Diffraction Award. August 12, 2015 Benno Schoenborn Benno Schoenborn Communications Office (505) 667-7000 Schoenborn is honored for his pioneering research in macromolecular neutron crystallography and the design and development of the neutron crystallography beamline (Protein

  8. Switchable radioactive neutron source device

    DOE Patents [OSTI]

    Stanford, G.S.; Rhodes, E.A.; Devolpi, A.; Boyar, R.E.

    1987-11-06

    This invention is a switchable neutron generating apparatus comprised of a pair of plates, the first plate having an alpha emitter section on it and the second plate having a target material portion on it which generates neutrons when its nuclei absorb an alpha particle. In operation, the alpha portion of the first plate is aligned with the neutron portion of the second plate to produce neutrons and brought out of alignment to cease production of neutrons. 3 figs.

  9. Neutron resonance averaging

    SciTech Connect (OSTI)

    Chrien, R.E.

    1986-10-01

    The principles of resonance averaging as applied to neutron capture reactions are described. Several illustrations of resonance averaging to problems of nuclear structure and the distribution of radiative strength in nuclei are provided. 30 refs., 12 figs.

  10. 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

  11. 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

  12. A SEARCH FOR POINT SOURCES OF EeV NEUTRONS

    SciTech Connect (OSTI)

    Abreu, P.; Andringa, S.; Aglietta, M.; Ahlers, M.; Ahn, E. J.; Albuquerque, I. F. M.; Allard, D.; Allekotte, I.; Allen, J.; Allison, P.; Almela, A.; Alvarez Castillo, J.; Alvarez-Muniz, J.; Alves Batista, R.; Ambrosio, M.; Aramo, C.; Aminaei, A.; Anchordoqui, L.; Antici'c, T.; Arganda, E.; Collaboration: Pierre Auger Collaboration; and others

    2012-12-01

    A thorough search of the sky exposed at the Pierre Auger Cosmic Ray Observatory reveals no statistically significant excess of events in any small solid angle that would be indicative of a flux of neutral particles from a discrete source. The search covers from -90 Degree-Sign to +15 Degree-Sign in declination using four different energy ranges above 1 EeV (10{sup 18} eV). The method used in this search is more sensitive to neutrons than to photons. The upper limit on a neutron flux is derived for a dense grid of directions for each of the four energy ranges. These results constrain scenarios for the production of ultrahigh energy cosmic rays in the Galaxy.

  13. Neutron tomography of axially symmetric objects using 14 MeV neutrons from a portable neutron generator

    SciTech Connect (OSTI)

    Andersson, P. Andersson-Sunden, E.; Sjöstrand, H.; Jacobsson-Svärd, S.

    2014-08-01

    In nuclear boiling water reactor cores, the distribution of water and steam (void) is essential for both safety and efficiency reasons. In order to enhance predictive capabilities, void distribution assessment is performed in two-phase test-loops under reactor-relevant conditions. This article proposes the novel technique of fast-neutron tomography using a portable deuterium-tritium neutron generator to determine the time-averaged void distribution in these loops. Fast neutrons have the advantage of high transmission through the metallic structures and pipes typically concealing a thermal-hydraulic test loop, while still being fairly sensitive to the water/void content. However, commercially available fast-neutron generators also have the disadvantage of a relatively low yield and fast-neutron detection also suffers from relatively low detection efficiency. Fortunately, some loops are axially symmetric, a property which can be exploited to reduce the amount of data needed for tomographic measurement, thus limiting the interrogation time needed. In this article, three axially symmetric test objects depicting a thermal-hydraulic test loop have been examined; steel pipes with outer diameter 24 mm, thickness 1.5 mm, and with three different distributions of the plastic material POM inside the pipes. Data recorded with the FANTOM fast-neutron tomography instrument have been used to perform tomographic reconstructions to assess their radial material distribution. Here, a dedicated tomographic algorithm that exploits the symmetry of these objects has been applied, which is described in the paper. Results are demonstrated in 20 rixel (radial pixel) reconstructions of the interior constitution and 2D visualization of the pipe interior is demonstrated. The local POM attenuation coefficients in the rixels were measured with errors (RMS) of 0.025, 0.020, and 0.022 cm{sup −1}, solid POM attenuation coefficient. The accuracy and precision is high enough to provide a useful indication on the flow mode, and a visualization of the radial material distribution can be obtained. A benefit of this system is its potential to be mounted at any axial height of a two-phase test section without requirements for pre-fabricated entrances or windows. This could mean a significant increase in flexibility of the void distribution assessment capability at many existing two-phase test loops.

  14. MAGNETIC NEUTRON SCATTERING

    SciTech Connect (OSTI)

    ZALIZNYAK,I.A.; LEE,S.H.

    2004-07-30

    Much of our understanding of the atomic-scale magnetic structure and the dynamical properties of solids and liquids was gained from neutron-scattering studies. Elastic and inelastic neutron spectroscopy provided physicists with an unprecedented, detailed access to spin structures, magnetic-excitation spectra, soft-modes and critical dynamics at magnetic-phase transitions, which is unrivaled by other experimental techniques. Because the neutron has no electric charge, it is an ideal weakly interacting and highly penetrating probe of matter's inner structure and dynamics. Unlike techniques using photon electric fields or charged particles (e.g., electrons, muons) that significantly modify the local electronic environment, neutron spectroscopy allows determination of a material's intrinsic, unperturbed physical properties. The method is not sensitive to extraneous charges, electric fields, and the imperfection of surface layers. Because the neutron is a highly penetrating and non-destructive probe, neutron spectroscopy can probe the microscopic properties of bulk materials (not just their surface layers) and study samples embedded in complex environments, such as cryostats, magnets, and pressure cells, which are essential for understanding the physical origins of magnetic phenomena. Neutron scattering is arguably the most powerful and versatile experimental tool for studying the microscopic properties of the magnetic materials. The magnitude of the cross-section of the neutron magnetic scattering is similar to the cross-section of nuclear scattering by short-range nuclear forces, and is large enough to provide measurable scattering by the ordered magnetic structures and electron spin fluctuations. In the half-a-century or so that has passed since neutron beams with sufficient intensity for scattering applications became available with the advent of the nuclear reactors, they have became indispensable tools for studying a variety of important areas of modern science, ranging from large-scale structures and dynamics of polymers and biological systems, to electronic properties of today's technological materials. Neutron scattering developed into a vast field, encompassing many different experimental techniques aimed at exploring different aspects of matter's atomic structure and dynamics. Modern magnetic neutron scattering includes several specialized techniques designed for specific studies and/or particular classes of materials. Among these are magnetic reflectometry aimed at investigating surfaces, interfaces, and multilayers, small-angle scattering for the large-scale structures, such as a vortex lattice in a superconductor, and neutron spin-echo spectroscopy for glasses and polymers. Each of these techniques and many others offer exciting opportunities for examining magnetism and warrant extensive reviews, but the aim of this chapter is not to survey how different neutron-scattering methods are used to examine magnetic properties of different materials. Here, we concentrate on reviewing the basics of the magnetic neutron scattering, and on the recent developments in applying one of the oldest methods, the triple axis spectroscopy, that still is among the most extensively used ones. The developments discussed here are new and have not been coherently reviewed. Chapter 2 of this book reviews magnetic small-angle scattering, and modern techniques of neutron magnetic reflectometry are discussed in Chapter 3.

  15. Geek-Up[1.28.2011]: Neutron Scattering and Full-Spectrum Solar Cells |

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

    Department of Energy .28.2011]: Neutron Scattering and Full-Spectrum Solar Cells Geek-Up[1.28.2011]: Neutron Scattering and Full-Spectrum Solar Cells January 28, 2011 - 5:11pm Addthis Detector tanks for the new SANS instruments at the High Flux Isotope Reactor. The Bio-SANS detector is on the right. Source: ORNL Detector tanks for the new SANS instruments at the High Flux Isotope Reactor. The Bio-SANS detector is on the right. Source: ORNL Niketa Kumar Niketa Kumar Public Affairs Specialist,

  16. Design Studies for a Multiple Application Thermal Reactor for Irradiation Experiments (MATRIX)

    SciTech Connect (OSTI)

    Pope, Michael A.; Gougar, Hans D.; Ryskamp, J. M.

    2015-03-01

    The Advanced Test Reactor (ATR) is a high power density test reactor specializing in fuel and materials irradiation. For more than 45 years, the ATR has provided irradiations of materials and fuels testing along with radioisotope production. Should unforeseen circumstances lead to the decommissioning of ATR, the U.S. Government would be left without a large-scale materials irradiation capability to meet the needs of its nuclear energy and naval reactor missions. In anticipation of this possibility, work was performed under the Laboratory Directed Research and Development (LDRD) program to investigate test reactor concepts that could satisfy the current missions of the ATR along with an expanded set of secondary missions. A survey was conducted in order to catalogue the anticipated needs of potential customers. Then, concepts were evaluated to fill the role for this reactor, dubbed the Multi-Application Thermal Reactor Irradiation eXperiments (MATRIX). The baseline MATRIX design is expected to be capable of longer cycle lengths than ATR given a particular batch scheme. The volume of test space in In-Pile-Tubes (IPTs) is larger in MATRIX than in ATR with comparable magnitude of neutron flux. Furthermore, MATRIX has more locations of greater volume having high fast neutron flux than ATR. From the analyses performed in this work, it appears that the lead MATRIX design can be designed to meet the anticipated needs of the ATR replacement reactor. However, this design is quite immature, and therefore any requirements currently met must be re-evaluated as the design is developed further.

  17. Development of the prototype pneumatic transfer system for ITER neutron activation system

    SciTech Connect (OSTI)

    Cheon, M. S.; Seon, C. R.; Pak, S.; Lee, H. G.; Bertalot, L.

    2012-10-15

    The neutron activation system (NAS) measures neutron fluence at the first wall and the total neutron flux from the ITER plasma, providing evaluation of the fusion power for all operational phases. The pneumatic transfer system (PTS) is one of the key components of the NAS for the proper operation of the system, playing a role of transferring encapsulated samples between the capsule loading machine, irradiation stations, counting stations, and disposal bin. For the validation and the optimization of the design, a prototype of the PTS was developed and capsule transfer tests were performed with the developed system.

  18. Neutron-deuteron breakup reaction as a tool for studying neutron-neutron interactions

    SciTech Connect (OSTI)

    Konobeevski, E. S., E-mail: konobeev@inr.ru; Zuyev, S. V.; Mordovskoy, M. V.; Potashev, S. I.; Sharapov, I. M. [Russian Academy of Sciences, Institute for Nuclear Research (Russian Federation)] [Russian Academy of Sciences, Institute for Nuclear Research (Russian Federation)

    2013-11-15

    An analysis of the most recent data on the reaction nd {yields} pnn revealed a serious discrepancy between theoretical predictions and cross sections measured for this reaction in various configurations where the role of neutron-neutron interactions is important. In view of this, it seems necessary both to develop theoretical approaches and to obtain new experimental data. For this purpose, a setup for studying the neutron-deuteron breakup reaction was created at the Institute for Nuclear Research on the basis of the neutron beam in the RADEX channel and deuterium targets. This facility makes it possible to perform experiments over a broad region of primary-neutron energies (10-60 MeV) and in various (final-state interaction, quasifree scattering, and spatial-star) configurations. Preliminary results of the respective experiment were obtained for configurations of final-state neutron-neutron interaction and quasifree neutron-neutron scattering.

  19. Beta ray flux measuring device

    DOE Patents [OSTI]

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

    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.

  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. 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.

  2. Neutron Detection Using an Embedded Sol-Gel Neutron Absorber...

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

    Date Patent 5,973,328 Patent 5,973,328 Neutron detector using sol-gel absorber An neutron detector composed of fissionable material having ions of lithium, uranium, thorium, ...

  3. Category:Neutron Log | Open Energy Information

    Open Energy Info (EERE)

    Looking for the Neutron Log page? For detailed information on Neutron Log, click here. Category:Neutron Log Add.png Add a new Neutron Log Technique Pages in category...

  4. Fast neutron fluence of yonggwang nuclear unit 1 reactor pressure vessel

    SciTech Connect (OSTI)

    Yoo, C.; Km, B.; Chang, K.; Leeand, S.; Park, J.

    2006-07-01

    The Code of Federal Regulations, Title 10, Part 50, Appendix H, requires that the neutron dosimetry be present to monitor the reactor vessel throughout plant life. The Ex-Vessel Neutron Dosimetry System has been installed for Yonggwang Nuclear Unit 1 after complete withdrawal of all six in-vessel surveillance capsules. This system has been installed in the reactor cavity annulus in order to measure the fast neutron spectrum coming out through the reactor pressure vessel. Cycle specific neutron transport calculations were performed to obtain the energy dependent neutron flux throughout the reactor geometry including dosimetry positions. Comparisons between calculations and measurements were performed for the reaction rates of each dosimetry sensors and results show good agreements. (authors)

  5. Portable Neutron Sensors for Emergency Response Operations

    SciTech Connect (OSTI)

    Mukhopadhyay, S., Maurer, R., Detweiler, R.

    2012-06-22

    This slide-show presents neutron measurement work, including design, use and performance of different neutron detection systems.

  6. METHOD OF PRODUCING NEUTRONS

    DOE Patents [OSTI]

    Imhoff, D.H.; Harker, W.H.

    1964-02-01

    A method for producing neutrons is described in which there is employed a confinement zone defined between longitudinally spaced localized gradient regions of an elongated magnetic field. Changed particles and neutralizing electrons, more specifically deuterons and tritons and neutralizng electrons, are injected into the confinement field from ion sources located outside the field. The rotational energy of the parrticles is increased at the gradients by imposing an oscillating transverse electrical field thereacross. The imposition of such oscillating transverse electrical fields improves the reflection capability of such gradient fielda so that the reactive particles are retained more effectively within the zone. With the attainment of appropriate densities of plasma particles and provided that such particles are at a sufficiently high temperature, neutron-producing reactions ensue and large quantities of neutrons emerge from the containment zone. (AEC)

  7. Ultrafast neutron detector

    DOE Patents [OSTI]

    Wang, Ching L.

    1987-01-01

    The invention comprises a neutron detector (50) of very high temporal resolution that is particularly well suited for measuring the fusion reaction neutrons produced by laser-driven inertial confinement fusion targets. The detector comprises a biased two-conductor traveling-wave transmission line (54, 56, 58, 68) having a uranium cathode (60) and a phosphor anode (62) as respective parts of the two conductors. A charge line and Auston switch assembly (70, 72, 74) launch an electric field pulse along the transmission line. Neutrons striking the uranium cathode at a location where the field pulse is passing, are enabled to strike the phosphor anode and produce light that is recorded on photographic film (64). The transmission line may be variously configured to achieve specific experimental goals.

  8. Spherical neutron generator

    DOE Patents [OSTI]

    Leung, Ka-Ngo

    2006-11-21

    A spherical neutron generator is formed with a small spherical target and a spherical shell RF-driven plasma ion source surrounding the target. A deuterium (or deuterium and tritium) ion plasma is produced by RF excitation in the plasma ion source using an RF antenna. The plasma generation region is a spherical shell between an outer chamber and an inner extraction electrode. A spherical neutron generating target is at the center of the chamber and is biased negatively with respect to the extraction electrode which contains many holes. Ions passing through the holes in the extraction electrode are focused onto the target which produces neutrons by D-D or D-T reactions.

  9. Personnel electronic neutron dosimeter

    DOE Patents [OSTI]

    Falk, R.B.; Tyree, W.H.

    1982-03-03

    A personnel electronic dosimeter includes a neutron-proton and neutron-alpha converter for providing an electrical signal having a magnitude proportional to the energy of a detected proton or alpha particle produced from the converter, a pulse generator circuit for generating a pulse having a duration controlled by the weighed effect of the amplitude of the electrical signal, an oscillator enabled by the pulse for generating a train of clock pulses for a time dependent upon the pulse length, a counter for counting the clock pulses, and an indicator for providing a direct reading and aural alarm when the count indicates that the wearer has been exposed to a selected level of neutron dose equivalent.

  10. Personnel electronic neutron dosimeter

    DOE Patents [OSTI]

    Falk, Roger B.; Tyree, William H.

    1984-12-18

    A personnel electronic dosimeter includes a neutron-proton and neutron-alpha converter for providing an electrical signal having a magnitude proportional to the energy of a detected proton or alpha particle produced from the converter, a pulse generator circuit for generating a pulse having a duration controlled by the weighed effect of the amplitude of the electrical signal, an oscillator enabled by the pulse for generating a train of clock pulses for a time dependent upon the pulse length, a counter for counting the clock pulses, and an indicator for providing a direct reading and aural alarm when the count indicates that the wearer has been exposed to a selected level of neutron dose equivalent.

  11. 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.

  12. 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.

  13. Neutron Log | Open Energy Information

    Open Energy Info (EERE)

    Dictionary.png Neutron Log: The neutron log responds primarily to the amount of hydrogen in the formation which is contained in oil, natural gas, and water. The amount of...

  14. Corrosion resistant neutron absorbing coatings

    DOE Patents [OSTI]

    Choi, Jor-Shan; Farmer, Joseph C; Lee, Chuck K; Walker, Jeffrey; Russell, Paige; Kirkwood, Jon; Yang, Nancy; Champagne, Victor

    2013-11-12

    A method of forming a corrosion resistant neutron absorbing coating comprising the steps of spray or deposition or sputtering or welding processing to form a composite material made of a spray or deposition or sputtering or welding material, and a neutron absorbing material. Also a corrosion resistant neutron absorbing coating comprising a composite material made of a spray or deposition or sputtering or welding material, and a neutron absorbing material.

  15. Corrosion resistant neutron absorbing coatings

    DOE Patents [OSTI]

    Choi, Jor-Shan; Farmer, Joseph C.; Lee, Chuck K.; Walker, Jeffrey; Russell, Paige; Kirkwood, Jon; Yang, Nancy; Champagne, Victor

    2012-05-29

    A method of forming a corrosion resistant neutron absorbing coating comprising the steps of spray or deposition or sputtering or welding processing to form a composite material made of a spray or deposition or sputtering or welding material, and a neutron absorbing material. Also a corrosion resistant neutron absorbing coating comprising a composite material made of a spray or deposition or sputtering or welding material, and a neutron absorbing material.

  16. Introducing Nuclear Data Evaluations of Prompt Fission Neutron Spectra

    SciTech Connect (OSTI)

    Neudecker, Denise

    2015-06-17

    Nuclear data evaluations provide recommended data sets for nuclear data applications such as reactor physics, stockpile stewardship or nuclear medicine. The evaluated data are often based on information from multiple experimental data sets and nuclear theory using statistical methods. Therefore, they are collaborative efforts of evaluators, theoreticians, experimentalists, benchmark experts, statisticians and application area scientists. In this talk, an introductions is given to the field of nuclear data evaluation at the specific example of a recent evaluation of the outgoing neutron energy spectrum emitted promptly after fission from 239Pu and induced by neutrons from thermal to 30 MeV.

  17. Neutron Science Center

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

    Public to have rare opportunity to tour Neutron Science Center May 10, 2011 LANL Rosenfest will celebrate life of LANSCE founder Louis Rosen and offer tour LOS ALAMOS, New Mexico, May 10, 2011-Members of the public will have an unusual opportunity to tour the Los Alamos Neutron Science Center from 1 to 5 p.m. on Friday, May 20, 2011, as part of Rosenfest, a celebration of LANSCE founder Louis - 2 - Rosen and the remarkable facility he conceived. In addition, Rosenfest organizers are offering a

  18. Neutron activated switch

    DOE Patents [OSTI]

    Barton, David M.

    1991-01-01

    A switch for reacting quickly to a neutron emission. A rod consisting of fissionable material is located inside a vacuum tight body. An adjustable contact is located coaxially at an adjustable distance from one end of the rod. Electrical leads are connected to the rod and to the adjustable contact. With a vacuum drawn inside the body, a neutron bombardment striking the rod causes it to heat and expand longitudinally until it comes into contact with the adjustable contact. This circuit closing occurs within a period of a few microseconds.

  19. GUIDE FOR POLARIZED NEUTRONS

    DOE Patents [OSTI]

    Sailor, V.L.; Aichroth, R.W.

    1962-12-01

    The plane of polarization of a beam of polarized neutrons is changed by this invention, and the plane can be flipped back and forth quicitly in two directions in a trouble-free manner. The invention comprises a guide having a plurality of oppositely directed magnets forming a gap for the neutron beam and the gaps are spaced longitudinally in a spiral along the beam at small stepped angles. When it is desired to flip the plane of polarization the magnets are suitably rotated to change the direction of the spiral of the gaps. (AEC)

  20. FAST NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Snell, A.H.

    1957-12-01

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

  1. Gas core nuclear thermal rocket engine research and development in the former USSR

    SciTech Connect (OSTI)

    Koehlinger, M.W.; Bennett, R.G.; Motloch, C.G.; Gurfink, M.M.

    1992-09-01

    Beginning in 1957 and continuing into the mid 1970s, the USSR conducted an extensive investigation into the use of both solid and gas core nuclear thermal rocket engines for space missions. During this time the scientific and engineering. problems associated with the development of a solid core engine were resolved. At the same time research was undertaken on a gas core engine, and some of the basic engineering problems associated with the concept were investigated. At the conclusion of the program, the basic principles of the solid core concept were established. However, a prototype solid core engine was not built because no established mission required such an engine. For the gas core concept, some of the basic physical processes involved were studied both theoretically and experimentally. However, no simple method of conducting proof-of-principle tests in a neutron flux was devised. This report focuses primarily on the development of the. gas core concept in the former USSR. A variety of gas core engine system parameters and designs are presented, along with a summary discussion of the basic physical principles and limitations involved in their design. The parallel development of the solid core concept is briefly described to provide an overall perspective of the magnitude of the nuclear thermal propulsion program and a technical comparison with the gas core concept.

  2. LiF/ZnS Neutron Multiplicity Counter

    SciTech Connect (OSTI)

    Stave, Sean C.; Bliss, Mary; Kouzes, Richard T.; Lintereur, Azaree T.; Robinson, Sean M.; Siciliano, Edward R.; Wood, Lynn S.

    2015-06-01

    Abstract: Alternatives to the use of 3He for the detection of thermal neutrons are being investigated. One of the most challenging applications for 3He alternatives is in neutron multiplicity counters. Neutron multiplicity counters are used to provide rapid assay of samples which contain an unknown amount of plutonium in a potentially unknown configuration. With appropriate detector design, the neutron single, double, and triple coincidence events can be used to extract information of three unknown parameters such as the 240Pu-effective mass, the sample self-multiplication, and the (α,n) rate. A project at PNNL has investigated replacing 3He-based tubes with LiF/ZnS neutron-scintillator sheets and wavelength shifting plastic for light pipes. A four-panel demonstrator module has been constructed, tested, and compared with detailed modeling results. The findings indicate that a full-scale system can be constructed with the same overall size as the most efficient 3He-based system and with improved performance. Remaining design challenges include electronics and robust neutron/gamma-ray discrimination based on pulse shape analysis at high rates. A review of the current effort and the most recent findings will be presented.

  3. Characterization of a Thermo Scientific D711 D-T Neutron Generator Located in a Low-Scatter Facility

    SciTech Connect (OSTI)

    Hayes, John W.; Finn, Erin C.; Greenwood, Lawrence R.; Wittman, Richard S.

    2014-03-21

    A dosimetry experiment used to measure the neutron flux and spectrum of a D-T neutron generator is presented. The D-T generator at Pacific Northwest National Laboratory is installed in the middle of a large room to minimize scatter of neutrons back to the sample. The efficacy of maintaining a pure fast neutron field for the sample is investigated. Twenty-one positions within 13 cm of the neutron source contained foils or wires of Fe, Ni, Al with additional Au, and In monitors at some locations. Spectral adjustment of the neutron flux at each position based on measured reaction rates and theoretical Monte Carlo calculations show that at least 99.1% of the spectrum lies above 110 keV for all measured positions, and neutrons above 14 MeV can account for as much as 91% at locations along the axis of the generator and close to the source. The 14 MeV component drops to 77% in radial positions far from the source. The largest total flux observed was 8.29E+08 n/cm2-s (1.4%) in the center of the cooling cap, although additional experiments have shown this value could be as high as 1.20E+09 n/cm2-s.

  4. Quantitative method for measuring heat flux emitted from a cryogenic object

    DOE Patents [OSTI]

    Duncan, Robert V.

    1993-01-01

    The present invention is a quantitative method for measuring the total heat flux, and of deriving the total power dissipation, of a heat-fluxing object which includes the steps of placing an electrical noise-emitting heat-fluxing object in a liquid helium bath and measuring the superfluid transition temperature of the bath. The temperature of the liquid helium bath is thereafter reduced until some measurable parameter, such as the electrical noise, exhibited by the heat-fluxing object or a temperature-dependent resistive thin film in intimate contact with the heat-fluxing object, becomes greatly reduced. The temperature of the liquid helum bath is measured at this point. The difference between the superfluid transition temperature of the liquid helium bath surrounding the heat-fluxing object, and the temperature of the liquid helium bath when the electrical noise emitted by the heat-fluxing object becomes greatly reduced, is determined. The total heat flux from the heat-fluxing object is determined as a function of this difference between these temperatures. In certain applications, the technique can be used to optimize thermal design parameters of cryogenic electronics, for example, Josephson junction and infra-red sensing devices.

  5. Quantitative method for measuring heat flux emitted from a cryogenic object

    DOE Patents [OSTI]

    Duncan, R.V.

    1993-03-16

    The present invention is a quantitative method for measuring the total heat flux, and of deriving the total power dissipation, of a heat-fluxing object which includes the steps of placing an electrical noise-emitting heat-fluxing object in a liquid helium bath and measuring the superfluid transition temperature of the bath. The temperature of the liquid helium bath is thereafter reduced until some measurable parameter, such as the electrical noise, exhibited by the heat-fluxing object or a temperature-dependent resistive thin film in intimate contact with the heat-fluxing object, becomes greatly reduced. The temperature of the liquid helum bath is measured at this point. The difference between the superfluid transition temperature of the liquid helium bath surrounding the heat-fluxing object, and the temperature of the liquid helium bath when the electrical noise emitted by the heat-fluxing object becomes greatly reduced, is determined. The total heat flux from the heat-fluxing object is determined as a function of this difference between these temperatures. In certain applications, the technique can be used to optimize thermal design parameters of cryogenic electronics, for example, Josephson junction and infrared sensing devices.

  6. LAHET calculations for accelerator neutron production

    SciTech Connect (OSTI)

    Prael, R.E.

    1993-07-01

    LAHET is a Monte Carlo code for the transport and interaction of nucleons, pions, muons, fight ions, and antinucleons in complex geometry; it is the result of a major effort at Los Alamos National Laboratory to develop a code system based on the LANL version of the HETC Monte Carlo code for the transport of nucleons, pions, and muons, which was originally developed at Oak Ridge National Laboratory. The system of codes based on LAHET is designated as the LAHET Code System (LCS). LAHET, as all the variants of HETC, has been widely used over the years for design of neutron production targets, facility shielding, and experimental analysis. LAHET is now widely used for medical accelerator facility design and application. Particle tracking uses the general geometry model of the LANL MCNP code, and shares the geometry description and input of MCNP, except for lattices and/or repeated structures. HMCNP is a modification of MCNP which accepts an. external neutron/photon source created by LAHET. Neutron transport from 20 MeV to thermal and all photon/electron transport is done with HMCNP.

  7. High sensitivity, solid state neutron detector

    DOE Patents [OSTI]

    Stradins, Pauls; Branz, Howard M; Wang, Qi; McHugh, Harold R

    2015-05-12

    An apparatus (200) for detecting slow or thermal neutrons (160). The apparatus (200) includes an alpha particle-detecting layer (240) that is a hydrogenated amorphous silicon p-i-n diode structure. The apparatus includes a bottom metal contact (220) and a top metal contact (250) with the diode structure (240) positioned between the two contacts (220, 250) to facilitate detection of alpha particles (170). The apparatus (200) includes a neutron conversion layer (230) formed of a material containing boron-10 isotopes. The top contact (250) is pixilated with each contact pixel extending to or proximate to an edge of the apparatus to facilitate electrical contacting. The contact pixels have elongated bodies to allow them to extend across the apparatus surface (242) with each pixel having a small surface area to match capacitance based upon a current spike detecting circuit or amplifier connected to each pixel. The neutron conversion layer (860) may be deposited on the contact pixels (830) such as with use of inkjet printing of nanoparticle ink.

  8. High sensitivity, solid state neutron detector

    DOE Patents [OSTI]

    Stradins, Pauls; Branz, Howard M.; Wang, Qi; McHugh, Harold R.

    2013-10-29

    An apparatus (200) for detecting slow or thermal neutrons (160) including an alpha particle-detecting layer (240) that is a hydrogenated amorphous silicon p-i-n diode structure. The apparatus includes a bottom metal contact (220) and a top metal contact (250) with the diode structure (240) positioned between the two contacts (220, 250) to facilitate detection of alpha particles (170). The apparatus (200) includes a neutron conversion layer (230) formed of a material containing boron-10 isotopes. The top contact (250) is pixilated with each contact pixel extending to or proximate to an edge of the apparatus to facilitate electrical contacting. The contact pixels have elongated bodies to allow them to extend across the apparatus surface (242) with each pixel having a small surface area to match capacitance based upon a current spike detecting circuit or amplifier connected to each pixel. The neutron conversion layer (860) may be deposited on the contact pixels (830) such as with use of inkjet printing of nanoparticle ink.

  9. Neutron Absorbing Alloys

    DOE Patents [OSTI]

    Mizia, Ronald E.; Shaber, Eric L.; DuPont, John N.; Robino, Charles V.; Williams, David B.

    2004-05-04

    The present invention is drawn to new classes of advanced neutron absorbing structural materials for use in spent nuclear fuel applications requiring structural strength, weldability, and long term corrosion resistance. Particularly, an austenitic stainless steel alloy containing gadolinium and less than 5% of a ferrite content is disclosed. Additionally, a nickel-based alloy containing gadolinium and greater than 50% nickel is also disclosed.

  10. Neutronic Reactor Structure

    DOE Patents [OSTI]

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

    1961-05-30

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

  11. NEUTRONIC REACTOR STRUCTURE

    DOE Patents [OSTI]

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

    1961-05-30

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

  12. NEUTRONIC REACTOR CONTROL ELEMENT

    DOE Patents [OSTI]

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

    1962-04-17

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

  13. 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.

  14. Novel Large Area High Resolution Neutron Detector for the Spallation Neutron Source

    SciTech Connect (OSTI)

    Lacy, Jeffrey L

    2009-05-22

    Neutron scattering is a powerful technique that is critically important for materials science and structural biology applications. The knowledge gained from past developments has resulted in far-reaching advances in engineering, pharmaceutical and biotechnology industries, to name a few. New facilities for neutron generation at much higher flux, such as the SNS at Oak Ridge, TN, will greatly enhance the capabilities of neutron scattering, with benefits that extend to many fields and include, for example, development of improved drug therapies and materials that are stronger, longer-lasting, and more impact-resistant. In order to fully realize this enhanced potential, however, higher neutron rates must be met with improved detection capabilities, particularly higher count rate capability in large size detectors, while maintaining practicality. We have developed a neutron detector with the technical and economic advantages to accomplish this goal. This new detector has a large sensitive area, offers 3D spatial resolution, high sensitivity and high count rate capability, and it is economical and practical to produce. The proposed detector technology is based on B-10 thin film conversion of neutrons in long straw-like gas detectors. A stack of many such detectors, each 1 meter in length, and 4 mm in diameter, has a stopping power that exceeds that of He-3 gas, contained at practical pressures within an area detector. With simple electronic readout methods, straw detector arrays can provide spatial resolution of 4 mm FWHM or better, and since an array detector of such form consists of several thousand individual elements per square meter, count rates in a 1 m^2 detector can reach 2?10^7 cps. Moreover, each individual event can be timetagged with a time resolution of less than 0.1 ?sec, allowing accurate identification of neutron energy by time of flight. Considering basic elemental cost, this novel neutron imaging detector can be commercially produced economically, probably at a small fraction of the cost of He-3 detectors. In addition to neutron scattering science, the fully developed base technology can be used as a rugged, low-cost neutron detector in area monitoring and surveying. Radiation monitors are used in a number of other settings for occupational and environmental radiation safety. Such a detector can also be used in environmental monitoring and remote nuclear power plant monitoring. For example, the Department of Energy could use it to characterize nuclear waste dumps, coordinate clean-up efforts, and assess the radioactive contaminants in the air and water. Radiation monitors can be used to monitor the age and component breakdown of nuclear warheads and to distinguish between weapons and reactor grade plutonium. The UN's International Atomic Energy Agency (IAEA) uses radiation monitors for treaty verification, remote monitoring, and enforcing the non-proliferation of nuclear weapons. As part of treaty verification, monitors can be used to certify the contents of containers during inspections. They could be used for portal monitoring to secure border checkpoints, sea ports, air cargo centers, public parks, sporting venues, and key government buildings. Currently, only 2% of all sea cargo shipped is inspected for radiation sources. In addition, merely the presence of radiation is detected and nothing is known about the radioactive source until further testing. The utilization of radiation monitors with neutron sensitivity and capability of operation in hostile port environments would increase the capacity and effectiveness of the radioactive scanning processes.

  15. Dose-equivalent neutron dosimeter

    DOE Patents [OSTI]

    Griffith, R.V.; Hankins, D.E.; Tomasino, L.; Gomaa, M.A.M.

    1981-01-07

    A neutron dosimeter is disclosed which provides a single measurement indicating the amount of potential biological damage resulting from the neutron exposure of the wearer, for a wide range of neutron energies. The dosimeter includes a detecting sheet of track etch detecting material such as a carbonate plastic, for detecting higher energy neutrons, and a radiator layer contaning conversion material such as /sup 6/Li and /sup 10/B lying adjacent to the detecting sheet for converting moderate energy neutrons to alpha particles that produce tracks in the adjacent detecting sheet.

  16. Revised neutron dosimetry results for the MOTA-2A experiment in FFTF

    SciTech Connect (OSTI)

    Greenwood, L.R.

    1994-09-01

    Revised neutron fluence and damage values are reported for the MOTA-2A experiment in the Fast Flux Test Facility (FFTF). This revision corrects an error with processing of the {sup 235}U(n,f) reaction. Net corrections are on the order of 5%.

  17. Call issued for Lujan Neutron Scattering Center proposals

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

    (polarized and unpolarized) Inelastic neutron scattering spectroscopy Small angle neutron scattering Neutron radiographytomography The Lujan instruments webpage...

  18. Coated Fiber Neutron Detector Test

    SciTech Connect (OSTI)

    Lintereur, Azaree T.; Ely, James H.; Kouzes, Richard T.; Stromswold, David C.

    2009-10-23

    Radiation portal monitors used for interdiction of illicit materials at borders include highly sensitive neutron detection systems. The main reason for having neutron detection capability is to detect fission neutrons from plutonium. The currently deployed radiation portal monitors (RPMs) from Ludlum and Science Applications International Corporation (SAIC) use neutron detectors based upon 3He-filled gas proportional counters, which are the most common large neutron detector. There is a declining supply of 3He in the world, and thus, methods to reduce the use of this gas in RPMs with minimal changes to the current system designs and sensitivity to cargo-borne neutrons are being investigated. Reported here are the results of tests of the 6Li/ZnS(Ag)-coated non-scintillating plastic fibers option. This testing measured the required performance for neutron detection efficiency and gamma ray rejection capabilities of a system manufactured by Innovative American Technology (IAT).

  19. A more accurate and penetrating method to measure the enrichment and mass of UF6 storage containers using passive neutron self-interrogation

    SciTech Connect (OSTI)

    Menlove, Howard O; Swinhoe, Martyn T; Miller, Karen A

    2010-01-01

    This paper describes an unattended mode neutron measurement that can provide the enrichment of the uranium in UF{sub 6} cylinders. The new passive neutron measurement provides better penetration into the uranium mass than prior gamma-ray enrichment measurement methods. The Passive Neutron Enrichment Monitor (PNEM) provides a new measurement technique that uses passive neutron totals and coincidence counting together with neutron self-interrogation to measure the enrichment in the cylinders. The measurement uses the neutron rates from two detector pods. One of the pods has a bare polyethylene surface next to the cylinder and the other polyethylene surface is covered with Cd to prevent thermal neutrons from returning to the cylinder. The primary neutron source from the enriched UF{sub 6} is the alpha-particle decay from the {sub 234}U that interacts with the fluorine to produce random neutrons. The singles neutron counting rate is dominated by the {sub 234}U neutrons with a minor contribution from the induced fissions in the {sub 235}U. However, the doubles counting rate comes primarily from the induced fissions (i.e., multiplication) in the {sub 235}U in enriched uranium. The PNEM concept makes use of the passive neutrons that are initially produced from the {sub 234}U reactions that track the {sub 235}U enrichment during the enrichment process. The induced fission reactions from the thermal-neutron albedo are all from the {sub 235}U and provide a measurement of the {sub 235}U. The Cd ratio has the desirable feature that all of the thermal-neutron-induced fissions in {sub 235}U are independent of the original neutron source. Thus, the ratio is independent of the uranium age, purity, and prior reactor history.

  20. Study of heat flux gages using sensitivity analysis

    SciTech Connect (OSTI)

    Dowding, K.J.; Blackwell, B.F.; Cochran, R.J.

    1998-08-01

    The response and operation of a heat flux gage is studied using sensitivity analysis. Sensitivity analysis is the process by which one determines the sensitivity of a model output to changes in the model parameters. This process uses sensitivity coefficients that are defined as partial derivatives of field variables--e.g., temperature--with respect to model parameters--e.g., thermal properties and boundary conditions. Computing sensitivity coefficients, in addition to the response of a heat flux gage, aids in identifying model parameters that significantly impact the temperature response. A control volume, finite element-based code is used to implement numerical sensitivity coefficient calculations, allowing general problems to be studied. Sensitivity coefficients are discussed for the well known Gardon gage.

  1. ARM - PI Product - Radiative Flux Analysis

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

    love to hear from you Send us a note below or call us at 1-888-ARM-DATA. Send PI Product : Radiative Flux Analysis The Radiative Flux Analysis is a technique for using...

  2. 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.

  3. 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.

  4. 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.

  5. Thermal Analysis of Closed Systems

    Energy Science and Technology Software Center (OSTI)

    1987-10-01

    TAP-LOOP is a finite-difference program designed for steady-state and transient thermal analysis of recirculating fluid loops and associated heat transfer equipment; however, it is not limited to loop analysis. TAP-LOOP was developed to perform scoping and conceptual design analyses for closed test loops in the Fast Flux Test Facility (FFTF), but it can handle a variety of problems which can be described in terms of potentials, sources, sinks, and storage including, in addition to heatmore » transfer problems, studies of potential fluid flow, electrical networks, and stress analysis.« less

  6. FAST NEUTRON SPECTROMETER

    DOE Patents [OSTI]

    Davis, F.J.; Hurst, G.S.; Reinhardt, P.W.

    1959-08-18

    An improved proton recoil spectrometer for determining the energy spectrum of a fast neutron beam is described. Instead of discriminating against and thereby"throwing away" the many recoil protons other than those traveling parallel to the neutron beam axis as do conventional spectrometers, this device utilizes protons scattered over a very wide solid angle. An ovoidal gas-filled recoil chamber is coated on the inside with a scintillator. The ovoidal shape of the sensitive portion of the wall defining the chamber conforms to the envelope of the range of the proton recoils from the radiator disposed within the chamber. A photomultiplier monitors the output of the scintillator, and a counter counts the pulses caused by protons of energy just sufficient to reach the scintillator.

  7. Porous material neutron detector

    DOE Patents [OSTI]

    Diawara, Yacouba; Kocsis, Menyhert

    2012-04-10

    A neutron detector employs a porous material layer including pores between nanoparticles. The composition of the nanoparticles is selected to cause emission of electrons upon detection of a neutron. The nanoparticles have a maximum dimension that is in the range from 0.1 micron to 1 millimeter, and can be sintered with pores thereamongst. A passing radiation generates electrons at one or more nanoparticles, some of which are scattered into a pore and directed toward a direction opposite to the applied electrical field. These electrons travel through the pore and collide with additional nanoparticles, which generate more electrons. The electrons are amplified in a cascade reaction that occurs along the pores behind the initial detection point. An electron amplification device may be placed behind the porous material layer to further amplify the electrons exiting the porous material layer.

  8. DESPIKING OF SPACECRAFT ENERGETIC PROTON FLUX TO STUDY GALACTIC COSMIC-RAY MODULATION

    SciTech Connect (OSTI)

    Qin, G.; Zhao, L.-L.; Chen, H.-C. E-mail: llzhao@spaceweather.ac.cn

    2012-06-20

    Galactic cosmic ray (GCR) is usually assumed as a stable 'background', with solar influence considered as a modulation. The violent solar energetic particle (SEP) events associated with solar activities change particle fluxes by several orders of magnitude in a few minutes. Thus, the flux observation of GCR provided by satellites may be heavily contaminated by spurious spikes due to SEPs, and that provided by ground-based neutron monitors (NMs) may be contaminated by the system error spikes and the ground level enhancement effect. To obtain the 'pure' background GCR flux for modulation research, the removal of multifarious spikes is necessary. In this article, we use a robust automatic despiking algorithm based on the Poincare map thresholding method provided by Goring and Nikora for 'purification' of the time-series GCR flux observations. We can show that the algorithm is good at cleaning up the heavily contaminated GCR intensity rates measured by both spacecraft and NMs without artificial parameters. In addition, using the algorithm to despike the spacecraft observations of relatively lower energetic proton flux, we get both 11 year and 27 day period cycles comparable to the much higher energy GCR flux data measured by the ground-based NMs.

  9. COMBINE7.1 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

    SciTech Connect (OSTI)

    Woo Y. Yoon; David W. Nigg

    2009-08-01

    COMBINE7.1 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.1 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 fine-group cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko self-shielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those self-shielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.1 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a one-dimensional, discrete-ordinate transport code, is incoprated into COMBINE7.1. As an option, the 167 fine-group constants generated by COMBINE portion in the program can be used to cacluate regionwise spectra in the ANISN portion, all internally to reflect the one-dimensional transport correction. Results for the criticality validation calculations are included as a part of verification and validation.

  10. COMBINE7.0 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

    SciTech Connect (OSTI)

    Woo Y. Yoon; David W. Nigg

    2008-09-01

    COMBINE7.0 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.0 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 finegroup cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko selfshielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those selfshielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.0 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a onedimensional, discrete-ordinate transport code, is incoprated into COMBINE7.0. As an option, the 167 fine-group constants generated by COMBINE portion in the program can be used to cacluate regionwise spectra in the ANISN portion, all internally to reflect the one-dimensional transport correction. Results for the criticality validation calculations are included as a part of verification and validation.

  11. 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.

  12. Neutron Science and Technology

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

    3 Neutron Science and Technology From a mountaintop in Mexico where we investigate gamma rays, to underground laboratories where we study the behavior of plutonium under extreme conditions, our research spans the spectrum from fundamental to applied. Contact Us Group Leader Frank Merrill Email Deputy Group Leader David Oro Email Deputy Group Leader (acting) Anna Llobet Email Deputy Group Leader (acting) Keith Rielage Email Group Office Email Inner workings of a wristwatch that was imaged by a

  13. METAL ALLOY SUITABLE FOR CONTROLLING THERMAL NEUTRON REACTORS

    DOE Patents [OSTI]

    Schier, R.J.

    1963-11-26

    A homogeneous arc-melted alloy suitable for nuclear reactor control material consisting essentially of titanium and up to about 8 wt% gadolinium as gadolinium oxide is described. (AEC)

  14. Neutron diffraction study and anomalous negative thermal expansion...

    Office of Scientific and Technical Information (OSTI)

    does not result in the emergence of superconductivity or any other new ground state. ... Country of Publication: United States Language: English Subject: 72 PHYSICS OF ELEMENTARY ...

  15. Review of Non-Neutron and Neutron Nuclear Data, 2004

    SciTech Connect (OSTI)

    Holden, Norman E.

    2005-05-24

    Review articles are in preparation for the 2004 edition of the CRC Handbook of Chemistry and Physics dealing with the evaluation of both non-neutron and neutron nuclear data. Data on the discovery of element 110, Darmstadtium, and element 111 have been officially accepted, while data on element 118 have been withdrawn. Data to be presented include revised values for very short-lived nuclides, long-lived nuclides, and beta-beta decay measurements. There has been a reassessment of the spontaneous fission (sf) half-lives, which distinguishes between sf decay half-lives and cluster decay half-lives, and with cluster-fission decay. New measurements of isotopic abundance values for many elements will be discussed with an emphasis on the minor isotopes of interest for use in neutron activation analysis measurements. Neutron resonance integrals will be discussed emphasizing the differences between the calculated values obtained from the analytical integration over neutron resonances and the measured values in a neutron reactor-spectrum, which does not quite conform to the assumed 1/E neutron energy spectrum. The method used to determine the neutron resonance integral from measurement, using neutron activation analysis, will be discussed.

  16. REVIEW OF NON-NEUTRON AND NEUTRON NUCLEAR DATA, 2004.

    SciTech Connect (OSTI)

    HOLDEN, N.E.

    2004-09-26

    Review articles are in preparation for the 2004 edition of the CRC Handbook of Chemistry and Physics dealing with the evaluation of both non-neutron and neutron nuclear data. Data on the discovery of element 110, Darmstadtium, and element 111 have been officially accepted, while data on element 11 8 have been withdrawn. Data to be presented include revised values for very short-lived nuclides, long-lived nuclides and beta-beta decay measurements. There has been a reassessment of the spontaneous fission (sf) half-lives, which distinguishes between sf decay half-lives and cluster decay half-lives and with cluster-fission decay. New measurements of isotopic abundance values for many elements will be discussed with an emphasis on the minor isotopes of interest for use in neutron activation analysis measurements. Neutron resonance integrals will be discussed emphasizing the differences between the calculated values obtained from the analytical integration over neutron resonances and the measured values in a neutron reactor-spectrum, which does not quite conform to the assumed 1/E neutron energy spectrum. The method used to determine the neutron resonance integral from measurement, using neutron activation analysis, will be discussed.

  17. Fast neutron imaging device and method

    DOE Patents [OSTI]

    Popov, Vladimir; Degtiarenko, Pavel; Musatov, Igor V.

    2014-02-11

    A fast neutron imaging apparatus and method of constructing fast neutron radiography images, the apparatus including a neutron source and a detector that provides event-by-event acquisition of position and energy deposition, and optionally timing and pulse shape for each individual neutron event detected by the detector. The method for constructing fast neutron radiography images utilizes the apparatus of the invention.

  18. Overview of recent U235 neutron cross section evaluation work

    SciTech Connect (OSTI)

    Lubitz, C. [Lockheed Martin Corp., Schenectady, NY (United States)

    1998-10-01

    This report is an overview (through 1997) of the U235 neutron cross section evaluation work at Oak Ridge National Laboratory (ORNL), AEA Technology (Harwell) and Lockheed Martin Corp.-Schenectady (LMS), which has influenced, or appeared in, ENDF/B-VI through Release 5. The discussion is restricted to the thermal and resolved resonance regions, apart from some questions about the unresolved region which still need investigation. The important role which benchmark testing has played will be touched on.

  19. A radial collimator for a time-of-flight neutron spectrometer

    SciTech Connect (OSTI)

    Stone, M. B.; Abernathy, D. L. [Quantum Condensed Matter Science Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Niedziela, J. L.; Loguillo, M. J.; Overbay, M. A. [Instrument and Source Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States)

    2014-08-15

    We have engineered and installed a radial collimator for use in the scattered beam of a neutron time-of-flight spectrometer at a spallation neutron source. The radial collimator may be used with both thermal and epithermal neutrons, reducing the detected scattering intensity due to material outside of the sample region substantially. The collimator is located inside of the sample chamber of the instrument, which routinely cycles between atmospheric conditions and cryogenic vacuum. The oscillation and support mechanism of the collimator allow it to be removed from use without breaking vacuum. We describe here the design and characterization of this radial collimator.

  20. Category:Gas Flux Sampling | Open Energy Information

    Open Energy Info (EERE)

    Gas Flux Sampling Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Geothermalpower.jpg Looking for the Gas Flux Sampling page? For detailed information on Gas Flux...

  1. An example of neutronic penalizations in reactivity transient analysis using 3D coupled chain HEMERA

    SciTech Connect (OSTI)

    Dubois, F.; Normand, B.; Sargeni, A.

    2012-07-01

    HEMERA (Highly Evolutionary Methods for Extensive Reactor Analyses), is a fully coupled 3D computational chain developed jointly by IRSN and CEA. It is composed of CRONOS2 (core neutronics, cross sections library from APOLLO2), FLICA4 (core thermal-hydraulics) and the system code CATHARE. Multi-level and multi-dimensional models are developed to account for neutronics, core thermal-hydraulics, fuel thermal analysis and system thermal-hydraulics, dedicated to best-estimate, conservative simulations and sensitivity analysis. In IRSN, the HEMERA chain is widely used to study several types of reactivity accidents and for sensitivity studies. Just as an example of the HEMERA possibilities, we present here two types of neutronic penalizations and their impact on a power transient due to a REA (Rod Ejection Accident): in the first one, we studied a bum-up distribution modification and in the second one, a delayed-neutron fraction modification. Both modifications are applied to the whole core or localized in a few assemblies. Results show that it is possible to use global or local changes but 1) in case of bum-up modification, the total core power can increase when assembly peak power decrease so, care has to be taken if the goal is to maximize a local power peak and 2) for delayed-neutron fraction, a local modification can have the same effect as the one on the whole core, provided that it is large enough. (authors)

  2. NEUTRON IMAGING, RADIOGRAPHY AND TOMOGRAPHY.

    SciTech Connect (OSTI)

    SMITH,G.C.

    2002-03-01

    Neutrons are an invaluable probe in a wide range of scientific, medical and commercial endeavors. Many of these applications require the recording of an image of the neutron signal, either in one-dimension or in two-dimensions. We summarize the reactions of neutrons with the most important elements that are used for their detection. A description is then given of the major techniques used in neutron imaging, with emphasis on the detection media and position readout principle. Important characteristics such as position resolution, linearity, counting rate capability and sensitivity to gamma-background are discussed. Finally, the application of a subset of these instruments in radiology and tomography is described.

  3. Portable neutron spectrometer and dosimeter

    DOE Patents [OSTI]

    Waechter, D.A.; Erkkila, B.H.; Vasilik, D.G.

    The disclosure relates to a battery operated neutron spectrometer/dosimeter utilizing a microprocessor, a built-in tissue equivalent LET neutron detector, and a 128-channel pulse height analyzer with integral liquid crystal display. The apparatus calculates doses and dose rates from neutrons incident on the detector and displays a spectrum of rad or rem as a function of keV per micron of equivalent tissue and also calculates and displays accumulated dose in millirads and millirem as well as neutron dose rates in millirads per hour and millirem per hour.

  4. Advanced Neutron Source (ANS) Project

    SciTech Connect (OSTI)

    Campbell, J.H.; Selby, D.L.; Harrington, R.M.; Peretz, F.J.

    1991-02-01

    This report discusses the research and development, design and safety of the Advanced Neutron Source at Oak Ridge National Laboratory. (LSP)

  5. Portable neutron spectrometer and dosimeter

    DOE Patents [OSTI]

    Waechter, David A. (Los Alamos, NM); Erkkila, Bruce H. (Los Alamos, NM); Vasilik, Dennis G. (Los Alamos, NM)

    1985-01-01

    The disclosure relates to a battery operated neutron spectrometer/dosimeter utilizing a microprocessor, a built-in tissue equivalent LET neutron detector, and a 128-channel pulse height analyzer with integral liquid crystal display. The apparatus calculates doses and dose rates from neutrons incident on the detector and displays a spectrum of rad or rem as a function of keV per micron of equivalent tissue and also calculates and displays accumulated dose in millirads and millirem as well as neutron dose rates in millirads per hour and millirem per hour.

  6. MPACT Fast Neutron Multiplicity System Design Concepts

    SciTech Connect (OSTI)

    D. L. Chichester; S. A. Pozzi; J. L. Dolan; M. T. Kinlaw; A. C. Kaplan; M. Flaska; A. Enqvist; J. T. Johnsom; S. M. Watson

    2012-10-01

    This report documents work performed by Idaho National Laboratory and the University of Michigan in fiscal year (FY) 2012 to examine design parameters related to the use of fast-neutron multiplicity counting for assaying plutonium for materials protection, accountancy, and control purposes. This project seeks to develop a new type of neutron-measurement-based plutonium assay instrument suited for assaying advanced fuel cycle materials. Some current-concept advanced fuels contain high concentrations of plutonium; some of these concept fuels also contain other fissionable actinides besides plutonium. Because of these attributes the neutron emission rates of these new fuels may be much higher, and more difficult to interpret, than measurements made of plutonium-only materials. Fast neutron multiplicity analysis is one approach for assaying these advanced nuclear fuels. Studies have been performed to assess the conceptual performance capabilities of a fast-neutron multiplicity counter for assaying plutonium. Comparisons have been made to evaluate the potential improvements and benefits of fast-neutron multiplicity analyses versus traditional thermal-neutron counting systems. Fast-neutron instrumentation, using for example an array of liquid scintillators such as EJ-309, have the potential to either a) significantly reduce assay measurement times versus traditional approaches, for comparable measurement precision values, b) significantly improve assay precision values, for measurement durations comparable to current-generation technology, or c) moderating improve both measurement precision and measurement durations versus current-generation technology. Using the MCNPX-PoliMi Monte Carlo simulation code, studies have been performed to assess the doubles-detection efficiency for a variety of counter layouts of cylindrical liquid scintillator detector cells over one, two, and three rows. Ignoring other considerations, the best detector design is the one with the most detecting volume. However, operational limitations guide a) the maximum acceptable size of each detector cell (due to PSD performance and maximum-acceptable per-channel data throughput rates, limited by pulse pile-up and the processing rate of the electronics components of the system) and b) the affordability of a system due to the number of total channels of data to be collected and processed. As a first estimate, it appears that a system comprised of two rows of detectors 5" Ø ? 3" would yield a working prototype system with excellent performance capabilities for assaying Pu-containing items and capable of handling high signal rates likely when measuring items with Pu and other actinides. However, it is still likely that gamma-ray shielding will be needed to reduce the total signal rate in the detectors. As a first step prior to working with these larger-sized detectors, it may be practical to perform scoping studies using small detectors, such as already-on-hand 3" Ø ? 3" detectors.

  7. BARYON LOADING OF ACTIVE GALACTIC NUCLEUS JETS MEDIATED BY NEUTRONS

    SciTech Connect (OSTI)

    Toma, K.; Takahara, F.

    2012-08-01

    Plasmas of geometrically thick, black hole (BH) accretion flows in active galactic nuclei (AGNs) are generally collisionless for protons, and involve magnetic field turbulence. Under such conditions a fraction of protons can be accelerated stochastically and create relativistic neutrons via nuclear collisions. These neutrons can freely escape from the accretion flow and decay into protons in the dilute polar region above the rotating BH to form relativistic jets. We calculate geometric efficiencies of the neutron energy and mass injections into the polar region, and show that this process can deposit luminosity as high as L{sub j}{approx}2 Multiplication-Sign 10{sup -3} M-dot c{sup 2} and mass loading M-dot{sub j}{approx}6 Multiplication-Sign 10{sup -4} M-dot for the case of the BH mass M {approx} 10{sup 8} M{sub Sun }, where M-dot is the mass accretion rate. The terminal Lorentz factors of the jets are {Gamma} {approx} 3, and they may explain the AGN jets having low luminosities. For higher luminosity jets, which can be produced by additional energy inputs such as Poynting flux, the neutron decay still can be a dominant mass loading process, leading to, e.g., {Gamma} {approx} 50 for L{sub j,tot}{approx}3 Multiplication-Sign 10{sup -2} M-dot c{sup 2}.

  8. Sci—Fri PM: Dosimetry—02: A Nested Neutron Spectrometer to Measure Neutron Spectra in Radiotherapy

    SciTech Connect (OSTI)

    Maglieri, R; Seuntjens, J; Kildea, J; Licea, A

    2014-08-15

    During high-energy radiotherapy treatments, neutrons are produced in the head of the linac through photonuclear interactions. This has been a concern for many years as photoneutrons contribute to the accepted, yet unwanted, out-of-field doses that pose an iatrogenic risk to patients and an occupational risk to personnel. Presently, in-room neutron measurements are difficult and time-consuming and have traditionally been carried out using Bonner spheres with activation foils and TLDs. In this work, a new detector, the Nested Neutron Spectrometer (NNS) is tested for use in radiotherapy bunkers. The NNS is designed for easy handling and is more practical than the traditional Bonner spheres. The NNS, operated in current mode, was used to measure the dose equivalent, average energy and energy spectrum at several positions in a radiotherapy bunker. The average energy and spectra were compared to Monte Carlo simulations while the dose equivalent was compared to bubble detector measurements. The average energies, as measured by the NNS and Monte Carlo simulations, differed by approximately 30% across the bunker. Measurements of the dose equivalent using the NNS and the bubble detectors agreed within 50% in the maze and less than 10% close to the linac head. Apart from some discrepancies at thermal energies, we also found reasonable agreement between NNS-measured and Monte Carlo-simulated spectra at a number of locations within our radiotherapy bunker. Our results demonstrate that the NNS is a suitable detector to be used in high dose-rate radiotherapy environments.

  9. Secondary neutron spectrum from 250-MeV passively scattered proton therapy: Measurement with an extended-range Bonner sphere system

    SciTech Connect (OSTI)

    Howell, Rebecca M.; Burgett, E. A.

    2014-09-15

    Purpose: Secondary neutrons are an unavoidable consequence of proton therapy. While the neutron dose is low compared to the primary proton dose, its presence and contribution to the patient dose is nonetheless important. The most detailed information on neutrons includes an evaluation of the neutron spectrum. However, the vast majority of the literature that has reported secondary neutron spectra in proton therapy is based on computational methods rather than measurements. This is largely due to the inherent limitations in the majority of neutron detectors, which are either not suitable for spectral measurements or have limited response at energies greater than 20 MeV. Therefore, the primary objective of the present study was to measure a secondary neutron spectrum from a proton therapy beam using a spectrometer that is sensitive to neutron energies over the entire neutron energy spectrum. Methods: The authors measured the secondary neutron spectrum from a 250-MeV passively scattered proton beam in air at a distance of 100 cm laterally from isocenter using an extended-range Bonner sphere (ERBS) measurement system. Ambient dose equivalent H*(10) was calculated using measured fluence and fluence-to-ambient dose equivalent conversion coefficients. Results: The neutron fluence spectrum had a high-energy direct neutron peak, an evaporation peak, a thermal peak, and an intermediate energy continuum between the thermal and evaporation peaks. The H*(10) was dominated by the neutrons in the evaporation peak because of both their high abundance and the large quality conversion coefficients in that energy interval. The H*(10) 100 cm laterally from isocenter was 1.6 mSv per proton Gy (to isocenter). Approximately 35% of the dose equivalent was from neutrons with energies ≥20 MeV. Conclusions: The authors measured a neutron spectrum for external neutrons generated by a 250-MeV proton beam using an ERBS measurement system that was sensitive to neutrons over the entire energy range being measured, i.e., thermal to 250 MeV. The authors used the neutron fluence spectrum to demonstrate experimentally the contribution of neutrons with different energies to the total dose equivalent and in particular the contribution of high-energy neutrons (≥20 MeV). These are valuable reference data that can be directly compared with Monte Carlo and experimental data in the literature.

  10. NEUTRONIC REACTOR CONSTRUCTION

    DOE Patents [OSTI]

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

    1958-09-01

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

  11. LANSCE | Lujan Center | Instruments | Neutron Radiography

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

    Neutron Radiography The recently acquired energy-dispersive neutron imaging detector can be used on various Lujan Center beam-lines to combine scattering with imaging. In most...

  12. High energy neutron Computed Tomography developed

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

    objects. May 9, 2014 Neutron tomography horizontal "slice" of a tungsten and polyethylene test object containing tungsten carbide BBs. Neutron tomography horizontal "slice"...

  13. Efficiency measurements using tagged neutrons (Conference) |...

    Office of Scientific and Technical Information (OSTI)

    Efficiency measurements using tagged neutrons Citation Details In-Document Search Title: Efficiency measurements using tagged neutrons You are accessing a document from the ...

  14. SciTech Connect: "neutron scattering"

    Office of Scientific and Technical Information (OSTI)

    neutron scattering" Find + Advanced Search Term Search Semantic Search Advanced Search All Fields: "neutron scattering" Semantic Semantic Term Title: Full Text: Bibliographic...

  15. Comparison of MCNP6 and experimental results for neutron counts, Rossi-{alpha}, and Feynman-{alpha} distributions

    SciTech Connect (OSTI)

    Talamo, A.; Gohar, Y.; Sadovich, S.; Kiyavitskaya, H.; Bournos, V.; Fokov, Y.; Routkovskaya, C.

    2013-07-01

    MCNP6, the general-purpose Monte Carlo N-Particle code, has the capability to perform time-dependent calculations by tracking the time interval between successive events of the neutron random walk. In fixed-source calculations for a subcritical assembly, the zero time value is assigned at the moment the neutron is emitted by the external neutron source. The PTRAC and F8 cards of MCNP allow to tally the time when a neutron is captured by {sup 3}He(n, p) reactions in the neutron detector. From this information, it is possible to build three different time distributions: neutron counts, Rossi-{alpha}, and Feynman-{alpha}. The neutron counts time distribution represents the number of neutrons captured as a function of time. The Rossi-a distribution represents the number of neutron pairs captured as a function of the time interval between two capture events. The Feynman-a distribution represents the variance-to-mean ratio, minus one, of the neutron counts array as a function of a fixed time interval. The MCNP6 results for these three time distributions have been compared with the experimental data of the YALINA Thermal facility and have been found to be in quite good agreement. (authors)

  16. 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.

  17. Methods for Neutron Spectrometry

    DOE R&D Accomplishments [OSTI]

    Brockhouse, Bertram N.

    1961-01-09

    The appropriate theories and the general philosophy of methods of measurement and treatment of data neutron spectrometry are discussed. Methods of analysis of results for liquids using the Van Hove formulation, and for crystals using the Born-von Karman theory, are reviewed. The most useful of the available methods of measurement are considered to be the crystal spectrometer methods and the pulsed monoenergetic beam/time-of-flight method. Pulsed-beam spectrometers have the advantage of higher counting rates than crystal spectrometers, especially in view of the fact that simultaneous measurements in several counters at different angles of scattering are possible in pulsed-beam spectrometers. The crystal spectrometer permits several valuable new types of specialized experiments to be performed, especially energy distribution measurements at constant momentum transfer. The Chalk River triple-axis crystal-spectrometer is discussed, with reference to its use in making the specialized experiments. The Chalk River rotating crystal (pulsed-beam) spectrometer is described, and a comparison of this type instrument with other pulsed-beam spectrometers is made. A partial outline of the theory of operation of rotating-crystal spectrometers is presented. The use of quartz-crystal filters for fast neutron elimination and for order elimination is discussed. (auth)

  18. Fast neutron environments.

    SciTech Connect (OSTI)

    Buchheit, Thomas Edward; Kotula, Paul Gabriel; Lu, Ping; Brewer, Luke N.; Goods, Steven Howard; Foiles, Stephen Martin; Puskar, Joseph David; Hattar, Khalid Mikhiel; Doyle, Barney Lee; Boyce, Brad Lee; Clark, Blythe G.

    2011-10-01

    The goal of this LDRD project is to develop a rapid first-order experimental procedure for the testing of advanced cladding materials that may be considered for generation IV nuclear reactors. In order to investigate this, a technique was developed to expose the coupons of potential materials to high displacement damage at elevated temperatures to simulate the neutron environment expected in Generation IV reactors. This was completed through a high temperature high-energy heavy-ion implantation. The mechanical properties of the ion irradiated region were tested by either micropillar compression or nanoindentation to determine the local properties, as a function of the implantation dose and exposure temperature. In order to directly compare the microstructural evolution and property degradation from the accelerated testing and classical neutron testing, 316L, 409, and 420 stainless steels were tested. In addition, two sets of diffusion couples from 316L and HT9 stainless steels with various refractory metals. This study has shown that if the ion irradiation size scale is taken into consideration when developing and analyzing the mechanical property data, significant insight into the structural properties of the potential cladding materials can be gained in about a week.

  19. High Dose Neutron Irradiation of Hi-Nicalon Type S Silicon Carbide Composites, Part 2. Mechanical and Physical Properties

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

    Katoh, Yutai; Nozawa, Takashi; Shih, Chunghao Phillip; Ozawa, Kazumi; Koyanagi, Takaaki; Porter, Wallace D; Snead, Lance Lewis

    2015-01-07

    Nuclear-grade silicon carbide (SiC) composite material was examined for mechanical and thermophysical properties following high-dose neutron irradiation in the High Flux Isotope Reactor at a temperature range of 573–1073 K. Likewise, the material was chemical vapor-infiltrated SiC-matrix composite with a two-dimensional satin weave Hi-Nicalon Type S SiC fiber reinforcement and a multilayered pyrocarbon/SiC interphase. Moderate (1073 K) to very severe (573 K) degradation in mechanical properties was found after irradiation to >70 dpa, whereas no evidence was found for progressive evolution in swelling and thermal conductivity. The swelling was found to recover upon annealing beyond the irradiation temperature, indicating themore » irradiation temperature, but only to a limited extent. Moreover, the observed strength degradation is attributed primarily to fiber damage for all irradiation temperatures, particularly a combination of severe fiber degradation and likely interphase damage at relatively low irradiation temperatures.« less

  20. First Results of Scanning Thermal Diffusivity Microscope (STDM) Measurements on Irradiated Monolithic and Dispersion Fuel

    SciTech Connect (OSTI)

    T. K. Huber; M. K. Figg; J. R. Kennedy; A. B. Robinson; D. M. Wachs

    2012-07-01

    The thermal conductivity of the fuel material in a reactor before and during irradiation is a sensitive and fundamental parameter for thermal hydraulic calculations that are useds to correctly determine fuel heat fluxes and meat temperatures and to simulate performance of the fuel elements during operation. Several techniques have been developed to measure the thermal properties of fresh fuel to support these calculations, but it is crucial to also investigate the change of thermal properties during irradiation.

  1. Nanocrystalline Separation Membrane for Improved Hydrogen Flux...

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

    for Improved Hydrogen Flux New processing technique to develop ionic transport membranes with improved ionic and electronic conductivity Savannah River National Laboratory...

  2. Fluxing agent for metal cast joining

    DOE Patents [OSTI]

    Gunkel, Ronald W.; Podey, Larry L.; Meyer, Thomas N.

    2002-11-05

    A method of joining an aluminum cast member to an aluminum component. The method includes the steps of coating a surface of an aluminum component with flux comprising cesium fluoride, placing the flux coated component in a mold, filling the mold with molten aluminum alloy, and allowing the molten aluminum alloy to solidify thereby joining a cast member to the aluminum component. The flux preferably includes aluminum fluoride and alumina. A particularly preferred flux includes about 60 wt. % CsF, about 30 wt. % AlF.sub.3, and about 10 wt. % Al.sub.2 O.sub.3.

  3. MAGNETAR GIANT FLARES AND THEIR PRECURSORS-FLUX ROPE ERUPTIONS WITH CURRENT SHEETS

    SciTech Connect (OSTI)

    Yu Cong

    2013-07-10

    We propose a catastrophic magnetospheric model for magnetar precursors and their successive giant flares. Axisymmetric models of the magnetosphere, which contain both a helically twisted flux rope and a current sheet, are established based on force-free field configurations. In this model, the helically twisted flux rope would lose its equilibrium and erupt abruptly in response to the slow and quasi-static variations at the ultra-strongly magnetized neutron star's surface. In a previous model without current sheets, only one critical point exists in the flux rope equilibrium curve. New features show up in the equilibrium curves for the flux rope when current sheets appear in the magnetosphere. The causal connection between the precursor and the giant flare, as well as the temporary re-entry of the quiescent state between the precursor and the giant flare, can be naturally explained. Magnetic energy would be released during the catastrophic state transitions. The detailed energetics of the model are also discussed. The current sheet created by the catastrophic loss of equilibrium of the flux rope provides an ideal place for magnetic reconnection. We point out the importance of magnetic reconnection for further enhancement of the energy release during eruptions.

  4. Thermal breeder fuel enrichment zoning

    DOE Patents [OSTI]

    Capossela, Harry J.; Dwyer, Joseph R.; Luce, Robert G.; McCoy, Daniel F.; Merriman, Floyd C.

    1992-01-01

    A method and apparatus for improving the performance of a thermal breeder reactor having regions of higher than average moderator concentration are disclosed. The fuel modules of the reactor core contain at least two different types of fuel elements, a high enrichment fuel element and a low enrichment fuel element. The two types of fuel elements are arranged in the fuel module with the low enrichment fuel elements located between the high moderator regions and the high enrichment fuel elements. Preferably, shim rods made of a fertile material are provided in selective regions for controlling the reactivity of the reactor by movement of the shim rods into and out of the reactor core. The moderation of neutrons adjacent the high enrichment fuel elements is preferably minimized as by reducing the spacing of the high enrichment fuel elements and/or using a moderator having a reduced moderating effect.

  5. Comparison of the high temperature heat flux sensor to traditional heat flux gages under high heat flux conditions.

    SciTech Connect (OSTI)

    Blanchat, Thomas K.; Hanks, Charles R.

    2013-04-01

    Four types of heat flux gages (Gardon, Schmidt-Boelter, Directional Flame Temperature, and High Temperature Heat Flux Sensor) were assessed and compared under flux conditions ranging between 100-1000 kW/m2, such as those seen in hydrocarbon fire or propellant fire conditions. Short duration step and pulse boundary conditions were imposed using a six-panel cylindrical array of high-temperature tungsten lamps. Overall, agreement between all gages was acceptable for the pulse tests and also for the step tests. However, repeated tests with the HTHFS with relatively long durations at temperatures approaching 1000%C2%B0C showed a substantial decrease (10-25%) in heat flux subsequent to the initial test, likely due to the mounting technique. New HTHFS gages have been ordered to allow additional tests to determine the cause of the flux reduction.

  6. Effects of thermal fluctuations on thermal inflation

    SciTech Connect (OSTI)

    Hiramatsu, Takashi; Miyamoto, Yuhei; Yokoyama, Jun’ichi

    2015-03-12

    The mechanism of thermal inflation, a relatively short period of accelerated expansion after primordial inflation, is a desirable ingredient for a certain class of particle physics models if they are not to be in contention with the cosmology of the early Universe. Though thermal inflation is most simply described in terms of a thermal effective potential, a thermal environment also gives rise to thermal fluctuations that must be taken into account. We numerically study the effects of these thermal fluctuations using lattice simulations. We conclude that though they do not ruin the thermal inflation scenario, the phase transition at the end of thermal inflation proceeds through phase mixing and is therefore not accompanied by the formations of bubbles nor appreciable amplitude of gravitational waves.

  7. High Temperature Thermal Array for Next Generation Solar Thermal...

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

    High Temperature Thermal Array for Next Generation Solar Thermal Power Production High Temperature Thermal Array for Next Generation Solar Thermal Power Production This ...

  8. Thermal-Hydrology Simulations of Disposal of High-Level Radioactive Waste in a Single Deep Borehole

    SciTech Connect (OSTI)

    Hadgu, Teklu; Stein, Emily; Hardin, Ernest; Freeze, Geoffrey A.; Hammond, Glenn Edward

    2015-11-01

    Simulations of thermal-hydrology were carried out for the emplacement of spent nuclear fuel canisters and cesium and strontium capsules using the PFLOTRAN simulator. For the cesium and strontium capsules the analysis looked at disposal options such as different disposal configurations and surface aging of waste to reduce thermal effects. The simulations studied temperature and fluid flux in the vicinity of the borehole. Simulation results include temperature and vertical flux profiles around the borehole at selected depths. Of particular importance are peak temperature increases, and fluxes at the top of the disposal zone. Simulations of cesium and strontium capsule disposal predict that surface aging and/or emplacement of the waste at the top of the disposal zone reduces thermal effects and vertical fluid fluxes. Smaller waste canisters emplaced over a longer disposal zone create the smallest thermal effect and vertical fluid fluxes no matter the age of the waste or depth of emplacement.

  9. AmeriFlux US-Atq Atqasuk

    SciTech Connect (OSTI)

    Oechel, Walt; Zona, Donatella

    2016-01-01

    This is the AmeriFlux version of the carbon flux data for the site US-Atq Atqasuk. Site Description - This site is 100 km south of Barrow, Alaska, Variety of moist-wet coastal sedge tundra, and moist-tussock tundra surfaces in the more well-drained upland.

  10. AmeriFlux US-Brw Barrow

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

    Oechel, Walt [San Diego State University; Zona, Donatella [San Diego State University

    2016-01-01

    This is the AmeriFlux version of the carbon flux data for the site US-Brw Barrow. Site Description - The local landscape surrounding the Barrow site has a history absent of any disturbances. The terrain was not heavily glaciated during the last period of glaciation. The vegetation is mature in an unmanaged and undisturbed Arctic tundra.

  11. AmeriFlux US-Ivo Ivotuk

    SciTech Connect (OSTI)

    Oechel, Walter; Zona, Donatella

    2016-01-01

    This is the AmeriFlux version of the carbon flux data for the site US-Ivo Ivotuk. Site Description - This site is 300 km south of Barrow and is located at the foothill of the Brooks Range and is classified as tussock sedge, dwarf-shrub, moss tundra.

  12. ARM - VAP Product - lblch1flux

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

    Productslbllblch1flux Documentation Data Management Facility Plots (Quick Looks) Citation DOI: 10.5439/1095322 [ What is this? ] Generate Citation 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 VAP Output : LBLCH1FLUX AERI: LBLRTM, channel 1

  13. ARM - VAP Product - lblch2flux

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

    Productslbllblch2flux Documentation Data Management Facility Plots (Quick Looks) Citation DOI: 10.5439/1095329 [ What is this? ] Generate Citation 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 VAP Output : LBLCH2FLUX AERI, line by line channel 2

  14. Neutron Irradiation Resistance of RAFM Steels

    SciTech Connect (OSTI)

    Gaganidze, Ermile; Dafferner, Bernhard; Aktaa, Jarir

    2008-07-01

    The neutron irradiation resistance of the reduced-activation ferritic/martensitic (RAFM) steel EUROFER97 and international reference steels (F82H-mod, OPTIFER-Ia, GA3X and MANET-I) have been investigated after irradiation in the Petten High Flux Reactor up to 16.3 dpa at different irradiation temperatures (250-450 deg. C). The embrittlement behavior and hardening are investigated by instrumented Charpy-V tests with sub-size specimens. Neutron irradiation-induced embrittlement and hardening of EUROFER97 was studied under different heat treatment conditions. Embrittlement and hardening of as-delivered EUROFER97 steel are comparable to those of reference steels. Heat treatment of EUROFER97 at a higher austenitizing temperature substantially improves the embrittlement behaviour at low irradiation temperatures. Analysis of embrittlement vs. hardening behavior of RAFM steels within a proper model in terms of the parameter C={delta}DBTT/{delta}{sigma} indicates hardening-dominated embrittlement at irradiation temperatures below 350 deg. C with 0.17 {<=} C {<=} 0.53 deg. C/MPa. Scattering of C at irradiation temperatures above 400 deg. C indicates non hardening embrittlement. A role of He in a process of embrittlement is investigated in EUROFER97 based steels, that are doped with different contents of natural B and the separated {sup 10}B-isotope (0.008-0.112 wt.%). Testing on small scale fracture mechanical specimens for determination of quasi-static fracture toughness will be also presented in a view of future irradiation campaigns. (authors)

  15. REVIEW OF FAST FLUX TEST FACILITY (FFTF) FUEL EXPERIMENTS FOR STORAGE IN INTERIM STORAGE CASKS (ISC)

    SciTech Connect (OSTI)

    CHASTAIN, S.A.

    2005-10-24

    Appendix H, Section H.3.3.10.11 of the Final Safety Analysis Report (FSAR), provides the limits to be observed for fueled components authorized for storage in the Fast Flux Test Facility (FFTF) spent fuel storage system. Currently, the authorization basis allows standard driver fuel assemblies (DFA), as described in the FSAR Chapter 17, Section 17.5.3.1, to be stored provided decay power per assembly is {le} 250 watts, post-irradiation time is four years minimum, average assembly burn-up is 150,000 MWD/MTHM maximum and the pre-irradiation enrichment is 29.3% maximum (per H.3.3.10.11). In addition, driver evaluation (DE), core characterizer assemblies (CCA), and run-to-cladding-breach (RTCB) assemblies are included based on their similarities to a standard DFA. Ident-69 pin containers with fuel pins from these DFAs can also be stored. Section H.3.3.10.11 states that fuel types outside the specification criteria above will be addressed on a case-by-case basis. There are many different types of fuel and blanket experiments that were irradiated in the FFTF which now require offload to the spent fuel storage system. Two reviews were completed for a portion of these special type fuel components to determine if placement into the Core Component Container (CCC)/Interim Storage Cask (ISC) would require any special considerations or changes to the authorization basis. Project mission priorities coupled with availability of resources and analysts prevented these evaluations from being completed as a single effort. Areas of review have included radiological accident release consequences, radiological shielding adequacy, criticality safety, thermal limits, confinement, and stress. The results of these reviews are available in WHC-SD-FF-RPT-005, Rev. 0 and 1, ''Review of FFTF Fuel Experiments for Storage at ISA'', (Reference I), which subsequently allowed a large portion of these components to be included in the authorization basis (Table H.3.3-21). The report also identified additional components and actions in Section 3.0 and Table 3 that require further evaluation. The purpose of this report is to evaluate another portion of the remaining inventory (i.e., delayed neutron signal fuel, blanket assemblies, highly enriched assemblies, newly loaded Ident-69 pin containers, and returned fuel) to ensure it can be safely off loaded to the FFTF spent fuel storage system.

  16. Thermal Control & System Integration

    Broader source: Energy.gov [DOE]

    The thermal control and system integration activity focuses on issues such as the integration of motor and power control technologies and the development of advanced thermal control technologies....

  17. Temperature and field dependence of the flux-line-lattice symmetry in V{sub 3}Si

    SciTech Connect (OSTI)

    Yethiraj, M.; Christen, D.K.; Gapud, A.A.; Paul, D. McK.; Crowe, S.J.; Dewhurst, C.D.; Cubitt, R.; Porcar, L.; Gurevich, A.

    2005-08-01

    In V{sub 3}Si, a first-order structural phase transition from hexagonal to square flux-line lattice occurs at approximately 1 T with H parallel to the a axis. In this paper, we demonstrate the reentrant structural transition in the flux-line lattice, which reverts to hexagonal symmetry as the magnetic field approached H{sub c2}(T). This behavior is described very well by a nonlocal London theory with thermal fluctuations. The phase diagram of the flux lattice topology is mapped out for this geometry.

  18. 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 facilities such as the High Flux Isotope Reactor at ORNL has been initiated by the US DOE and NASA for space exploration needs. Two Monte Carlo-based depletion codes, TRITON (ORNL) and VESTA (IRSN), were used to study the {sup 238}Pu production rates with varying target configurations in a typical HFIR fuel cycle. Preliminary studies have shown that approximately 11 grams and within 15 to 17 grams of {sup 238}Pu could be produced in the first irradiation cycle in one small and one large VXF facility, respectively, when irradiating fresh target arrays as those herein described. Important to note is that in this study we discovered that small differences in assumptions could affect the production rates of Pu-238 observed. The exact flux at a specific target location can have a significant impact upon production, so any differences in how the control elements are modeled as a function of exposure, will also cause differences in production rates. In fact, the surface plot of the large VXF target Pu-238 production shown in Figure 3 illustrates that the pins closest to the core can potentially have production rates as high as 3 times those of pins away from the core, thus implying that a cycle-to-cycle rotation of the targets may be well advised. A methodology for generating spatially-dependent, multi-group self-shielded cross sections and flux files with the KENO and CENTRM codes has been created so that standalone ORIGEN-S inputs can be quickly constructed to perform a variety of {sup 238}Pu production scenarios, i.e. combinations of the number of arrays loaded and the number of irradiation cycles. The studies herein shown with VESTA and TRITON/KENO will be used to benchmark the standalone ORIGEN.

  19. An Analysis Technique for Active Neutron Multiplicity Measurements Based on First Principles

    SciTech Connect (OSTI)

    Evans, Louise G; Goddard, Braden; Charlton, William S; Peerani, Paolo

    2012-08-13

    Passive neutron multiplicity counting is commonly used to quantify the total mass of plutonium in a sample, without prior knowledge of the sample geometry. However, passive neutron counting is less applicable to uranium measurements due to the low spontaneous fission rates of uranium. Active neutron multiplicity measurements are therefore used to determine the {sup 235}U mass in a sample. Unfortunately, there are still additional challenges to overcome for uranium measurements, such as the coupling of the active source and the uranium sample. Techniques, such as the coupling method, have been developed to help reduce the dependence of calibration curves for active measurements on uranium samples; although, they still require similar geometry known standards. An advanced active neutron multiplicity measurement method is being developed by Texas A&M University, in collaboration with Los Alamos National Laboratory (LANL) in an attempt to overcome the calibration curve requirements. This method can be used to quantify the {sup 235}U mass in a sample containing uranium without using calibration curves. Furthermore, this method is based on existing detectors and nondestructive assay (NDA) systems, such as the LANL Epithermal Neutron Multiplicity Counter (ENMC). This method uses an inexpensive boron carbide liner to shield the uranium sample from thermal and epithermal neutrons while allowing fast neutrons to reach the sample. Due to the relatively low and constant fission and absorption energy dependent cross-sections at high neutron energies for uranium isotopes, fast neutrons can penetrate the sample without significant attenuation. Fast neutron interrogation therefore creates a homogeneous fission rate in the sample, allowing for first principle methods to be used to determine the {sup 235}U mass in the sample. This paper discusses the measurement method concept and development, including measurements and simulations performed to date, as well as the potential limitations.

  20. Thermal Hydraulic Characteristics of Fuel Defects in Plate Type Nuclear

    Office of Scientific and Technical Information (OSTI)

    Research Reactors (Technical Report) | SciTech Connect Technical Report: Thermal Hydraulic Characteristics of Fuel Defects in Plate Type Nuclear Research Reactors Citation Details In-Document Search Title: Thermal Hydraulic Characteristics of Fuel Defects in Plate Type Nuclear Research Reactors Turbulent flow coupled with heat transfer is investigated for a High Flux Isotope Reactor (HFIR) fuel plate. The Reynolds Averaged Navier-Stokes Models are used for fluid dynamics and the transfer of

  1. Thermal island destabilization and the Greenwald limit (Journal Article) |

    Office of Scientific and Technical Information (OSTI)

    SciTech Connect Thermal island destabilization and the Greenwald limit Citation Details In-Document Search Title: Thermal island destabilization and the Greenwald limit Magnetic reconnection is ubiquitous in the magnetosphere, the solar corona, and in toroidal fusion research discharges. In a fusion device, a magnetic island saturates at a width which produces a minimum in the magnetic energy of the configuration. At saturation, the modified current density profile, a function of the flux in

  2. Laser Intertial Fusion Energy: Neutronic Design Aspects of a Hybrid Fusion-Fission Nuclear Energy System

    SciTech Connect (OSTI)

    Kramer, K

    2010-04-08

    This study investigates the neutronics design aspects of a hybrid fusion-fission energy system called the Laser Fusion-Fission Hybrid (LFFH). A LFFH combines current Laser Inertial Confinement fusion technology with that of advanced fission reactor technology to produce a system that eliminates many of the negative aspects of pure fusion or pure fission systems. When examining the LFFH energy mission, a significant portion of the United States and world energy production could be supplied by LFFH plants. The LFFH engine described utilizes a central fusion chamber surrounded by multiple layers of multiplying and moderating media. These layers, or blankets, include coolant plenums, a beryllium (Be) multiplier layer, a fertile fission blanket and a graphite-pebble reflector. Each layer is separated by perforated oxide dispersion strengthened (ODS) ferritic steel walls. The central fusion chamber is surrounded by an ODS ferritic steel first wall. The first wall is coated with 250-500 {micro}m of tungsten to mitigate x-ray damage. The first wall is cooled by Li{sub 17}Pb{sub 83} eutectic, chosen for its neutron multiplication and good heat transfer properties. The {sub 17}Pb{sub 83} flows in a jacket around the first wall to an extraction plenum. The main coolant injection plenum is immediately behind the Li{sub 17}Pb{sub 83}, separated from the Li{sub 17}Pb{sub 83} by a solid ODS wall. This main system coolant is the molten salt flibe (2LiF-BeF{sub 2}), chosen for beneficial neutronics and heat transfer properties. The use of flibe enables both fusion fuel production (tritium) and neutron moderation and multiplication for the fission blanket. A Be pebble (1 cm diameter) multiplier layer surrounds the coolant injection plenum and the coolant flows radially through perforated walls across the bed. Outside the Be layer, a fission fuel layer comprised of depleted uranium contained in Tristructural-isotropic (TRISO) fuel particles having a packing fraction of 20% in 2 cm diameter fuel pebbles. The fission blanket is cooled by the same radial flibe flow that travels through perforated ODS walls to the reflector blanket. This reflector blanket is 75 cm thick comprised of 2 cm diameter graphite pebbles cooled by flibe. The flibe extraction plenum surrounds the reflector bed. Detailed neutronics designs studies are performed to arrive at the described design. The LFFH engine thermal power is controlled using a technique of adjusting the {sup 6}Li/{sup 7}Li enrichment in the primary and secondary coolants. The enrichment adjusts system thermal power in the design by increasing tritium production while reducing fission. To perform the simulations and design of the LFFH engine, a new software program named LFFH Nuclear Control (LNC) was developed in C++ to extend the functionality of existing neutron transport and depletion software programs. Neutron transport calculations are performed with MCNP5. Depletion calculations are performed using Monteburns 2.0, which utilizes ORIGEN 2.0 and MCNP5 to perform a burnup calculation. LNC supports many design parameters and is capable of performing a full 3D system simulation from initial startup to full burnup. It is able to iteratively search for coolant {sup 6}Li enrichments and resulting material compositions that meet user defined performance criteria. LNC is utilized throughout this study for time dependent simulation of the LFFH engine. Two additional methods were developed to improve the computation efficiency of LNC calculations. These methods, termed adaptive time stepping and adaptive mesh refinement were incorporated into a separate stand alone C++ library name the Adaptive Burnup Library (ABL). The ABL allows for other client codes to call and utilize its functionality. Adaptive time stepping is useful for automatically maximizing the size of the depletion time step while maintaining a desired level of accuracy. Adaptive meshing allows for analysis of fixed fuel configurations that would normally require a computationally burdensome number of depletion zones. Alternatively, Adaptive Mesh Refinement (AMR) adjusts the depletion zone size according to the variation in flux across the zone or fractional contribution to total absorption or fission. A parametric analysis on a fully mixed fuel core was performed using the LNC and ABL code suites. The resulting system parameters are found to optimize performance metrics using a 20 MT DU fuel load with a 20% TRISO packing and a 300 {micro}m kernel radius operated with a fusion input power of 500 MW and a fission blanket gain of 4.0. LFFH potentially offers a proliferation resistant technology relative to other nuclear energy systems primarily because of no need for fuel enrichment or reprocessing. A figure of merit of the material attractiveness is examined and it is found that the fuel is effectively contaminated to an unattractive level shortly after the system is started due to fission product and minor actinide build up.

  3. On the possibility of neutron generation in an imploding TiD{sub 2} puff Z pinch

    SciTech Connect (OSTI)

    Baksht, Rina B.; Tel-Aviv University, Tel Aviv ; Oreshkin, Vladimir I.; Tomsk Polytechnic University, Tomsk ; Rousskikh, Alexander G.

    2013-08-15

    Simulation of implosion of a TiD{sub 2} puff Z pinch is reported. The Z pinch is supposed to be produced by the plasma flow generated by a vacuum arc, as described by Rousskikh et al.[Phys. Plasmas 18, 092707 (2011)]. To simulate the implosion, a one-dimensional two-temperature radiative magnetohydrodynamics code was used. The simulation has shown that neutrons are generated during the implosion of a TiD{sub 2} puff Z pinch due to thermalization of the pinch plasma stagnated on axis. It has been shown that the necessary condition for neutron generation is that the ion temperature must be substantially higher than the electron temperature. For a pinch current of 1 MA, the predicted yield of 'thermal' neutrons is 2.5 10{sup 9} neutrons/shot.

  4. Observation of d-d fusion neutrons during degassing of deuterium-loaded palladium

    SciTech Connect (OSTI)

    Bittner, M.; Meister, A.; Seeliger, D.; Schwierz, R.; Wuestner, P. )

    1993-05-01

    Experiments with two massive deuterium-loaded palladium samples designed to search for deuteron-deuteron (d-d) fusion during thermal degassing are described. In the heavier of the two samples, which has a total mass of [approximately] 0.5 kg, during deuterium expulsion from the metal, a significant neutron excess count rate was detected by two independent NE-213 scintillation neutron detectors. The maximum time-dependent excess count rate corresponds to a d-d reaction rate of (3 [+-] 1) [times] 10[sup [minus]25] per deuteron pair per second. From detector pulse height spectra, the energy of the neutrons is determined to be [approximately] 2.5 MeV, as expected for d-d fusion neutrons. 10 refs., 10 figs., 2 tabs.

  5. The nuclear physics of neutron stars

    SciTech Connect (OSTI)

    Piekarewicz, J.

    2014-05-09

    We explore the unique and fascinating structure of neutron stars. Although neutron stars are of interest in many areas of Physics, our aim is to provide an intellectual bridge between Nuclear Physics and Astrophysics. We argue against the naive perception of a neutron star as a uniform assembly of neutrons packed to enormous densities. Rather, by focusing on the many exotic phases that are speculated to exist in a neutron star, we show how the reality is different and far more interesting.

  6. High energy neutron Computed Tomography developed

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

    High energy neutron Computed Tomography developed High energy neutron Computed Tomography developed LANSCE now has a high-energy neutron imaging capability that can be deployed on WNR flight paths for unclassified and classified objects. May 9, 2014 Neutron tomography horizontal "slice" of a tungsten and polyethylene test object containing tungsten carbide BBs. Neutron tomography horizontal "slice" of a tungsten and polyethylene test object containing tungsten carbide BBs.

  7. Instability due to a two recirculation pump trip in a BWR using RAMONA-4B computer code with 3D neutron kinetics

    SciTech Connect (OSTI)

    Cheng, H.S.; Rohatgi, U.S.

    1993-06-01

    An investigation was made of the potential for thermal-hydraulic instabilities coupled to neutronic feedback in a BWR due to a two recirculation pump trip event using the RAMONA-4B computer code with 3D neutron kinetics. It is concluded that a high-power (100%) and low-flow (75%) initial condition would most likely lead to in-phase density wave oscillations after the tripping of both recirculation pumps, and that RAMONA-4B is capable of predicting such thermal-hydraulic instabilities coupled to neutronic feedback in BWR and in SBWR.

  8. The Spallation Neutron Source Project

    Broader source: Energy.gov [DOE]

    When the Department of Energy (DOE) set out in the 1990s to develop a neutron scattering research facility that was ten times more powerful than the state of the art, the concept for the project...

  9. Ion chamber based neutron detectors

    DOE Patents [OSTI]

    Derzon, Mark S; Galambos, Paul C; Renzi, Ronald F

    2014-12-16

    A neutron detector with monolithically integrated readout circuitry, including: a bonded semiconductor die; an ion chamber formed in the bonded semiconductor die; a first electrode and a second electrode formed in the ion chamber; a neutron absorbing material filling the ion chamber; and the readout circuitry which is electrically coupled to the first and second electrodes. The bonded semiconductor die includes an etched semiconductor substrate bonded to an active semiconductor substrate. The readout circuitry is formed in a portion of the active semiconductor substrate. The ion chamber has a substantially planar first surface on which the first electrode is formed and a substantially planar second surface, parallel to the first surface, on which the second electrode is formed. The distance between the first electrode and the second electrode may be equal to or less than the 50% attenuation length for neutrons in the neutron absorbing material filling the ion chamber.

  10. Development of a Time-tagged Neutron Source for SNM Detection

    SciTech Connect (OSTI)

    Ji, Qing; Ludewigt, Bernhard; Wallig, Joe; Waldron, Will; Tinsley, Jim

    2015-06-18

    Associated particle imaging (API) is a powerful technique for special nuclear material (SNM) detection and characterization of fissile material configurations. A sealed-tube neutron generator has been under development by Lawrence Berkeley National Laboratory to reduce the beam spot size on the neutron production target to 1 mm in diameter for a several-fold increase in directional resolution and simultaneously increases the maximum attainable neutron flux. A permanent magnet 2.45 GHz microwave-driven ion source has been adopted in this time-tagged neutron source. This type of ion source provides a high plasma density that allows the use of a sub-millimeter aperture for the extraction of a sufficient ion beam current and lets us achieve a much reduced beam spot size on target without employing active focusing. The design of this API generator uses a custom-made radial high voltage insulator to minimize source to neutron production target distance and to provide for a simple ion source cooling arrangement. Preliminary experimental results showed that more than 100 µA of deuterium ions have been extracted, and the beam diameter on the neutron production target is around 1 mm.

  11. Development of a Time-tagged Neutron Source for SNM Detection

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

    Ji, Qing; Ludewigt, Bernhard; Wallig, Joe; Waldron, Will; Tinsley, Jim

    2015-06-18

    Associated particle imaging (API) is a powerful technique for special nuclear material (SNM) detection and characterization of fissile material configurations. A sealed-tube neutron generator has been under development by Lawrence Berkeley National Laboratory to reduce the beam spot size on the neutron production target to 1 mm in diameter for a several-fold increase in directional resolution and simultaneously increases the maximum attainable neutron flux. A permanent magnet 2.45 GHz microwave-driven ion source has been adopted in this time-tagged neutron source. This type of ion source provides a high plasma density that allows the use of a sub-millimeter aperture for themore » extraction of a sufficient ion beam current and lets us achieve a much reduced beam spot size on target without employing active focusing. The design of this API generator uses a custom-made radial high voltage insulator to minimize source to neutron production target distance and to provide for a simple ion source cooling arrangement. Preliminary experimental results showed that more than 100 µA of deuterium ions have been extracted, and the beam diameter on the neutron production target is around 1 mm.« less

  12. 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.

  13. 239Pu Resonance Evaluation for Thermal Benchmark System Calculations

    SciTech Connect (OSTI)

    Leal, Luiz C; Noguere, G; De Saint Jean, C; Kahler, A.

    2013-01-01

    Analyses of thermal plutonium solution critical benchmark systems have indicated a deciency in the 239Pu resonance evaluation. To investigate possible solutions to this issue, the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) Working Party for Evaluation Cooperation (WPEC) established Subgroup 34 to focus on the reevaluation of the 239Pu resolved resonance parameters. In addition, the impacts of the prompt neutron multiplication (nubar) and the prompt neutron ssion spectrum (PFNS) have been investigated. The objective of this paper is to present the results of the 239Pu resolved resonance evaluation eort.

  14. CONTROL MEANS FOR NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Tonks, L.

    1962-08-01

    A control device surrounding the active portion of a nuclear reactor is described. The control device consists of a plurality of contiguous cylinders partly filled with a neutron absorbing material and partly filled with a neutron reflecting material, each cylinder having a longitudinal reentrant surface into which a portion of an adjacent cylinder extends, one of the cylinders having two re-entrant surfaces, and means for rotating the cylinders one at a time. (AEC)

  15. Alternative Neutron Detection Testing Summary

    SciTech Connect (OSTI)

    Kouzes, Richard T.; Ely, James H.; Erikson, Luke E.; Kernan, Warnick J.; Lintereur, Azaree T.; Siciliano, Edward R.; Stromswold, David C.; Woodring, Mitchell L.

    2010-04-08

    Radiation portal monitors used for interdiction of illicit materials at borders include highly sensitive neutron detection systems. The main reason for having neutron detection capability is to detect fission neutrons from plutonium. Most currently deployed radiation portal monitors (RPMs) use neutron detectors based upon 3He-filled gas proportional counters, which are the most common large area neutron detector. This type of neutron detector is used in the TSA and other RPMs installed in international locations and in the Ludlum and Science Applications International Corporation RPMs deployed primarily for domestic applications. There is a declining supply of 3He in the world and, thus, methods to reduce the use of this gas in RPMs with minimal changes to the current system designs and sensitivity to cargo-borne neutrons are being investigated. Four technologies have been identified as being currently commercially available, potential alternative neutron detectors to replace the use of 3He in RPMs. These technologies are: 1) Boron trifluoride-filled proportional counters, 2) Boron-lined proportional counters, 3) Lithium-loaded glass fibers, and 4) Coated wavelength-shifting plastic fibers. Reported here is a summary of the testing carried out at Pacific Northwest National Laboratory on these technologies to date, as well as measurements on 3He tubes at various pressures. Details on these measurements are available in the referenced reports. Sponsors of these tests include the Department of Energy (DOE), Department of Homeland Security (DHS), and the Department of Defense (DoD), as well as internal Pacific Northwest National Laboratory funds.

  16. SETTABLE NEUTRON RADIATION SHIELDING MATERIAL

    DOE Patents [OSTI]

    Axelrad, I.R.

    1960-11-22

    A settable, viscous, putty-like shielding composition is described. It consists of an intimate admixture of a major proportion of a compound having a ratio of hydrogen atoms to all other atoms therein within the range of from 0.5: 1 to 2:l. from 0.5 to 10% by weight of boron, and a fluid resinous carrier This composition when cured is adapted to attenuate fast moving neutrons and capture slow moving neutrons.

  17. Neutron detectors comprising boron powder

    DOE Patents [OSTI]

    Wang, Zhehui; Morris, Christopher; Bacon, Jeffrey Darnell; Makela, Mark F; Spaulding, Randy Jay

    2013-05-21

    High-efficiency neutron detector substrate assemblies comprising a first conductive substrate, wherein a first side of the substrate is in direct contact with a first layer of a powder material comprising .sup.10boron, .sup.10boron carbide or combinations thereof, and wherein a conductive material is in proximity to the first layer of powder material; and processes of making said neutron detector substrate assemblies.

  18. 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.

  19. Development of a new flux map processing code for moveable detector system in PWR

    SciTech Connect (OSTI)

    Li, W.; Lu, H.; Li, J.; Dang, Z.; Zhang, X.

    2013-07-01

    This paper presents an introduction to the development of the flux map processing code MAPLE developed by China Nuclear Power Technology Research Institute (CNPPJ), China Guangdong Nuclear Power Group (CGN). The method to get the three-dimensional 'measured' power distribution according to measurement signal has also been described. Three methods, namely, Weight Coefficient Method (WCM), Polynomial Expand Method (PEM) and Thin Plane Spline (TPS) method, have been applied to fit the deviation between measured and predicted results for two-dimensional radial plane. The measured flux map data of the LINGAO nuclear power plant (NPP) is processed using MAPLE as a test case to compare the effectiveness of the three methods, combined with a 3D neutronics code COCO. Assembly power distribution results show that MAPLE results are reasonable and satisfied. More verification and validation of the MAPLE code will be carried out in future. (authors)

  20. NEUTRON RADIOGRAPHY (NRAD) REACTOR 64-ELEMENT CORE UPGRADE

    SciTech Connect (OSTI)

    John D. Bess

    2014-03-01

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