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1

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

Science Conference Proceedings (OSTI)

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

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

2011-12-15T23:59:59.000Z

2

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

DOE Patents (OSTI)

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.

Bowman, Charles D. (Los Alamos, NM)

1992-01-01T23:59:59.000Z

3

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

DOE Patents (OSTI)

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.

Bowman, C.D.

1992-11-03T23:59:59.000Z

4

NEUTRON FLUX INTENSITY DETECTION  

DOE Patents (OSTI)

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)

Russell, J.T.

1964-04-21T23:59:59.000Z

5

Thermal neutron detection system  

DOE Patents (OSTI)

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.

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

2000-01-01T23:59:59.000Z

6

THERMAL NEUTRONIC REACTOR  

DOE Patents (OSTI)

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

Spinrad, B.I.

1960-01-12T23:59:59.000Z

7

TEMPEST II--A NEUTRON THERMALIZATION CODE  

SciTech Connect

The TEMPEST II neutron thermalization code in Fortran for IBM 709 or 7090 calculates thermal neutron flux spectra based upon the Wigner-Wilkins equation, the Wilkins equation, or the Maxwellian distribution. When a neutron spectrum is obtained, TEMPEST II provides microscopic and macroscopic cross section averages over that spectrum. Equations used by the code and sample input and output data are given. (auth)

Shudde, R.H.; Dyer, J.

1962-06-01T23:59:59.000Z

8

THERMAL NEUTRON BACKSCATTER IMAGING.  

DOE Green Energy (OSTI)

Objects of various shapes, with some appreciable hydrogen content, were exposed to fast neutrons from a pulsed D-T generator, resulting in a partially-moderated spectrum of backscattered neutrons. The thermal component of the backscatter was used to form images of the objects by means of a coded aperture thermal neutron imaging system. Timing signals from the neutron generator were used to gate the detection system so as to record only events consistent with thermal neutrons traveling the distance between the target and the detector. It was shown that this time-of-flight method provided a significant improvement in image contrast compared to counting all events detected by the position-sensitive {sup 3}He proportional chamber used in the imager. The technique may have application in the detection and shape-determination of land mines, particularly non-metallic types.

VANIER,P.; FORMAN,L.; HUNTER,S.; HARRIS,E.; SMITH,G.

2004-10-16T23:59:59.000Z

9

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

DOE Patents (OSTI)

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.

Crane, Thomas W. (Los Alamos, NM)

1986-01-01T23:59:59.000Z

10

Neutronics Modeling of the High Flux Isotope Reactor using COMSOL  

Science Conference Proceedings (OSTI)

The High Flux Isotope Reactor located at the Oak Ridge National Laboratory is a versatile 85 MWth research reactor with cold and thermal neutron scattering, materials irradiation, isotope production, and neutron activation analysis capabilities. HFIR staff members are currently in the process of updating the thermal hydraulic and reactor transient modeling methodologies. COMSOL Multiphysics has been adopted for the thermal hydraulic analyses and has proven to be a powerful finite-element-based simulation tool for solving multiple physics-based systems of partial and ordinary differential equations. Modeling reactor transients is a challenging task because of the coupling of neutronics, heat transfer, and hydrodynamics. This paper presents a preliminary COMSOL-based neutronics study performed by creating a two-dimensional, two-group, diffusion neutronics model of HFIR to study the spatially-dependent, beginning-of-cycle fast and thermal neutron fluxes. The 238-group ENDF/B-VII neutron cross section library and NEWT, a two-dimensional, discrete-ordinates neutron transport code within the SCALE 6 code package, were used to calculate the two-group neutron cross sections required to solve the diffusion equations. The two-group diffusion equations were implemented in the COMSOL coefficient form PDE application mode and were solved via eigenvalue analysis using a direct (PARDISO) linear system solver. A COMSOL-provided adaptive mesh refinement algorithm was used to increase the number of elements in areas of largest numerical error to increase the accuracy of the solution. The flux distributions calculated by means of COMSOL/SCALE compare well with those calculated with benchmarked three-dimensional MCNP and KENO models, a necessary first step along the path to implementing two- and three-dimensional models of HFIR in COMSOL for the purpose of studying the spatial dependence of transient-induced behavior in the reactor core.

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

2011-01-01T23:59:59.000Z

11

Electrospun Polymer Nanofiber Composite as Thermal Neutron ...  

Science Conference Proceedings (OSTI)

Lithium-6 isotope has a significant thermal neutron cross-section and produces high energy charged particles on thermal neutron absorption. In this research ...

12

NEUTRONIC REACTOR HAVING LOCALIZED AREAS OF HIGH THERMAL NEUTRON DENSITIES  

DOE Patents (OSTI)

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.

Newson, H.W.

1958-06-01T23:59:59.000Z

13

Modulating the Neutron Flux from a Mirror Neutron Source  

Science Conference Proceedings (OSTI)

A 14-MeV neutron source based on a Gas-Dynamic Trap will provide a high flux of 14 MeV neutrons for fusion materials and sub-component testing. In addition to its main goal, the source has potential applications in condensed matter physics and biophysics. In this report, the author considers adding one more capability to the GDT-based neutron source, the modulation of the neutron flux with a desired frequency. The modulation may be an enabling tool for the assessment of the role of non-steady-state effects in fusion devices as well as for high-precision, low-signal basic science experiments favoring the use of the synchronous detection technique. A conclusion is drawn that modulation frequency of up to 1 kHz and modulation amplitude of a few percent is achievable. Limitations on the amplitude of modulations at higher frequencies are discussed.

Ryutov, D D

2011-09-01T23:59:59.000Z

14

Neutronic reactor thermal shield  

DOE Patents (OSTI)

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

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

1976-06-15T23:59:59.000Z

15

A Novel Detector for High Neutron Flux Measurements  

Science Conference Proceedings (OSTI)

Measuring alpha particles from a neutron induced break-up reaction with a mass spectrometer can be an excellent tool for detecting neutrons in a high neutron flux environment. Break-up reactions of {sup 6}Li and {sup 12}C can be used in the detection of slow and fast neutrons, respectively. A high neutron flux detection system that integrates the neutron energy sensitive material and helium mass spectrometer has been developed. The description of the detector configuration is given and it is soon to be tested at iThemba LABS, South Africa.

Singo, T. D.; Wyngaardt, S. M. [Department of Physics, University of Stellenbosch, Private bag X1, Matieland, Stellenbosch (South Africa); Papka, P. [Department of Physics, University of Stellenbosch, Private bag X1, Matieland, Stellenbosch (South Africa); Nuclear Physics group, iThemba labs, P. O. Box 722, Somerset West 7129 (South Africa); Dobson, R. T. [Department of Mechanical and Mechatronic Engineering, University of Stellenbosch, Private bag X1, Matieland, Stellenbosch (South Africa)

2010-01-05T23:59:59.000Z

16

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

Science Conference Proceedings (OSTI)

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)

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)

2012-07-01T23:59:59.000Z

17

Measurements of the Thermal Neutron Scattering Kernel  

E-Print Network (OSTI)

world's most powerful neutron source, the $1.4 billion Spallation Neutron Source At 1.4MW, SNS produces. SNS will feature 24 beamlines for physics, chemistry, biology, materials research. www.sns.gov #12 · Coproduction of epithermal, thermal and cold neutrons #12;SNS Instrument Beam Lines 1st experimentproposed 2nd

Danon, Yaron

18

Thermal neutron shield and method of manufacture  

DOE Patents (OSTI)

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.

Brindza, Paul Daniel; Metzger, Bert Clayton

2013-05-28T23:59:59.000Z

19

Neutron flux profile monitor for use in a fission reactor  

DOE Patents (OSTI)

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.

Kopp, Manfred K. (Oak Ridge, TN); Valentine, Kenneth H. (Lenoir City, TN)

1983-01-01T23:59:59.000Z

20

A neutronic feasibility study for LEU conversion of the high flux beam reactor (HFBR).  

SciTech Connect

A neutronic feasibility study for converting the High Flux Beam Reactor at Brookhaven National Laboratory from HEU to LEU fuel was performed at Argonne National Laboratory. The purpose of this study is to determine what LEU fuel density would be needed to provide fuel lifetime and neutron flux performance similar to the current HEU fuel. The results indicate that it is not possible to convert the HFBR to LEU fuel with the current reactor core configuration. To use LEU fuel, either the core needs to be reconfigured to increase the neutron thermalization or a new LEU reactor design needs to be considered. This paper presents results of reactor calculations for a reference 28-assembly HEU-fuel core configuration and for an alternative 18-assembly LEU-fuel core configuration with increased neutron thermalization. Neutronic studies show that similar in-core and ex-core neutron fluxes, and fuel cycle length can be achieved using high-density LEU fuel with about 6.1 gU/cm{sup 3} in an altered reactor core configuration. However, hydraulic and safety analyses of the altered HFBR core configuration needs to be performed in order to establish the feasibility of this concept.

Pond, R. B.

1998-01-16T23:59:59.000Z

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


21

Determination of thermal neutron capture gamma yields.  

E-Print Network (OSTI)

A method of analysing Ge(Li) thermal neutron capture gamma spectra to obtain total gamma yields has been developed. Tie method determines both the yields from the well resolved gamma peaks in a spectrum as well as the gamma ...

Harper, Thomas Lawrence

1969-01-01T23:59:59.000Z

22

Type II superconductivity and magnetic flux transport in neutrons stars  

E-Print Network (OSTI)

The transition to a type II proton superconductor which is believed to occur in a cooling neutron star is accompanied by changes in the equation of hydrostatic equilibrium and by the formation of proton vortices with quantized magnetic flux. Analysis of the electron Boltzmann equation for this system and of the proton supercurrent distribution formed at the transition leads to the derivation of a simple expression for the transport velocity of magnetic flux in the liquid interior of a neutron star. This shows that flux moves easily as a consequence of the interaction between neutron and proton superfluid vortices during intervals of spin-down or spin-up in binary systems. The differences between the present analysis and those of previous workers are reviewed and an error in the paper of Jones (1991) is corrected.

P. B. Jones

2005-10-13T23:59:59.000Z

23

Compound Refractive Lenses for Thermal Neutron Applications  

SciTech Connect

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.

Gary, Charles K.

2013-11-12T23:59:59.000Z

24

Thermal testing of solid neutron shielding materials  

Science Conference Proceedings (OSTI)

Two legal-weight truck casks the GA-4 and GA-9, will carry four PWR and nine BWR spent fuel assemblies, respectively. Each cask has a solid neutron shielding material separating the steel body and the outer steel skin. In the thermal accident specified by NRC regulations in 10CFR Part 71, the cask is subjected to an 800[degree]C environment for 30 minutes. The neutron shield need not perform any shielding function during or after the thermal accident, but its behavior must not compromise the ability of the cask to contain the radioactive contents. In May-June 1989 the first series of full-scale thermal tests was performed on three shielding materials: Bisco Products NS-4-FR, and Reactor Experiments RX-201 and RX-207. The tests are described in Thermal Testing of Solid Neutron Shielding Materials, GA-AL 9897, R. H. Boonstra, General Atomics (1990), and demonstrated the acceptability of these materials in a thermal accident. Subsequent design changes to the cask rendered these materials unattractive in terms of weight or adequate service temperature margin. For the second test series, a material specification was developed for a polypropylene based neutron shield with a softening point of at least 280[degree]F. The neutron shield materials tested were boronated (0.8--4.5%) polymers (polypropylene, HDPE, NS-4). The Envirotech and Bisco materials are not polypropylene, but were tested as potential backup materials in the event that a satisfactory polypropylene could not be found.

Boonstra, R.H.

1992-09-01T23:59:59.000Z

25

CREEP RUPTURE IN THE PRESENCE OF A FAST NEUTRON FLUX  

SciTech Connect

Possible mechanisms for creep rupture during irradiation are examined. Evidence that the rupture occurs by grain boundary sliding alone, or by vacancy condensation, is compared. It is observed that vacancy condensation is the more probable mechanism, and that this mechanism predicts a reduction in creep rupture life for metals exposed to a fast neutron flux (neglecting effects of radiation annealing). (T.F.H.)

Gregory, D.P.

1959-01-14T23:59:59.000Z

26

High precision thermal neutron detectors  

Science Conference Proceedings (OSTI)

Two-dimensional position sensitive detectors are indispensable in neutron diffraction experiments for determination of molecular and crystal structures in biology, solid-state physics and polymer chemistry. Some performance characteristics of these detectors are elementary and obvious, such as the position resolution, number of resolution elements, neutron detection efficiency, counting rate and sensitivity to gamma-ray background. High performance detectors are distinguished by more subtle characteristics such as the stability of the response (efficiency) versus position, stability of the recorded neutron positions, dynamic range, blooming or halo effects. While relatively few of them are needed around the world, these high performance devices are sophisticated and fairly complex; their development requires very specialized efforts. In this context, we describe here a program of detector development, based on {sup 3}He filled proportional chambers, which has been underway for some years at Brookhaven. Fundamental approaches and practical considerations are outlined that have resulted in a series of high performance detectors with the best known position resolution, position stability, uniformity of reliability over time of this type.

Radeka, V.; Schaknowski, N.A.; Smith, G.C.; and Yu, B.

1994-10-01T23:59:59.000Z

27

THERMAL PERFORMANCE OF A FAST NEUTRON TEST CONCEPT FOR THE ADVANCED TEST REACTOR  

Science Conference Proceedings (OSTI)

Since 1967, the Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL) has provided state-of-the-art experimental irradiation testing capability. A unique design is investigated herein for the purpose of providing a fast neutron flux test capability in the ATR. This new test capability could be brought on line in approximately 5 or 6 years, much sooner than a new test reactor could be built, to provide an interim fast-flux test capability in the timeframe before a fast-flux research reactor could be built. The proposed cost for this system is approximately $63M, much less than the cost of a new fast-flux test reactor. A concept has been developed to filter out a large portion of the thermal flux component by using a thermally conductive neutron absorber block. The objective of this study is to determine the feasibility of this experiment cooling concept.

Donna Post Guillen

2008-06-01T23:59:59.000Z

28

High Flux Isotope Reactor cold neutron source reference design concept  

SciTech Connect

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

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

1998-05-01T23:59:59.000Z

29

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

SciTech Connect

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

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

2013-04-19T23:59:59.000Z

30

HFIR Technical Parameters | ORNL Neutron Sciences  

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

a thorough understanding of how elements react to neutron bombardment. Facilities at HFIR Two Pneumatic Tubes: PT-1: Thermal Neutron Flux: 4 1014 n cm-2 s-1...

31

Methods for absorbing neutrons  

DOE Patents (OSTI)

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.

Guillen, Donna P. (Idaho Falls, ID); Longhurst, Glen R. (Idaho Falls, ID); Porter, Douglas L. (Idaho Falls, ID); Parry, James R. (Idaho Falls, ID)

2012-07-24T23:59:59.000Z

32

Transient heat flux shielding using thermal metamaterials  

E-Print Network (OSTI)

We have developed a heat shield based on a metamaterial engineering approach to shield a region from transient diffusive heat flow. The shield is designed with a multilayered structure to prescribe the appropriate spatial profile for heat capacity, density, and thermal conductivity of the effective medium. The heat shield was experimentally compared to other isotropic materials.

Narayana, Supradeep; Sato, Yuki

2013-01-01T23:59:59.000Z

33

Study of the Neutron Flux and Dpa Attenuation in the Reactor Pressure-Vessel Wall  

Science Conference Proceedings (OSTI)

The study of the neutron flux and dpa attenuation in the reactor pressure vessel (PV) wall presented in this work was performed with state-of-the art methods currently used to determine PV fluxes, the BUGLE-96 cross-section library, and the iron displacement cross sections derived from ENDF/B-VI data. The calculations showed that the RG 1.99, Rev. 2, extrapolation formula predicts slower--and therefore conservative--attenuation of the neutron flux (E > 1MeV) in the PV wall. More importantly, the calculations gave slower attenuation of the dpa rate in the PV wall than the attenuation predicted by the formula. The slower dpa rate attenuation was observed for all the cases considered, which included two different PWRs, and several configurations obtained by varying the PV wall thickness and thermal shield thickness. For example, for a PV wall thickness of {approximately}24 cm, the calculated ratio of the dpa rate at 1/4 and 3/4 of the PV wall thickness to the dpa value on the inner PV surface is {approximately}14% and 19% higher, respectively, than predicted by the RG 1.99, Rev. 2, formula.

Remec, I.

1999-06-01T23:59:59.000Z

34

Comparison of HEU and LEU Fuel Neutron Spectrum for ATR Fuel Element and ATR Flux-Trap Positions  

SciTech Connect

The Advanced Test Reactor (ATR) is a high power and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the high total core power and high neutron flux, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. An optimized low-enriched uranium (LEU) (U-10Mo) core conversion case, which can meet the project requirements, has been selected. However, LEU contains a significant quantity of high density U-238 (80.3 wt.%), which will harden the neutron spectrum in the core region. Based on the reference ATR HEU and the optimized LEU full core plate-by-plate (PBP) models, the present work investigates and compares the neutron spectra differences in the fuel element (FE), Northeast flux trap (NEFT), Southeast flux trap (SEFT), and East flux trap (EFT) positions. A detailed PBP MCNP ATR core model was developed and validated for fuel cycle burnup comparison analysis. The current ATR core with HEU U 235 enrichment of 93.0wt.% was used as the reference model. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm (20 mil). In this work, an optimized LEU (U-10Mo) core conversion case with a nominal fuel meat thickness of 0.330 mm (13 mil) and the U-235 enrichment of 19.7 wt.% was used to calculate the impact of the neutron spectrum in FE and FT positions. MCNP-calculated results show that the neutron spectrum in the LEU FE is slightly harder than in the HEU FE, as expected. However, when neutrons transport through water coolant and beryllium (Be), the neutrons are thermalized to an equilibrium neutron spectrum as a function of water volume fraction in the investigated FT positions. As a result, the neutron spectrum differences of the HEU and LEU in the NEFT, SEFT, and EFT are negligible. To demonstrate that the LEU core fuel cycle performance can meet the Updated Final Safety Analysis Report (UFSAR) safety requirements, additional studies will be necessary to evaluate and compare safety parameters such as void reactivity and Doppler coefficients, control components worth (outer shim control cylinders, safety rods and regulating rod), and shutdown margins between the HEU and LEU cores.

G. S. Chang

2008-10-01T23:59:59.000Z

35

Neutron flux measurements in the side-core region of Hunterston B advanced gas-cooled reactor  

Science Conference Proceedings (OSTI)

The core restraints of advanced gas-cooled reactors are important structural components that are required to maintain the geometric integrity of the cores. A review of neutron dosimetry for the sister stations Hunterston B and Hinkley Point B identified that earlier conservative assessments predicted high thermal neutron dose rates to key components of the restraint structure (the restraint rod welds), with the implication that some of them may be predicted to fail during a seismic event. A revised assessment was therefore undertaken [Thornton, D. A., Allen, D. A., Tyrrell, R. J., Meese, T. C., Huggon, A.P., Whiley, G. S., and Mossop, J. R., 'A Dosimetry Assessment for the Core Restraint of an Advanced Gas Cooled Reactor,' Proceedings of the 13. International Symposium on Reactor Dosimetry (ISRD-13, May 2008), World Scientific, River Edge, NJ, 2009, W. Voorbraak, L. Debarberis, and P. D'hondt, Eds., pp. 679-687] using a detailed 3D model and a Monte Carlo radiation transport program, MCBEND. This reassessment resulted in more realistic fast and thermal neutron dose recommendations, the latter in particular being much lower than had been thought previously. It is now desirable to improve confidence in these predictions by providing direct validation of the MCBEND model through the use of neutron flux measurements. This paper describes the programme of work being undertaken to deploy two neutron flux measurement 'stringers' within the side-core region of one of the Hunterston B reactors for the purpose of validating the MCBEND model. The design of the stringers and the determination of the preferred deployment locations have been informed by the use of detailed MCBEND flux calculations. These computational studies represent a rare opportunity to design a flux measurement beforehand, with the clear intention of minimising the anticipated uncertainties and obtaining measurements that are known to be representative of the neutron fields to which the vulnerable steel restraint components are exposed. (authors)

Allen, D.A. [Serco, Rutherford House, Quedgeley, Gloucester, GL2 4NF (United Kingdom); Shaw, S.E. [British Energy, Barnett Way, Barnwood, Gloucester, GL4 3RS (United Kingdom); Huggon, A.P.; Steadman, R.J.; Thornton, D.A. [Serco, Rutherford House, Quedgeley, Gloucester, GL2 4NF (United Kingdom); Whiley, G.S. [British Energy, Barnett Way, Barnwood, Gloucester, GL4 3RS (United Kingdom)

2011-07-01T23:59:59.000Z

36

DIRECT MEASUREMENT OF HEAT FLUX FROM COOLING LAKE THERMAL IMAGERY  

SciTech Connect

Laboratory experiments show a linear relationship between the total heat flux from a water surface to air and the standard deviation of the surface temperature field, {sigma}, derived from thermal images of the water surface over a range of heat fluxes from 400 to 1800 Wm{sup -2}. Thermal imagery and surface data were collected at two power plant cooling lakes to determine if the laboratory relationship between heat flux and {sigma} exists in large heated bodies of water. The heat fluxes computed from the cooling lake data range from 200 to 1400 Wm{sup -2}. The linear relationship between {sigma} and Q is evident in the cooling lake data, but it is necessary to apply band pass filtering to the thermal imagery to remove camera artifacts and non-convective thermal gradients. The correlation between {sigma} and Q is improved if a correction to the measured {sigma} is made that accounts for wind speed effects on the thermal convection. Based on more than a thousand cooling lake images, the correlation coefficients between {sigma} and Q ranged from about 0.8 to 0.9.

Garrett, A; Eliel Villa-Aleman, E; Robert Kurzeja, R; Malcolm Pendergast, M; Timothy Brown, T; Saleem Salaymeh, S

2007-12-19T23:59:59.000Z

37

Non-destructive assay of mechanical components using gamma-rays and thermal neutrons  

SciTech Connect

This work presents the results obtained in the inspection of several mechanical components through neutron and gamma-ray transmission radiography. The 4.46 Multiplication-Sign 10{sup 5} n.cm{sup -2}.s{sup -1} thermal neutron flux available at the main port of the Argonauta research reactor in Instituto de Engenharia Nuclear has been used as source for the neutron radiographic imaging. The 412 keV {gamma}-ray emitted by {sup 198}Au, also produced in that reactor, has been used as interrogation agent for the gamma radiography. Imaging Plates - IP specifically designed to operate with thermal neutrons or with X-rays have been employed as detectors and storage devices for each of these radiations.

Souza, Erica Silvani; Avelino, Mila R. [PPG-EM/UERJ, R. Sao Francisco Xavier, 524, Maracana - Rio de Janeiro - RJ (Brazil); Almeida, Gevaldo L. de; Souza, Maria Ines S. [IEN/CNEN, Rua Helio de Almeida, 75, Ilha do Fundao, Rio de Janeiro - RJ (Brazil)

2013-05-06T23:59:59.000Z

38

Measurements of neutron flux from an inertial-electrostatic confinement device  

SciTech Connect

A neutron-detection system was built for the purpose of measuring the neutron flux from an Inertial-Electrostatic Confinement Device located at Brigham Young University. A BF$sub 3$ proportional counter was used for absolute flux measurements and a pair of scintillation detectors was used to compare neutron output under different operating conditions. The detectors were designed to be compatible with the operating conditions of the device and to be able to measure small changes in neutron output. The detectors were calibrated using a Pu-Be source with corrections made for laboratory conditions. Performance of the counting system was checked and data were collected on the neutron flux from the device. (auth)

Westenskow, G.A.

1975-08-01T23:59:59.000Z

39

Neutron-flux profile monitor for use in a fission reactor  

DOE Patents (OSTI)

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.

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

1981-09-15T23:59:59.000Z

40

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

Science Conference Proceedings (OSTI)

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 diffraction and laid the foundation for an active neutron scattering effort that continued through the 1950s, using the Oak Ridge Research reactor after 1958, and, starting in 1966, the High Flux Isotope Reactor, or HFIR.

Nagler, Stephen E [ORNL; Mook Jr, Herbert A [ORNL

2008-01-01T23:59:59.000Z

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


41

Thermal Modeling and Feedback Requirements for LIFE Neutronic Simulations  

Science Conference Proceedings (OSTI)

An initial study is performed to determine how temperature considerations affect LIFE neutronic simulations. Among other figures of merit, the isotopic mass accumulation, thermal power, tritium breeding, and criticality are analyzed. Possible fidelities of thermal modeling and degrees of coupling are explored. Lessons learned from switching and modifying nuclear datasets is communicated.

Seifried, J E

2009-07-15T23:59:59.000Z

42

Gamma discrimination in pillar structured thermal neutron detectors  

Science Conference Proceedings (OSTI)

Solid-state thermal neutron detectors are desired to replace {sup 3}He tube based technology for the detection of special nuclear materials. {sup 3}He tubes have some issues with stability, sensitivity to microphonics and very recently, a shortage of {sup 3}He. There are numerous solid-state approaches being investigated that utilize various architectures and material combinations. By using the combination of high-aspect-ratio silicon PIN pillars, which are 2 {micro}m wide with a 2 {micro}m separation, arranged in a square matrix, and surrounded by {sup 10}B, the neutron converter material, a high efficiency thermal neutron detector is possible. Besides intrinsic neutron detection efficiency, neutron to gamma discrimination is an important figure of merit for unambiguous signal identification. In this work, theoretical calculations and experimental measurements are conducted to determine the effect of structure design of pillar structured thermal neutron detectors including: intrinsic layer thickness, pillar height, substrate doping and incident gamma energy on neutron to gamma discrimination.

Shao, Q; Radev, R P; Conway, A M; Voss, L F; Wang, T F; Nikolic, R J; Deo, N; Cheung, C L

2012-03-26T23:59:59.000Z

43

Thermal-neutron capture for A=36-44  

SciTech Connect

The prompt gamma-ray data of thermal- neutron captures fornuclear mass number A=26-35 had been evaluated and published in "ATOMICDATA AND NUCLEAR DATA TABLES, 26, 511 (1981)". Since that time the manyexperimental data of the thermal-neutron captures have been measured andpublished. The update of the evaluated prompt gamma-ray data is verynecessary for use in PGAA of high-resolution analytical prompt gamma-rayspectroscopy. Besides, the evaluation is also very needed in theEvaluated Nuclear Structure Data File, ENSDF, because there are no promptgamma-ray data in ENSDF. The levels, prompt gamma-rays and decay schemesof thermal-neutron captures fornuclides (26Mg, 27Al, 28Si, 29Si, 30Si,31P, 32S, 33S, 34S, and 35Cl) with nuclear mass number A=26-35 have beenevaluated on the basis of all experimental data. The normalizationfactors, from which absolute prompt gamma-ray intensity can be obtained,and necessary comments are given in the text. The ENSDF format has beenadopted in this evaluation. The physical check (intensity balance andenergy balance) of evaluated thermal-neutron capture data has been done.The evaluated data have been put into Evaluated Nuclear Structure DataFile, ENSDF. This evaluation may be considered as an update of the promptgamma-ray from thermal-neutron capture data tables as published in"ATOMIC DATA AND NUCLEAR DATA TABLES, 26, 511 (1981)".

Chunmei, Z.; Firestone, R.B.

2003-01-01T23:59:59.000Z

44

Modeling high-energy cosmic ray induced terrestrial and atmospheric neutron flux: A lookup table  

E-Print Network (OSTI)

Under current conditions, the cosmic ray spectrum incident on the Earth is dominated by particles with energies solar flares, supernovae and gamma ray bursts produce high energy cosmic rays (HECRs) with drastically higher energies. The Earth is likely episodically exposed to a greatly increased HECR flux from such events, some of which lasting thousands to millions of years. The air showers produced by HECRs ionize the atmosphere and produce harmful secondary particles such as muons and neutrons. Neutrons currently contribute a significant radiation dose at commercial passenger airplane altitude. With higher cosmic ray energies, these effects will be propagated to ground level. This work shows the results of Monte Carlo simulations quantifying the neutron flux due to high energy cosmic rays at various primary energies and altitudes. We provide here lookup tables that can be used to determine neutron fluxes from primaries with total energies 1 GeV - 1 PeV. By convolution, one can compute the neutron flux for any arbitrary CR spectrum. Our results demonstrate that deducing the nature of primaries from ground level neutron enhancements would be very difficult.

Andrew Overholt; Adrian Melott; Dimitra Atri

2012-06-22T23:59:59.000Z

45

Fabrication of Pillar-Structured Thermal Neutron Detectors  

SciTech Connect

Pillar detector is an innovative solid state device structure that leverages advanced semiconductor fabrication technology to produce a device for thermal neutron detection. State-of-the-art thermal neutron detectors have shortcomings in achieving simultaneously high efficiency, low operating voltage while maintaining adequate fieldability performance. By using a 3-dimensional silicon PIN diode pillar array filled with isotopic boron 10, ({sup 10}B) a high efficiency device is theoretically possible. The fabricated pillar structures reported in this work are composed of 2 {micro}m diameter silicon pillars with a 4 {micro}m pitch and pillar heights of 6 and 12 {micro}m. The pillar detector with a 12 {micro}m height achieved a thermal neutron detection efficiency of 7.3% at 2V.

Nikolic, R J; Conway, A M; Reinhardt, C E; Graff, R T; Wang, T F; Deo, N; Cheung, C L

2007-11-19T23:59:59.000Z

46

The prototype of a detector for monitoring the cosmic radiation neutron flux on ground  

SciTech Connect

This work presents a comparison between the results of experimental tests and Monte Carlo simulations of the efficiency of a detector prototype for on-ground monitoring the cosmic radiation neutron flux. The experimental tests were made using one conventional {sup 241}Am-Be neutron source in several incidence angles and the results were compared to that ones obtained with a Monte Carlo simulation made with MCNPX Code.

Lelis Goncalez, Odair; Federico, Claudio Antonio; Mendes Prado, Adriane Cristina; Galhardo Vaz, Rafael; Tizziani Pazzianotto, Mauricio [Instituto de Estudos Avancados - IEAv/DCTA - Sao Jose dos Campos, SP (Brazil); Semmler, Renato [Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP - Sao Paulo, SP (Brazil)

2013-05-06T23:59:59.000Z

47

Comparison of three-dimensional neutron flux calculations for Maine Yankee  

SciTech Connect

Calculations have been performed on the Maine Yankee Power Plant to obtain three-dimensional neutron fluxes using the spatial synthesis with the two-dimensional discrete ordinates code DORT, the three-dimensional discrete ordinates code THREEDANT and the three-dimensional Monte Carlo code MCNP. Neutron fluxes are compared for energies above 0.1 MeV and 1.0 MeV as well as dpa. Results were obtained at the Yankee dosimetry locations and special test regions within the pressure vessel, in the reactor cavity, and in a shield tank detector well.

Urban, W.T.; Crotzer, L.A.; Waters, L.S.; Parsons, D.K.; Alcouffe, R.E. [Los Alamos National Lab., NM (United States); Spinney, K.B.; Cacciapouti, R.J. [Yankee Atomic Electric Co., Bolton, MA (United States)

1996-10-01T23:59:59.000Z

48

Temperature, thermal-conductivity, and heat-flux data,Raft River...  

Open Energy Info (EERE)

Temperature, thermal-conductivity, and heat-flux data,Raft River area, Cassia County, Idaho (1974-1976) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Report:...

49

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

E-Print Network (OSTI)

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

Kempf, Stephanie Anne

2011-01-01T23:59:59.000Z

50

A unique method of neutron flux determination from experimental data  

DOE Patents (OSTI)

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

Paxton, Frank A.

1998-12-01T23:59:59.000Z

51

Ortho- and para-hydrogen in neutron thermalization  

DOE Green Energy (OSTI)

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.

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

1998-09-01T23:59:59.000Z

52

Development of the OSO-6 high-energy neutron detector and correlation of measured solar neutron fluxes to solar flares  

SciTech Connect

Thesis. The development of a directional high energy (20 to 160 MeV) neutron detector which was flown to satellite altitudes (500 km; circular equatorial orbit) in the NASA Orbiting Solar Observatory (OSO-6) in August 1969 is described. Both the angle of incidence and the energy of the neutron are determined by a proton-recoil telescope (Pilot B scintillation plastic) which provides the source for proton-recoils and defines the dE/dX versus E method for particle identification and energy determination. The telescope is embedded in a scintillation plastic guard counter envelope which eliminates the unwanted charged particle background as well as recoil protons (electrons) whose energies and direction do not satisfy neutron (gamma-ray) detection requirements, respectively. Results from a Monte Carlo calculation indicate that the overall average efficiency within an average angular acceptance of about 29 deg (FWHM) is approximately (2.25 plus or minus 0.113) x 10/sup -4/. The inflight calibration procedure, the main frame data bit error analysis, and the method for determining the orientation of the detector axis in the spacecraft spin plane are described. Results indicate a discrepancy in the measured (0.461 x 10/sup -2/ plus or minus 0.254 x 10/sup -2/ n/cm/sup 2/sec) and theoretical (2 to 70 n/cm/ sup 2/sec) neutron flux es which suggests a lack of basic underatanding of mechanisms leading to high energy neutron production at the sun. (auth)

Young, F.

1974-02-01T23:59:59.000Z

53

DIFFERENTIAL NEUTRON THERMALIZATION. Annual Summary Report, October 1, 1961 through September 30, 1962  

DOE Green Energy (OSTI)

Experimental and theoretical work on the interaction mechanisms by which neutrons exchange energy with H atoms involves treating neutron thermalization as neutron interactions with energy levels in the atoms. Cold moderators are presently being studied in order to optimize the source of cold neutrons. Cold neutrons are provided from an accelerator arrangement that directs electrons against a Fansteel target producing fast neutrons. Thermal neutrons, produced by moderation of fast neutrons, are passed through a chopper. Several moderators are evaluated, and neutron emission time measurements by crystal diffraction and beam chopper techniques point out emission time dependence on thickness, moderator, and temperature. The neutron beam chopper used presently is described, and results of neutron scattering by liquid para- and orthohydrogen are displayed and compared with theoretical predictions made with a perfect hydrogen gas model. Inelastic scattering of neutrons by liquid H is discussed, and theoretical and experimental results of inelastic scattering by polyethylene are also included. (D.C.W.)

Whittemore, W L

1962-11-28T23:59:59.000Z

54

Effect of high-energy neutron flux on fiber optics in an active diagnostic on TFTR  

Science Conference Proceedings (OSTI)

A bundle of 1-mm-diam fused silica optical fibers on an existing TFTR diagnostic has been exposed to 11 high-power DT discharges. Each shot subjected the fibers to a peak fast (14.7 MeV) neutron flux of [approx]2[times]10[sup 12] n/cm[sup 2]/s and a [gamma]-dose rate of 500 rad(Si)/s for 0.75--1.0 s. The total fast-neutron fluence for these shots was [approx]5[times]10[sup 12] n/cm[sup 2]. A 15-m-long section of the bundle ran along the tokamak's toroidal field coils and the remaining 15 m ran radially away from the reactor. Fiber luminescence at 660 nm was [approx]10[sup 10] photons/s/sr/cm[sup 2]/A for the above flux ([approx]5%--10% of the bremsstrahlung emission), and varied linearly with DT neutron rate. Luminescence at 530 nm was 50% stronger, consistent with a Cerenkov radiation spectrum. Sensitivity to 3.5 MeV DD neutrons was [approx]1/3 to 1/2 of that for DT neutrons. Fiber transmission decreased with the time integral of the neutron source rate and was reduced by 4% for the above flux. The fiber recovered rapidly: within 10 s, the transmission loss was only 2.5%. Shortly thereafter, the rate of recovery slowed to [approx]0.05% per minute, but was sufficient to restore 75% of the transmission loss within two to four discharges. Recovery continued at [approx]0.1% per hour and slowed overnight to [approx]0.1% per day. Within the relative error of [lt]0.2%, full transmission was recovered after five days.

Paul, S.F.; Goldstein, J.L.; Durst, R.D.; Fonck, R.J. (Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 (United States))

1995-02-01T23:59:59.000Z

55

Neutron Spectral Brightness of Cold Guide 4 at the High Flux Isotope Reactor  

DOE Green Energy (OSTI)

The High Flux Isotope Reactor resumed operation in June of 2007 with a super-critical hydrogen cold source in horizontal beam tube 4. Cold guide 4 is a guide system designed to deliver neutrons from this source at reasonable flux at wavelengths greater than 4 Å to several instruments, and includes a 15-m, 96-section, 4-channel bender. A time-of-flight spectrum with calibrated detector was recorded at port C of cold guide 4, and compared to McStas simulations, to generate a brightness spectrum.

Winn,B.L.; Robertson, J.L.; Iverson, E.B.; Selby, D.L.

2009-05-03T23:59:59.000Z

56

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

Science Conference Proceedings (OSTI)

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

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

2009-01-01T23:59:59.000Z

57

Quiescent thermal emission from neutron stars in LMXBs  

E-Print Network (OSTI)

The quiescent thermal emission from neutron stars in low mass X-ray binaries after active periods of intense activity in x-rays (outbursts) has been monitored. The theoretical modeling of the thermal relaxation of the neutron star crust may be used to establish constraints on the crust and envelope composition and transport properties, depending on the astrophysical scenarios assumed. We perform numerical simulations of the neutron star crust thermal evolution and compare them with inferred surface temperatures for five sources: MXB 1659-29, KS 1731-260, EXO 0748-676, XTE J1701-462 and IGR J17480-2446. We also present stationary envelope models to be used as a boundary condition for the crustal cooling models. We obtain a relation between the mass accretion rate and the temperature reached at the crust-envelope interface at the end of the active phase that accounts for early observations and reduces the number of free parameters of the problem. With this relation we are also able to set constraints to the envelope composition depending on the accretion mass rate. We find that the evolution of MXB 1659-29, KS 1731-260 and EXO 0748-676 can be well described within a deep crustal cooling scenario. Conversely, we find that other two sources can only be explained with models beyond crustal cooling. For the peculiar emission of XTE J1701-462 we propose alternative scenarios like residual accretion during quiescence, additional heat sources in the outer crust and/or thermal isolation of the inner crust due to a buried magnetic field. We also explain the very recent reported temperature of IGR J17480-2446 with an extra heat deposition in the outer crust coming from shallow sources.

Anabela Turlione; Deborah N. Aguilera; José A. Pons

2013-09-16T23:59:59.000Z

58

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

Science Conference Proceedings (OSTI)

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

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

2011-03-01T23:59:59.000Z

59

Temperature, thermal-conductivity, and heat-flux data,Raft River area,  

Open Energy Info (EERE)

Temperature, thermal-conductivity, and heat-flux data,Raft River area, Temperature, thermal-conductivity, and heat-flux data,Raft River area, Cassia County, Idaho (1974-1976) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Report: Temperature, thermal-conductivity, and heat-flux data,Raft River area, Cassia County, Idaho (1974-1976) Details Activities (1) Areas (1) Regions (0) Abstract: Basin and Range Province; Cassia County Idaho; economic geology; exploration; geophysical surveys; geothermal energy; heat flow; heat flux; Idaho; North America; Raft River basin; south-central Idaho; surveys; temperature; thermal conductivity; United States; USGS Author(s): Urban, T.C.; Diment, W.H.; Nathenson, M.; Smith, E.P.; Ziagos, J.P.; Shaeffer, M.H. Published: Open-File Report - U. S. Geological Survey, 1/1/1986 Document Number: Unavailable

60

Parameterization of the Dry Convective Boundary Layer Based on a Mass Flux Representation of Thermals  

Science Conference Proceedings (OSTI)

Presented is a mass flux parameterization of vertical transport in the convective boundary layer. The formulation of the new parameterization is based on an idealization of thermal cells or rolls. The parameterization is validated by comparison ...

Frédéric Hourdin; Fleur Couvreux; Laurent Menut

2002-03-01T23:59:59.000Z

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


61

The Effect of Soil Thermal Conductivity Parameterization on Surface Energy Fluxes and Temperatures  

Science Conference Proceedings (OSTI)

The sensitivity of sensible and latent heat fluxes and surface temperatures to the parameterization of the soil thermal conductivity is demonstrated using a soil vegetation atmosphere transfer scheme (SVATS) applied to intensive field campaigns (...

C. D. Peters-Lidard; E. Blackburn; X. Liang; E. F. Wood

1998-04-01T23:59:59.000Z

62

Heat and Momentum Fluxes Induced by Thermal Inhomogeneities with and without Large-Scale Flow  

Science Conference Proceedings (OSTI)

The authors Present an analytical evaluation of the vertical heat and momentum fluxes associated with mesoscale flow generated by periodic and isolated thermal inhomogeneities within the convective boundary layer (CBL). The influence of larger-...

G. A. Dalu; R. A. Pielke; M. Baldi; X. Zeng

1996-11-01T23:59:59.000Z

63

New aspects in the evaluation of thermal neutron cross sections  

Science Conference Proceedings (OSTI)

Because of recent advances in experimental techniques, which improved the accuracies of thermal capture and scattering cross sections by an order of magnitude, a more stringent approach in the evaluation of the thermal constants is developed. In the present approach, the following aspects are introduced: (1) a consistency between thermal cross sections, coherent and incoherent scattering lengths, and neutron resonance parameters is achieved; (2) a consistency between the isotopic and element cross sections is sought; in addition, for each isotope, the requirement that the partial cross sections add up to the total is fulfilled; (3) where possible, charged particle data particularly derived from (d,p) reactions on light and medium weight isotopes are used in locating the positions and strengths of bound levels. Such a procedure is useful in the evaluation of the shape of the cross sections in the thermal region; and (4) the Lane-Lynn theory of direct capture is called upon to calculate thermal cross sections and check for consistencies for certain isotopes. Extensive examples to illustrate these procedures are presented.

Mughabghab, S.F.

1980-01-01T23:59:59.000Z

64

Investigation of the MTC noise estimation with a coupled neutronic/thermal-hydraulic dedicated model - 'Closing the loop'  

Science Conference Proceedings (OSTI)

This paper investigates the reliability of different noise estimators aimed at determining the Moderator Temperature Coefficient (MTC) of reactivity in Pressurized Water Reactors. By monitoring the inherent fluctuations in the neutron flux and moderator temperature, an on-line monitoring of the MTC without perturbing reactor operation is possible. In order to get an accurate estimation of the MTC by noise analysis, the point-kinetic component of the neutron noise and the core-averaged moderator temperature noise have to be used. Because of the scarcity of the in-core instrumentation, the determination of these quantities is difficult, and several possibilities thus exist for estimating the MTC by noise analysis. Furthermore, the effect of feedback has to be negligible at the frequency chosen for estimating the MTC in order to get a proper determination of the MTC. By using an integrated neutronic/thermal- hydraulic model specifically developed for estimating the three-dimensional distributions of the fluctuations in neutron flux, moderator properties, and fuel temperature, different approaches for estimating the MTC by noise analysis can be tested individually. It is demonstrated that a reliable MTC estimation can only be provided if the core is equipped with a sufficient number of both neutron detectors and temperature sensors, i.e. if the core contain in-core detectors monitoring both the axial and radial distributions of the fluctuations in neutron flux and moderator temperature. It is further proven that the effect of feedback is negligible for frequencies higher than 0.1 Hz, and thus the MTC noise estimations have to be performed at higher frequencies. (authors)

Demaziere, C.; Larsson, V. [Div. of Nuclear Engineering, Dept. of Applied Physics, Chalmers Univ. of Technology, SE-41296 Gothenburg (Sweden)

2012-07-01T23:59:59.000Z

65

The High Flux Isotope Reactor at Oak Ridge National Laboratory  

NLE Websites -- All DOE Office Websites

The High Flux Isotope Reactor at ORNL The High Flux Isotope Reactor at ORNL Aerial of the High Flux Isotope Reactor Site The High Flux Isotope Reactor site is located on the south side of the ORNL campus and is about a three-minute drive from her sister neutron facility, the Spallation Neutron Source. Operating at 85 MW, HFIR is the highest flux reactor-based source of neutrons for research in the United States, and it provides one of the highest steady-state neutron fluxes of any research reactor in the world. The thermal and cold neutrons produced by HFIR are used to study physics, chemistry, materials science, engineering, and biology. The intense neutron flux, constant power density, and constant-length fuel cycles are used by more than 500 researchers each year for neutron scattering research into

66

Neutron Flux Interpolation with Finite Element Method in the Nuclear Fuel Cell Calculation using Collision Probability Method  

SciTech Connect

Nuclear reactor design and analysis of next-generation reactors require a comprehensive computing which is better to be executed in a high performance computing. Flat flux (FF) approach is a common approach in solving an integral transport equation with collision probability (CP) method. In fact, the neutron flux distribution is not flat, even though the neutron cross section is assumed to be equal in all regions and the neutron source is uniform throughout the nuclear fuel cell. In non-flat flux (NFF) approach, the distribution of neutrons in each region will be different depending on the desired interpolation model selection. In this study, the linear interpolation using Finite Element Method (FEM) has been carried out to be treated the neutron distribution. The CP method is compatible to solve the neutron transport equation for cylindrical geometry, because the angle integration can be done analytically. Distribution of neutrons in each region of can be explained by the NFF approach with FEM and the calculation results are in a good agreement with the result from the SRAC code. In this study, the effects of the mesh on the k{sub eff} and other parameters are investigated.

Shafii, M. Ali [Departmen of Physics Bandung Institute of Technology, Jl. Ganesha 10, Bandung 40134 (Indonesia); Physics Department, Andalas University, Kampus Limau Manis, Padang, Sumatera Barat (Indonesia); Su'ud, Zaki; Waris, Abdul; Kurniasih, Neny [Departmen of Physics Bandung Institute of Technology, Jl. Ganesha 10, Bandung 40134 (Indonesia); Ariani, Menik [Physics Department, Sriwijaya University, Kampus Indralaya, Ogan Ilir, Sumatera Selatan (Indonesia); Departmen of Physics Bandung Institute of Technology, Jl. Ganesha 10, Bandung 40134 (Indonesia); Yulianti, Yanti [Physics Department, Lampung University, Jl.Sumantri Brojonegoro no 1, Lampung (Indonesia); Departmen of Physics Bandung Institute of Technology, Jl. Ganesha 10, Bandung 40134 (Indonesia)

2010-12-23T23:59:59.000Z

67

Supercool Neutrons (Ultracold Neutrons)  

E-Print Network (OSTI)

in the USA. Why neutrons? Neutrons possess physical properties that make them valuable investigative tools Spallation Neutron Source (SNS) The world's most intense pulsed accelerator-based neutron source. High Flux Isotope Reactor (HFIR) The highest flux reactor-based neutron source for condensed matter research

Martin, Jeff

68

Thermal Neutron Capture y's (CapGam)  

DOE Data Explorer (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)

69

Method for measuring dose-equivalent in a neutron flux with an unknown energy spectra and means for carrying out that method  

DOE Patents (OSTI)

A method for measuring the dose-equivalent for exposure to an unknown and/or time varing neutron flux which comprises simultaneously exposing a plurality of neutron detecting elements of different types to a neutron flux and combining the measured responses of the various detecting elements by means of a function, whose value is an approximate measure of the dose-equivalent, which is substantially independent of the energy spectra of the flux. Also, a personnel neutron dosimeter, which is useful in carrying out the above method, comprising a plurality of various neutron detecting elements in a single housing suitable for personnel to wear while working in a radiation area.

Distenfeld, Carl H. (Mattituck, NY)

1978-01-01T23:59:59.000Z

70

CONSTRUCTION OF A 14 Mev NEUTRON GENERATOR UTILIZING T$sup 3$(d,n)He$sup 4$ REACTION AND MEASUREMENT OF FAST NEUTRON FLUX  

SciTech Connect

Construction of a low-voltage accelerating machine for accelerating deuterons which is utilized as a source of production of 14-Mev neutrons by T/sup 3/(d,n)He/sup 4/ reaction is described. Neutron counting has been done by counting recoil protons in a suitable scintillation counter. Neutron yield hss also been measured indirectly from saturated activity of an irradiated thin silver foil. Increase in relative flux with increasing deuteron energy trom 30 kev to 0.1 Mev has been obtained. (auth)

Mitra, B.

1959-04-01T23:59:59.000Z

71

The impact of magnetic field on the thermal evolution of neutron stars  

E-Print Network (OSTI)

The impact of strong magnetic fields B>10e13 G on the thermal evolution of neutron stars is investigated, including crustal heating by magnetic field decay. For this purpose, we perform 2D cooling simulations with anisotropic thermal conductivity considering all relevant neutrino emission processes for realistic neutron stars. The standard cooling models of neutron stars are called into question by showing that the magnetic field has relevant (and in many cases dominant) effects on the thermal evolution. The presence of the magnetic field significantly affects the thermal surface distribution and the cooling history of these objects during both, the early neutrino cooling era and the late photon cooling era. The minimal cooling scenario is thus more complex than generally assumed. A consistent magneto-thermal evolution of magnetized neutron stars is needed to explain the observations.

Deborah N. Aguilera; José A. Pons; Juan A. Miralles

2007-12-09T23:59:59.000Z

72

The impact of magnetic field on the thermal evolution of neutron stars  

E-Print Network (OSTI)

The impact of strong magnetic fields B>10e13 G on the thermal evolution of neutron stars is investigated, including crustal heating by magnetic field decay. For this purpose, we perform 2D cooling simulations with anisotropic thermal conductivity considering all relevant neutrino emission processes for realistic neutron stars. The standard cooling models of neutron stars are called into question by showing that the magnetic field has relevant (and in many cases dominant) effects on the thermal evolution. The presence of the magnetic field significantly affects the thermal surface distribution and the cooling history of these objects during both, the early neutrino cooling era and the late photon cooling era. The minimal cooling scenario is thus more complex than generally assumed. A consistent magneto-thermal evolution of magnetized neutron stars is needed to explain the observations.

Aguilera, Deborah N; Miralles, Juan A

2007-01-01T23:59:59.000Z

73

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

SciTech Connect

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

Mo, S. C.

1998-01-14T23:59:59.000Z

74

METHOD OF TESTING THERMAL NEUTRON FISSIONABLE MATERIAL FOR PURITY  

DOE Patents (OSTI)

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.

Fermi, E.; Anderson, H.L.

1961-01-24T23:59:59.000Z

75

A U.S. high-flux neutron facility for fusion materials development  

SciTech Connect

Materials for a fusion reactor first wall and blanket structure must be able to reliably function in an extreme environment that includes 10-15 MW-year/m{sup 2} neutron and heat fluences. The various materials and structural challenges are as difficult and important as achieving a burning plasma. Overcoming radiation damage degradation is the rate-controlling step in fusion materials development. Recent advances with oxide dispersion strengthened ferritic steels show promise in meeting reactor requirements, while multi-timescale atomistic simulations of defect-grain boundary interactions in model copper systems reveal surprising self-annealing phenomenon. While these results are promising, simultaneous evaluation of radiation effects displacement damage ({le} 200 dpa) and in-situ He generation ({le} 2000 appm) at prototypical reactor temperatures and chemical environments is still required. There is currently no experimental facility in the U.S. that can meet these requirements for macroscopic samples. The E.U. and U.S. fusion communities have recently concluded that a fusion-relevant, high-flux neutron source for accelerated characterization of the effects of radiation damage to materials is a top priority for the next decade. Data from this source will be needed to validate designs for the multi-$B next-generation fusion facilities such as the CTF, ETF, and DEMO, that are envisioned to follow ITER and NIF.

Rei, Donald J [Los Alamos National Laboratory

2010-01-01T23:59:59.000Z

76

MSc thesis topic: Coupled Thermal-hydraulic MC neutronic calculations  

E-Print Network (OSTI)

Institute of Nuclear Physics the source of epithermal neutrons based on a vacuum insulation tandem At the Budker Institute of Nuclear Physics the VITA-facility for the boron neutron capture therapyNew technical solution for using the time-of-flight technique to measure neutron spectra V. Aleynik

Haviland, David

77

Report on Thermal Neutron Diffusion Length Measurement in Reactor Grade Graphite Using MCNP and COMSOL Multiphysics  

E-Print Network (OSTI)

Neutron diffusion length in reactor grade graphite is measured both experimentally and theoretically. The experimental work includes Monte Carlo (MC) coding using 'MCNP' and Finite Element Analysis (FEA) coding suing 'COMSOL Multiphysics' and Matlab. The MCNP code is adopted to simulate the thermal neutron diffusion length in a reactor moderator of 2m x 2m with slightly enriched uranium ($^{235}U$), accompanied with a model designed for thermal hydraulic analysis using point kinetic equations, based on partial and ordinary differential equation. The theoretical work includes numerical approximation methods including transcendental technique to illustrate the iteration process with the FEA method. Finally collision density of thermal neutron in graphite is measured, also specific heat relation dependability of collision density is also calculated theoretically, the thermal neutron diffusion length in graphite is evaluated at $50.85 \\pm 0.3cm$ using COMSOL Multiphysics and $50.95 \\pm 0.5cm$ using MCNP. Finally ...

Mirfayzi, S R

2013-01-01T23:59:59.000Z

78

NEUTRONIC AND THERMAL HYDRAULIC DESIGNS OF ANNULAR FUEL FOR HIGH POWER DENSITY BWRS  

E-Print Network (OSTI)

As a promising new fuel for high power density light water reactors, the feasibility of using annular fuel for BWR services is explored from both thermal hydraulic and neutronic points of view. Keeping the bundle size ...

Morra, P.

79

Study of gamma cascades of 59Ni by thermal neutron reaction  

E-Print Network (OSTI)

The quantum properties are important to study nuclear structure. The energy, spin, parity, transition order are usually interesting to research. In this experiment, 59Ni is activated by thermal neutron on 3rd horizontal channel of Dalat nuclear Reactor.

Nguyen An Son

2013-10-07T23:59:59.000Z

80

Event-by-event study of neutron observables in spontaneous and thermal fission  

E-Print Network (OSTI)

The event-by-event fission model FREYA is extended to spontaneous fission of actinides and a variety of neutron observables are studied for spontaneous fission and fission induced by thermal neutrons with a view towards possible applications for detection of special nuclear materials.

Vogt, R

2011-01-01T23:59:59.000Z

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


81

Event-by-event study of neutron observables in spontaneous and thermal fission  

E-Print Network (OSTI)

The event-by-event fission model FREYA is extended to spontaneous fission of actinides and a variety of neutron observables are studied for spontaneous fission and fission induced by thermal neutrons with a view towards possible applications for detection of special nuclear materials.

R. Vogt; J. Randrup

2011-09-17T23:59:59.000Z

82

MASS YIELDS FROM FISSION BY NEUTRONS BETWEEN THERMAL AND 14.7 MEV  

SciTech Connect

Radiochemically determined mass-yield curves are given for the fission of U/sup 235/ and U/sup 238/ by l4.7-Mev neutrons. Symmetric and to a less extent, very asymmetric modes of fission are more probable at that energy than in thermal fission. Yields of four fission products from the fission of U/sup 235/ have been measured as a function of neutron energy in the range thermal to 14 Mev. The yields of eleven masses have been measured from the fission of Np/sup 237/ by degraded fission spectrum neutrons. The mass-yield curve is similar to that from the thermal fission of Pu/sup 239/ with a ratio of peak to valley yields of approximately 175. Relative yields of one peak fission product and four valley fission products have been determdned under the following conditions: fission of U/sup 235/ and Pu/sup 239/ with thermal neutrons; fission of U/sup 235/ Pu/sup 239/ and U/sup 238/ with fission spectrum neutrons; and fission of U//sup 235/ and Pu/sup 239/ with the intermediate neutron spectrum at the center of the Los Alamos Fast Reactor. Absolute yields of Moss have been determined from the fission of U/sup 235/,Pu/sup 239/ with thermal neutrons. (auth)

Ford, G.P.; Gilmore, J.S.; Ames, D.P.; Balagna, J.P.; Barnes, J.W.; Comstock, A.A.; Cowan, G.A.; Elkin, P.B.; Hoffman, D.C.; Knobeloch, G.W.; Lang, E.J.; Melnick, M.A.; Minkkinen, C.O.; Pollock, B.D.; Sattizshn, J.E.; Stanley, C.W.; Warren, B.

1956-02-01T23:59:59.000Z

83

NEUTRONIC REACTOR  

DOE Patents (OSTI)

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

Wigner, E.P.

1958-04-22T23:59:59.000Z

84

Mid-latitude composition of mars from thermal and epithermal neutrons  

DOE Green Energy (OSTI)

Epithermal neutron data acquired by Mars Odyssey have been analyzed to determine global maps of water-equivalent hydrogen abundance. By assuming that hydrogen was distributed uniformly with depth within the surface, a map of minimum water abundance was obtained. The addition of thermal neutrons to this analysis could provide information needed to determine water stratigraphy. For example, thermal and epithermal neutrons have been used together to determine the depth and abundance of waterequivalent hydrogen of a buried layer in the south polar region. Because the emission of thermal neutrons from the Martian surface is sensitive to absorption by elements other than hydrogen, analysis of stratigraphy requires that the abundance of these elements be known. For example, recently published studies of the south polar region assumed that the Mars Pathfinder mean soil composition is representative of the regional soil composition, This assumption is partially motivated by the fact that Mars appears to have a well-mixed global dust cover and that the Pathfinder soil composition is representative of the mean composition of the Martian surface. In this study, we have analyzed thermal and epithermal neutron data measured by the neutron spectrometer subsystem of the gamma ray spectrometer to determine the spatial distribution of the composition of elements other than hydrogen. We have restricted our analysis to mid-latitude regions for which we have corrected the neutron counting data for variations in atmospheric thickness.

Prettyman, T. H. (Thomas H.); Feldman, W. C. (William C.); Elphic, R. C. (Richard C.); Boynton, W. V. (William V.); Bish, D. L. (David L.); Vaniman, D. T. (David T.); Funsten, H. O. (Herbert O.); Lawrence, David J. (David Jeffery),; Maurice, S. (Sylvestre); McKinney, G. W. (Gregg W.); Moore, K. R. (Kurt R.); Tokar, R. L. (Robert L.)

2003-01-01T23:59:59.000Z

85

Experimental and Computational Study of the Flux Spectrum in Materials Irradiation Facilities of the High Flux Isotope Reactor  

Science Conference Proceedings (OSTI)

This report compares the available experimental neutron flux data in the High Flux Isotope Reactor (HFIR) to computational models of the HFIR loosely based on the experimental loading of cycle 400. Over the last several decades, many materials irradiation experiments have included fluence monitors which were subsequently used to reconstruct a coarse-group energy-dependent flux spectrum. Experimental values for thermal and fast neutron flux in the flux trap about the midplane are found to be 1.78 0.27 and 1.05 0:06 1E15 n/cm sec, respectively. The reactor physics code MCNP is used to calculate neutron flux in the HFIR at irradiation locations. The computational results are shown to correspond to closely to experimental data for thermal and fast neutron flux with calculated percent differences ranging from 0:55 13.20%.

McDuffee, Joel Lee [ORNL; Daly, Thomas F [ORNL

2012-01-01T23:59:59.000Z

86

Effect of high-energy neutron flux on fiber optics in an active diagnostic on TFTR (abstract)[sup a  

SciTech Connect

A bundle of 1-mm-diam fused silica optical fibers on an existing TFTR diagnostic has been exposed to 11 high-power DT discharges. Each shot subjected the fibers to a peak fast (14.7 MeV) neutron flux of [approx]2[times]10[sup 12] n/cm[sup 2]/s and a [gamma]-dose rate of 500 rad(Si)/s for 0.75--1.0 s. The total fast-neutron fluence for these shots was [approx]5[times]10[sup 12] n/cm[sup 2]. A 15-m-long section of the bundle ran along the tokamak's toroidal field coils and the remaining 15 m ran radially away from the reactor. Fiber luminescence at 660 nm was [approx]10[sup 10] photons/s/sr/cm[sup 2]/A for the above flux ([approx]5%--10% of the bremsstrahlung emission), and varied linearly with DT neutron rate. Luminescence at 530 nm was 50% stronger, consistent with a Cerenkov radiation spectrum. Sensitivity to 3.5 MeV DD neutrons was [approx]1/3 to 1/2 of that for DT neutrons. Fiber transmission decreased with the time integral of the neutron source rate and was reduced by 4% for the above flux. The fiber recovered rapidly: within 10 s, the transmission loss was only 2.5%. Shortly thereafter, the rate of recovery slowed to [approx]0.05% per minute, but was sufficient to restore 75% of the transmission loss within two to four discharges. Recovery continued at [approx]0.1% per hour and slowed overnight to [approx]0.1% per day. Within the relative error of [lt]0.2%, full transmission was recovered after five days.

Paul, S.F.; Goldstein, J.L.; Durst, R.D.; Fonck, R.J. (Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 (United States))

1995-01-01T23:59:59.000Z

87

Neutron beam characterization measurements at the Manuel Lujan Jr. neutron scattering center  

DOE Green Energy (OSTI)

We have measured the neutron beam characteristics of neutron moderators at the Manuel Lujan Jr. Neutron Scattering Center at LANSCE. The absolute thermal neutron flux, energy spectra and time emission spectra were measured for the high resolution and high intensity decoupled water, partially coupled liquid hydrogen and partially coupled water moderators. The results of our experimental study will provide an insight into aging of different target-moderator-reflector-shield components as well as new experimental data for benchmarking of neutron transport codes.

Mocko, Michal [Los Alamos National Laboratory; Muhrer, Guenter [Los Alamos National Laboratory; Daemen, Luke L [Los Alamos National Laboratory; Kelsey, Charles T [Los Alamos National Laboratory; Duran, Michael A [Los Alamos National Laboratory; Tovesson, Fredrik K [Los Alamos National Laboratory

2010-01-01T23:59:59.000Z

88

Optimizing diode thickness for thin-film solid state thermal neutron detectors  

Science Conference Proceedings (OSTI)

In this work, we investigate the optimal thickness of a semiconductor diode for thin-film solid state thermal neutron detectors. We evaluate several diode materials, Si, CdTe, GaAs, C (diamond), and ZnO, and two neutron converter materials, {sup 10}B and {sup 6}LiF. Investigating a coplanar diode/converter geometry, we determine the minimum semiconductor thickness needed to achieve maximum neutron detection efficiency. By keeping the semiconductor thickness to a minimum, gamma rejection is kept as high as possible. In this way, we optimize detector performance for different thin-film semiconductor materials.

Murphy, John W.; Mejia, Israel; Quevedo-Lopez, Manuel A.; Gnade, Bruce [Department of Materials and Science, University of Texas at Dallas, Richardson, Texas 75080 (United States); Kunnen, George R.; Allee, David [Flexible Display Center at Arizona State University, Tempe, Arizona 85284 (United States)

2012-10-01T23:59:59.000Z

89

Report on Thermal Neutron Diffusion Length Measurement in Reactor Grade Graphite Using MCNP and COMSOL Multiphysics  

E-Print Network (OSTI)

Neutron diffusion length in reactor grade graphite is measured both experimentally and theoretically. The experimental work includes Monte Carlo (MC) coding using 'MCNP' and Finite Element Analysis (FEA) coding suing 'COMSOL Multiphysics' and Matlab. The MCNP code is adopted to simulate the thermal neutron diffusion length in a reactor moderator of 2m x 2m with slightly enriched uranium ($^{235}U$), accompanied with a model designed for thermal hydraulic analysis using point kinetic equations, based on partial and ordinary differential equation. The theoretical work includes numerical approximation methods including transcendental technique to illustrate the iteration process with the FEA method. Finally collision density of thermal neutron in graphite is measured, also specific heat relation dependability of collision density is also calculated theoretically, the thermal neutron diffusion length in graphite is evaluated at $50.85 \\pm 0.3cm$ using COMSOL Multiphysics and $50.95 \\pm 0.5cm$ using MCNP. Finally the total neutron cross-section is derived using FEA in an inverse iteration form.

S. R. Mirfayzi

2013-01-08T23:59:59.000Z

90

Light-water-reactor coupled neutronic and thermal-hydraulic codes  

Science Conference Proceedings (OSTI)

An overview is presented of computer codes that model light water reactor cores with coupled neutronics and thermal-hydraulics. This includes codes for transient analysis and codes for steady state analysis which include fuel depletion and fission product buildup. Applications in nuclear design, reactor operations and safety analysis are given and the major codes in use in the USA are identified. The neutronic and thermal-hydraulic methodologies and other code features are outlined for three steady state codes (PDQ7, NODE-P/B and SIMULATE) and four dynamic codes (BNL-TWIGL, MEKIN, RAMONA-3B, RETRAN-02). Speculation as to future trends with such codes is also presented.

Diamond, D.J.

1982-01-01T23:59:59.000Z

91

Flux  

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

5000 5000 6000 7000 8000 Wavelength (Angstroms) Flux (in arbitrary units) SN 1990N SN 1989B SN 1993O SN 1981B SN 1994D SN 1997ap Iron Peak Blends Ca II Si II & Co II Fe II & III Day -7 Day -5 Day -4 Day -2 ± 2 Day 0 Day +2 * -50 0 50 100 150 Observed days from peak Observed I magnitude 27 26 25 24 23 Observed R magnitude 27 26 25 24 Observed I magnitude 27 26 25 24 23 R band Ground-based I band HST I band (b) (c) (a) Pre-SN observation 3.5 4.0 4.5 5.0 5.5 log(cz) 14 16 18 20 22 24 26 effective m B 0.02 0.05 0.1 0.2 0.5 1.0 redshift z Hamuy et al (A.J. 1996) Supernova Cosmology Project 6 8 % 9 0 % 0.5 1.0 1.5 2.0 2.5 3.0 ! M Age < 9.6 Gyr (H = 50 km s -1 Mpc -1 ) No Big Bang 0.0 0.5 1.0 1.5 2.0 2.5 3.0 -3 -2 -1 0 1 2 3 -3 -2 -1 0 1 2 3 ! " z ~ 0 . 4 z = 0 . 8 3 6 8 % 9 0 % 0.5 1.0 1.5 2.0 2.5 3.0 ! M Age < 9.6 Gyr (H=50 km/s/Mpc)

92

SHARP Neutronics Expanded | Department of Energy  

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

SHARP Neutronics Expanded SHARP 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 existing high-fidelity solver package was extended to be independent of reactor technology and demonstrated with 2-D MOC and Sn method simulations of LWR core configurations. Efforts to support verification and validation of the DeCART code, used as one reference solution method by the SHARP code

93

SPHERES, Juelich's high-flux neutron backscattering spectrometer at FRM II  

SciTech Connect

SPHERES is a third-generation neutron backscattering spectrometer, located at the 20 MW German neutron source FRM II and operated by the Juelich Centre for Neutron Science. It offers an energy resolution (fwhm) better than 0.65 {mu}eV, a dynamic range of {+-} 31 {mu}eV, and a signal-to-noise ratio of up to 1750:1.

Wuttke, Joachim; Budwig, Alfred; Drochner, Matthias; Kaemmerling, Hans; Kayser, Franz-Joseph; Kleines, Harald; Ossovyi, Vladimir; Pardo, Luis Carlos; Prager, Michael; Richter, Dieter; Schneider, Gerald J.; Schneider, Harald; Staringer, Simon [Forschungszentrum Juelich GmbH, 52425 Juelich (Germany)

2012-07-15T23:59:59.000Z

94

On the participant-spectator matter and thermalization of neutron-rich systems in heavy-ion collisions  

E-Print Network (OSTI)

We study the participant-spectator matter at the energy of vanishing flow for neutron-rich systems. Our study reveals similar behaviour of articipant-spectator for neutron-rich systems as for stable systems and also points towards nearly mass independence behaviour of participant-spectator matter for neutron-rich systems at the energy of vanishing flow. We also study the thermalization reached in the reactions of neutron-rich systems.

Sakshi Gautam; Rajeev K. Puri

2011-07-27T23:59:59.000Z

95

Portable Gamma and Thermal Neutron Detector Using LiI(Eu) Crystals  

Science Conference Proceedings (OSTI)

Simultaneous detection of gamma rays and neutrons emanating from an unknown source has been of special significance and importance to consequence management and first responders since the earliest days of the program. Bechtel Nevada scientists have worked with 6LiI(Eu) crystals and 6Li glass to develop field-operable dual sensors that detect gamma rays and neutrons simultaneously. The prototype 6LiI(Eu) counter, which has been built and tested, is designed to collect data for periods of one second to more than eight hours. The collection time is controlled by thumbwheel switches. A fourpole, high pass filter at 90 KHz reduces microphonic noise from shock or vibration. 6LiI(Eu) crystals generate completely separable gamma-ray and thermal neutron responses. The 6LiI(Eu) rate meter consists of a single crystal 3.8 x 3.8 cm (1.5 x 1.5 in) with a 2.54-cm-(1-in-) thick, annular, high-density, polyethylene ring around the cylinder. Special features are (1) thermal and epithermal neutron detection (0.025eV to 250keV) and (2) typical gamma resolution of 8% at 661.6 keV. Monte Carlo N-Particle calculations for characteristics of gamma spectral behavior, neutron attenuation length, relative neutron and gamma detection efficiency are reported.

S. Mukhopadhyay

2003-06-01T23:59:59.000Z

96

MEASUREMENT OF THE HUMIDITY OF SOILS BY DIFFUSION OF A BEAM OF THERMAL NEUTRONS  

SciTech Connect

From earlier results on the measurement of soil humidity an apparatus was constructed and calibrated for the measurement of the humidity of soils by diffusion of a beam of thermal neutrons. The construction and calibration of this apparatus are described in detail. (J.S.R.)

Wack, B.

1962-02-01T23:59:59.000Z

97

MCNP modeling of the Swiss LWRs for the calculation of the in- and ex-vessel neutron flux distributions  

SciTech Connect

MCNP models of all Swiss Nuclear Power Plants have been developed by the National Cooperative for the Disposal of Radioactive Waste (Nagra), in collaboration with the utilities and ETH Zurich, for the 2011 decommissioning cost study. The estimation of the residual radionuclide inventories and corresponding activity levels of irradiated structures and components following the NPP shut-down is of crucial importance for the planning of the dismantling process, the waste packaging concept and, consequently, for the estimation of the decommissioning costs. Based on NPP specific data, the neutron transport simulations lead to the best yet knowledge of the neutron spectra necessary for the ensuing activation calculations. In this paper, the modeling concept towards the MCNP-NPPs is outlined and the resulting flux distribution maps are presented. (authors)

Pantelias, M.; Volmert, B.; Caruso, S. [National Cooperative for the Disposal of Radioactive Waste Nagra, Hardstrasse 73, 5430, Wettingen (Switzerland); Zvoncek, P. [Laboratory for Nuclear Energy Systems, ETH Zurich, Sonneggstrasse 3, 8092, Zurich (Switzerland); Bitterli, B. [Kernkraftwerk Goesgen-Daeniken AG, 4658 Daeniken (Switzerland); Neukaeter, E.; Nissen, W. [BKW FMB Energie AG-Kernkraftwerk Muehleberg, 3203 Muehleberg (Switzerland); Ledergerber, G. [Kernkraftwerk Leibstadt AG, 5325 Leibstadt (Switzerland); Vielma, R. [Axpo AG-Kernkraftwerk Beznau, 5312 Doettingen (Switzerland)

2012-07-01T23:59:59.000Z

98

Neutron Flux Measurements and Calculations in the Gamma Irradiation Facility Using MCNPX.  

E-Print Network (OSTI)

??The gamma irradiation facility at the High Flux Isotope Reactor (HFIR)is used to deliver a pure gamma dose to any target of interest. in addition… (more)

Giuliano, Dominic Richard

2010-01-01T23:59:59.000Z

99

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

DOE Patents (OSTI)

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel 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.

Schreiber, Roger B. (Penn Twp., PA); Fero, Arnold H. (New Kensington, PA); Sejvar, James (Murrysville, PA)

1997-01-01T23:59:59.000Z

100

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

DOE Patents (OSTI)

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel 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.

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

1997-12-16T23:59:59.000Z

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


101

Neutronic and thermal design considerations for heat-pipe reactors  

SciTech Connect

SABRE (Space-Arena Baseline Reactor) is a 100-kW/sub e/, heat-pipe-cooled, beryllium-reflected, fast reactor that produces heat at a temperature of 1500/sup 0/K and radiatively transmits it to high-temperature thermoelectric (TE) conversion elements. The use of heat pipes for core heat removal eliminates single-point failure mechanisms in the reactor cooling system, and provides minimal temperature drop radiative coupling to the TE array, as well as automatic, self-actuating removal of reactor afterheat. The question of how the failure of a fuel module heat pipe will affect neighboring fuel modules in the core is discussed, as is fission density peaking that occurs at the core/reflector interface. Results of neutronic calculations of the control margin available are described. Another issue that is addressed is that of helium generation in the heat pipes from neutron reactions in the core with the heat pipe fluid. Finally, the growth potential of the SABRE design to much higher powers is examined.

Ranken, W.A.; Koenig, D.R.

1983-01-01T23:59:59.000Z

102

THERMAL SIGNATURES OF TETHER-CUTTING RECONNECTIONS IN PRE-ERUPTION CORONAL FLUX ROPES: HOT CENTRAL VOIDS IN CORONAL CAVITIES  

SciTech Connect

Using a three-dimensional MHD simulation, we model the quasi-static evolution and the onset of eruption of a coronal flux rope. The simulation begins with a twisted flux rope emerging at the lower boundary and pushing into a pre-existing coronal potential arcade field. At a chosen time the emergence is stopped with the lower boundary taken to be rigid. Then the coronal flux rope settles into a quasi-static rise phase during which an underlying, central sigmoid-shaped current layer forms along the so-called hyperbolic flux tube (HFT), a generalization of the X-line configuration. Reconnections in the dissipating current layer effectively add twisted flux to the flux rope and thus allow it to rise quasi-statically, even though the magnetic energy is decreasing as the system relaxes. We examine the thermal features produced by the current layer formation and the associated 'tether-cutting' reconnections as a result of heating and field aligned thermal conduction. It is found that a central hot, low-density channel containing reconnected, twisted flux threading under the flux rope axis forms on top of the central current layer. When viewed in the line of sight roughly aligned with the hot channel (which is roughly along the neutral line), the central current layer appears as a high-density vertical column with upward extensions as a {sup U-}shaped dense shell enclosing a central hot, low-density void. Such thermal features have been observed within coronal prominence cavities. Our MHD simulation suggests that they are the signatures of the development of the HFT topology and the associated tether-cutting reconnections, and that the central void grows and rises with the reconnections, until the flux rope reaches the critical height for the onset of the torus instability and dynamic eruption ensues.

Fan, Y. [High Altitude Observatory, National Center for Atmospheric Research, 3080 Center Green Drive, Boulder, CO 80301 (United States)

2012-10-10T23:59:59.000Z

103

COMSOL Simulations for Steady State Thermal Hydraulics Analyses of ORNL s High Flux Isotope Reactor  

SciTech Connect

Simulation models for steady state thermal hydraulics analyses of Oak Ridge National Laboratory s High Flux Isotope Reactor (HFIR) have been developed using the COMSOL Multiphysics simulation software. A single fuel plate and coolant channel of each type of HFIR fuel element was modeled in three dimensions; coupling to adjacent plates and channels was accounted for by using periodic boundary conditions. The standard k- turbulence model was used in simulating turbulent flow with conjugate heat transfer. The COMSOL models were developed to be fully parameterized to allow assessing impacts of fuel fabrication tolerances and uncertainties related to low enriched uranium (LEU) fuel design and reactor operating parameters. Heat source input for the simulations was obtained from separate Monte Carlo N Particle calculations for the axially non-contoured LEU fuel designs at the beginning of the reactor cycle. Mesh refinement studies have been performed to calibrate the models against the pressure drop measured across the HFIR core.

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

2012-01-01T23:59:59.000Z

104

Neutron Science User Program  

E-Print Network (OSTI)

provides a user gateway for SNS and HFIR 11 Managed by UT-Battelle for the U.S. Department of Energy #12.) · Complementary to SNS HFIR produces the world's highest thermal neutron flux #12;13 UT-Battelle Department infrastructure (REDC, HFIR, etc.): $3B+ national asset ORNL is uniquely positioned to support advanced nuclear

105

A Weighted Point Model for the Thermal Neutron Multiplicity Assay of High-Mass Plutonium Samples  

Science Conference Proceedings (OSTI)

A weighted point model for thermal neutron multiplicity counting has been developed for the assay of impure plutonium metal samples. Weighting factors are introduced for the spontaneous fission and ({alpha},n) contributions to the doubles and triples rates to account for the variations in neutron multiplication in these samples. The weighting factors are obtained from Monte Carlo simulations using the MCNPX code, which supports the simulation of spontaneous fission sources and can tally the source and detected neutron multiplicity distributions. Systematic behavior of the weighting factors was studied as a function of sample mass and geometry. Simulations were performed to evaluate the potential accuracy of assays performed with weighted point model analysis. Comparisons with experimental data are presented. The possible use of quads rates is explored.

M.S. Krick; W.H. Geist; D.R. Mayo

2005-10-01T23:59:59.000Z

106

Neutrons  

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

School on Neutron and X-ray Scattering Oak Ridge 10-24 August 2013 John M. Carpenter ANL, ORNLSNS 18 August 2013 2 Neutron Detection How does one detect a neutron? - It is...

107

THERMAL HYDRAULIC ANALYSIS OF A LIQUID-METAL-COOLED NEUTRON SPALLATION TARGET  

Science Conference Proceedings (OSTI)

We have carried out numerical simulations of the thermal hydraulic behavior of a neutron spallation target where liquid metal lead-bismuth serves as both coolant and as a neutron spallation source. The target is one of three designs provided by the Institute of Physics and Power Engineering (IPPE) in Russia. This type of target is proposed for Accelerator-driven Transmutation of Waste (ATW) to eliminate plutonium from hazardous fission products. The thermal hydraulic behavior was simulated by use of a commercial CFD computer code called CFX. Maximum temperatures in the diaphragm window and in the liquid lead were determined. In addition the total pressure drop through the target was predicted. The results of the CFX analysis were close to those results predicted by IPPE in their preliminary analysis.

W. GREGORY; R. MARTIN; T. VALACHOVIC

2000-07-01T23:59:59.000Z

108

TAX: Backscattering Spectrometer at SNS | ORNL Neutron Sciences  

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

The Triple-Axis Spectrometer at HFIR Triple-Axis Spectrometer (HB-3) Triple-Axis Spectrometer (HB-3). HB-3 is a colossal flux thermal neutron three-axis spectrometer designed for...

109

Materials Reliability Program: Codes Comparison for Reactor Internals Neutronics, Thermal, and Mechanical Calculations (MRP-354)  

Science Conference Proceedings (OSTI)

During the last few years, the Electric Power Research Institute (EPRI) has led an effort to conduct functionality analyses, which have been performed by AREVA and Westinghouse, to evaluate the potential degradation of reactor internals during life extension to 60 years. The methodologies employed in these functionality analyses make use of different thermal-hydraulic, neutronics, and structural models and computer codes and involve different geometries and loading histories. A comparative ...

2013-09-18T23:59:59.000Z

110

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

DOE Patents (OSTI)

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.

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

1980-11-05T23:59:59.000Z

111

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

DOE Patents (OSTI)

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.

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

1982-01-01T23:59:59.000Z

112

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

E-Print Network (OSTI)

with neutron tubes that had RTD bundles attached on theequipped with combined RTD and neutron logging devices toyear 2 years 3 years 4 years RTD Location Scale meters Wing

Birkholzer, Jens T.

2005-01-01T23:59:59.000Z

113

Algebraic Turbulent Heat Flux Model for Prediction of Thermal Stratification in Piping Systems  

Science Conference Proceedings (OSTI)

Technical Paper / Special Issue on the 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) / Thermal Hydraulics

M. Pellegrini; H. Endo; E. Merzari; H. Ninokata

114

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

SciTech Connect

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.

Montoya, M. [Instituto Peruano de Energia Nuclear, Av. Canada 1470, Lima 41 (Peru); Facultad de Ciencias, Universidad Nacional de Ingenieria, Av. Tupac Amaru 210, Apartado 31-139, Lima (Peru); Rojas, J. [Instituto Peruano de Energia Nuclear, Av. Canada 1470, Lima 41 (Peru); Facultad de Ciencias Fisicas, Universidad Nacional Mayor de San Marcos, Av. Venezuela Cdra 34, Apartado Postal 14-0149, Lima 1 (Peru); Lobato, I. [Facultad de Ciencias, Universidad Nacional de Ingenieria, Av. Tupac Amaru 210, Apartado 31-139, Lima (Peru)

2010-08-04T23:59:59.000Z

115

New thermal neutron scattering files for ENDF/B-VI release 2  

SciTech Connect

At thermal neutron energies, the binding of the scattering nucleus in a solid, liquid, or gas affects the cross section and the distribution of secondary neutrons. These effects are described in the thermal sub-library of Version VI of the Evaluated Nuclear Data Files (ENDF/B-VI) using the File 7 format. In the original release of the ENDF/B-VI library, the data in File 7 were obtained by converting the thermal scattering evaluations of ENDF/B-III to the ENDF-6 format. These original evaluations were prepared at General Atomics (GA) in the late sixties, and they suffer from accuracy limitations imposed by the computers of the day. This report describes new evaluations for six of the thermal moderator materials and six new cold moderator materials. The calculations were made with the LEAPR module of NJOY, which uses methods based on the British code LEAP, together with the original GA physics models, to obtain new ENDF files that are accurate over a wider range of energy and momentum transfer than the existing files. The new materials are H in H{sub 2}O, Be metal, Be in BeO, C in graphite, H in ZrH, Zr in ZrH, liquid ortho-hydrogen, liquid para-hydrogen, liquid ortho-deuterium, liquid para-deuterium liquid methane, and solid methane.

MacFarlane, R.E.

1994-03-01T23:59:59.000Z

116

Event-by-event study of neutron observables in spontaneous and thermal fission  

Science Conference Proceedings (OSTI)

The event-by-event fission model FREYA is extended to spontaneous fission of actinides and a variety of neutron observables are studied for spontaneous fission and fission induced by thermal neutrons with a view towards possible applications for SNM detection. We have shown that event-by-event models of fission, such as FREYA, provide a powerful tool for studying fission neutron correlations. Our results demonstrate that these correlations are significant and exhibit a dependence on the fissioning nucleus. Since our method is phenomenological in nature, good input data are especially important. Some of the measurements employed in FREYA are rather old and statistics limited. It would be useful to repeat some of these studies with modern detector techniques. In addition, most experiments made to date have not made simultaneous measurements of the fission products and the prompt observables, such as neutron and photons. Such data, while obviously more challenging to obtain, would be valuable for achieving a more complete understanding of the fission process.

Vogt, R; Randrup, J

2011-09-14T23:59:59.000Z

117

Active and passive mode calibration of the Combined Thermal Epithermal Neutron (CTEN) system  

SciTech Connect

The Combined Thermal/Epithermal Neutron (CTEN) non-destructive assay (NDA) system was designed to assay transuranic waste by employing an induced active neutron interrogation and/or a spontaneous passive neutron measurement. This is the second of two papers, and focuses on the passive mode, relating the net double neutron coincidence measurement to the plutonium mass via the calibration constant. National Institute of Standards and Technology (NIST) calibration standards were used and the results verified with NIST-traceable verification standards. Performance demonstration program (PDP) 'empty' 208-L matrix drum was used for the calibration. The experimentally derived calibration constant was found to be 0.0735 {+-} 0.0059 g {sup 240}Pu effective per unit response. Using this calibration constant, the Waste Isolation Pilot Plant (WIPP) criteria was satisfied with five minute waste assays in the range from 3 to 177g Pu. CTEN also participated in the PDP Cycle 8A blind assay with organic sludge and metal matrices and passed the criteria for accuracy and precision in both assay modes. The WIPP and EPA audit was completed March 1, 2002 and full certification is awaiting the closeout of one finding during the audit. With the successful closeout of the audit, the CTEN system will have shown that it can provide very fast assays (five minutes or less) of waste in the range from the minimum detection limit (about 2 mg Pu) to 177 g Pu.

Veilleux, J. M. (John M.)

2002-06-01T23:59:59.000Z

118

Cold Neutron and Ultracold Neutron Sources  

Science Conference Proceedings (OSTI)

... Moderators • Solid Methane – CH 4 – CD 4 ... In a cold neutron flux with a continuous spectrum, more neutrons could ... Magneto-vibrational Scatt. + ...

2009-07-13T23:59:59.000Z

119

Correlation of /sup 239/Pu thermal and fast reactor fission yields with neutron energy  

SciTech Connect

The relative isotopic abundances and the fisson yields for over 40 stable and long-lived fission products from /sup 239/Pu fast fission were evaluated to determine if the data could be correlated with neutron energy. Only mass spectrometric data were used in this study. For some nuclides changes of only a few percent in the relative isotopic abundance or the fission yields over the energy range of thermal to 1 MeV are easily discernable and significant; for others the data are too sparse and scattered to obtain a good correlation. The neutron energy index usedin this study is the /sup 150/Nd//sup 143/Nd isotopic ratio. The results of this correlation study compared to the US Evaluated Nuclear Data File (ENDF) fast fission yield compilation. Several discrepancies are noted and suggestions for future work are presented.

Maeck, W.J.

1981-10-01T23:59:59.000Z

120

Deterministic Multigroup Modeling of Thermal Effect on Neutron Scattering by Heavy Nuclides.  

E-Print Network (OSTI)

??The principal physical phenomenon underlying the computation of neutron spectra is the nuclear reaction in which neutrons lose or gain energy, i.e., the neutron scattering… (more)

Ghrayeb, Shadi

2013-01-01T23:59:59.000Z

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


121

MEASUREMENT OF THE NEUTRON SPECTRUM OF THE HB-4 COLD SOURCE AT THE HIGH FLUX ISOTOPE REACTOR AT OAK RIDGE NATIONAL LABORATORY  

DOE Green Energy (OSTI)

Measurements of the cold neutron spectrum from the super critical hydrogen cold source at the High Flux Isotope Reactor at Oak Ridge National Laboratory were made using time-of-flight spectroscopy. Data were collected at reactor power levels of 8.5MW, 42.5MW and 85MW. The moderator temperature was also varied. Data were collected at 17K and 25K while the reactor power was at 8.5MW, 17K and 25K while at 42.5MW and 18K and 22K while at 85MW. The purpose of these measurements was to characterize the brightness of the cold source and to better understand the relationship between reactor power, moderator temperature, and cold neutron production. The authors will discuss the details of the measurement, the changes observed in the neutron spectrum, and the process for determining the source brightness from the measured neutron intensity.

Robertson, Lee [ORNL; Iverson, Erik B [ORNL

2009-01-01T23:59:59.000Z

122

Directorate Organization | ORNL Neutron Sciences  

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

ORNL Neutron Sciences Directorate The Neutron Sciences Directorate (NScD) manages and operates the Spallation Neutron Source and the High Flux Isotope Reactor, two of the world's...

123

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

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

High Flux Isotope Reactor High Flux Isotope Reactor May 30, 2013 The High Flux Isotope Reactor (HFIR) first achieved criticality on August 25, 1965, and achieved full power in August 1966. It is a versatile 85-MW isotope production, research, and test reactor with the capability and facilities for performing a wide variety of irradiation experiments and a world-class neutron scattering science program. HFIR is a beryllium-reflected, light water-cooled and moderated flux-trap type swimming pool reactor that uses highly enriched uranium-235 as fuel. HFIR typically operates seven 23-to-27 day cycles per year. Irradiation facility capabilities include Flux trap positions: Peak thermal flux of 2.5X1015 n/cm2/s with similar epithermal and fast fluxes (Highest thermal flux available in the

124

Measurement of neutron capture on $^{48}$Ca at thermal and thermonuclear energies  

E-Print Network (OSTI)

At the Karlsruhe pulsed 3.75\\,MV Van de Graaff accelerator the thermonuclear $^{48}$Ca(n,$\\gamma$)$^{49}$Ca(8.72\\,min) cross section was measured by the fast cyclic activation technique via the 3084.5\\,keV $\\gamma$-ray line of the $^{49}$Ca-decay. Samples of CaCO$_3$ enriched in $^{48}$Ca by 77.87\\,\\% were irradiated between two gold foils which served as capture standards. The capture cross-section was measured at the neutron energies 25, 151, 176, and 218\\,keV, respectively. Additionally, the thermal capture cross-section was measured at the reactor BR1 in Mol, Belgium, via the prompt and decay $\\gamma$-ray lines using the same target material. The $^{48}$Ca(n,$\\gamma$)$^{49}$Ca cross-section in the thermonuclear and thermal energy range has been calculated using the direct-capture model combined with folding potentials. The potential strengths are adjusted to the scattering length and the binding energies of the final states in $^{49}$Ca. The small coherent elastic cross section of $^{48}$Ca+n is explained through the nuclear Ramsauer effect. Spectroscopic factors of $^{49}$Ca have been extracted from the thermal capture cross-section with better accuracy than from a recent (d,p) experiment. Within the uncertainties both results are in agreement. The non-resonant thermal and thermonuclear experimental data for this reaction can be reproduced using the direct-capture model. A possible interference with a resonant contribution is discussed. The neutron spectroscopic factors of $^{49}$Ca determined from shell-model calculations are compared with the values extracted from the experimental cross sections for $^{48}$Ca(d,p)$^{49}$Ca and $^{48}$Ca(n,$\\gamma$)$^{49}$Ca.

H. Beer; C. Coceva; P. V. Sedyshev; Yu. P. Popov; H. Herndl; R. Hofinger; P. Mohr; H. Oberhummer

1996-08-07T23:59:59.000Z

125

The Influence of Thermal Pressure on Hypermassive Neutron Star Merger Remnants  

E-Print Network (OSTI)

The merger of two neutron stars leaves behind a rapidly spinning hypermassive object whose survival is believed to depend on the maximum mass supported by the nuclear equation of state, angular momentum redistribution by (magneto-)rotational instabilities, and spindown by gravitational waves. The high temperatures (~5-40 MeV) prevailing in the merger remnant may provide thermal pressure support that could increase its maximum mass and, thus, its life on a neutrino-cooling timescale. We investigate the role of thermal pressure support in hypermassive merger remnants by computing sequences of spherically-symmetric and axisymmetric uniformly and differentially rotating equilibrium solutions to the general-relativistic stellar structure equations. Using a set of finite-temperature nuclear equations of state, we find that hot maximum-mass critically spinning configurations generally do not support larger baryonic masses than their cold counterparts. However, subcritically spinning configurations with mean density of less than a few times nuclear saturation density yield a significantly thermally enhanced mass. Even without decreasing the maximum mass, cooling and other forms of energy loss can drive the remnant to an unstable state. We infer secular instability by identifying approximate energy turning points in equilibrium sequences of constant baryonic mass parametrized by maximum density. Energy loss carries the remnant along the direction of decreasing gravitational mass and higher density until instability triggers collapse. Since configurations with more thermal pressure support are less compact and thus begin their evolution at a lower maximum density, they remain stable for longer periods after merger.

J. D. Kaplan; C. D. Ott; E. P. O'Connor; K. Kiuchi; L. Roberts; M. Duez

2013-06-17T23:59:59.000Z

126

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

SciTech Connect

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.

Ebert, D.

1997-07-01T23:59:59.000Z

127

As-Run Thermal Analysis of the GTL-1 Experiment Irradiated in the ATR South Flux Trap  

Science Conference Proceedings (OSTI)

The GTL-1 experiment was conducted to assess corrosion the performance of the proposed Boosted Fast Flux Loop booster fuel at heat flux levels {approx}30% above the design operating condition. Sixteen miniplates fabricated from 25% enriched, high-density U3Si2/Al dispersion fuel with 6061 aluminum cladding were subjected to peak beginning of cycle (BOC) heat fluxes ranging from 411 W/cm2 to 593 W/cm2. Miniplates fabricated with three different fuel variations (without fines, annealed, and with standard powder) performed equally well, with negligible irradiation-induced swelling and a normal fission density gradient. Both the standard and the modified prefilm procedures produced hydroxide films that adequately protected the miniplates from failure. A detailed finite element model was constructed to calculate temperatures and heat flux for an as-run cycle average effective south lobe power of 25.4 MW(t). Results of the thermal analysis are given at four times during the cycle: BOC at 0 effective full power days (EFPD), middle of cycle (MOC) at 18 EFPD, MOC at 36 EFPD, and end of cycle at 48.9 EFPD. The highest temperatures and heat fluxes occur at the BOC and decrease in a linear manner throughout the cycle. Miniplate heat flux levels and fuel, cladding, hydroxide, and coolant-hydroxide interface temperatures were calculated using the average measured hydroxide thickness on each miniplate. The hydroxide layers are the largest on miniplates nearest to the core midplane, where heat flux and temperature are highest. The hydroxide layer thickness averages 20.4 {mu}m on the six hottest miniplates (B3, B4, C1, C2, C3, and C4). This tends to exacerbate the heating of these miniplates, since a thicker hydroxide layer reduces the heat transfer from the fuel to the coolant. These six hottest miniplates have the following thermal characteristics at BOC: (1) Peak fuel centerline temperature >300 C; (2) Peak cladding temperature >200 C; (3) Peak hydroxide temperature >190 C; (4) Peak hydroxide-water interface temperature >140 C; and (5) Peak heat flux >565 W/cm2.

Donna P. Guillen

2011-05-01T23:59:59.000Z

128

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

Science Conference Proceedings (OSTI)

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.

Ayman Hawari

2008-06-20T23:59:59.000Z

129

Parameterization of Surface Heat Fluxes above Forest with Satellite Thermal Sensing and Boundary-Layer Soundings  

Science Conference Proceedings (OSTI)

Satellite-derived surface temperature measurements can be used in conjunction with temperature and wind soundings in the boundary layer to determine the surface sensible heat flux from forests at the regional scale. The underlying formulation is ...

Wilfried Brutsaert; A. Y. Hsu; Thomas J. Schmugge

1993-05-01T23:59:59.000Z

130

Thermal evaluation of uranium silicide miniplates irradiated at high heat flux  

Science Conference Proceedings (OSTI)

The Gas Test Loop (GTL)-1 irradiation experiment was conducted in the Advanced Test Reactor (ATR) to assess corrosion performance of proposed booster fuel at heat flux levels ~30% above the design operating condition. Sixteen miniplates fabricated from 25% enriched, high-density (4.8 g U/cm3) U3Si2/Al dispersion fuel with 6061 aluminum cladding were subjected to peak beginning of cycle (BOC) heat fluxes ranging from 411 to 593 W/cm2. No adverse impacts to the miniplates were observed at these high heat flux levels. A detailed finite element model was constructed to calculate temperatures and heat flux for an as-run cycle average effective ATR south lobe power of 25.4 MW(t). Miniplate heat flux levels and fuel, cladding, hydroxide, and coolant–hydroxide interface temperatures were calculated using the average hydroxide thickness on each miniplate measured during post-irradiation examination. The purpose of this study was to obtain a best estimate of the as-run experiment temperatures to aid in establishing acceptable heat flux levels and designing fuel qualification experiments for this fuel type.

Donna P. Guillen

2012-09-01T23:59:59.000Z

131

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

SciTech Connect

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.

Montoya, M. [Instituto Peruano de Energia Nuclear, Av. Canada 1470, Lima 41 (Peru); Facultad de Ciencias, Universidad Nacional de Ingenieria, Av. Tupac Amaru 210, Apartado 31-139, Lima (Peru); Rojas, J. [Instituto Peruano de Energia Nuclear, Av. Canada 1470, Lima 41 (Peru); Facultad de Ciencias Fisicas, Universidad Nacional Mayor de San Marcos, Av. Venezuela Cdra 34, Apartado Postal 14-0149, Lima 1 (Peru); Saetone, E. [Facultad de Ciencias, Universidad Nacional de Ingenieria, Av. Tupac Amaru 210, Apartado 31-139, Lima (Peru)

2007-10-26T23:59:59.000Z

132

The Thermal Delocalization of the Flux Tubes in Mesons and Baryons  

Science Conference Proceedings (OSTI)

The gluon action density in a static mesonic system is analyzed at finite temperature using lattice QCD techniques in quenched QCD. The obtained results are compared to predictions of bosonic string models for the flux-tube profiles to understand the changes of the flux-tube profiles with temperature. The mesonic flux tube curved-width profile is found to compare well with that of the bosonic string at large distances. In the intermediate distance region, a free bosonic string behaviour is observed for analysis performed on highly UV-filtered gauge configurations. Extending the analysis to the static baryon reveals a delocalization of the baryonic node in the Y-shape gluonic configuration observed at zero temperature. At finite temperature, a filled delta-shaped configuration is observed, even at large distances. We study a baryonic string model at finite temperature.

Bakry, Ahmed S.; Leinweber, Derek B.; Williams, Anthony G. [Special Research Center for the Subatomic Structure of Matter, Departement of Physics, Adelaide, South Australia 5005 (Australia)

2011-05-24T23:59:59.000Z

133

An evaporation-based model of thermal neutron induced ternary fission of plutonium  

E-Print Network (OSTI)

Ternary fission probabilities for thermal neutron induced fission of plutonium are analyzed within the framework of an evaporation-based model where the complexity of time-varying potentials, associated with the neck collapse, are included in a simplistic fashion. If the nuclear temperature at scission and the fission-neck-collapse time are assumed to be ~1.2 MeV and ~10^-22 s, respectively, then calculated relative probabilities of ternary-fission light-charged-particle emission follow the trends seen in the experimental data. The ability of this model to reproduce ternary fission probabilities spanning seven orders of magnitude for a wide range of light-particle charges and masses implies that ternary fission is caused by the coupling of an evaporation-like process with the rapid re-arrangement of the nuclear fluid following scission.

J. P. Lestone

2007-03-10T23:59:59.000Z

134

Thermal neutron absorption cross sections for igneous rocks: Newberry Caldera, Oregon  

DOE Green Energy (OSTI)

The thermal neutron absorption cross sections of geologic materials are of first-order importance to the interpretation of pulsed neutron porosity logs and of second-order importance to the interpretation of steady-state porosity logs using dual detectors. Even in the latter case, uncertainties in log response can be excessive whenever formations are encountered that possess absorption properties appreciably greater than the limestones used in most tool calibrations. These effects are of importance to logging operations directed at geothermal applications where formation vary from igneous to sedimentary and which may contain solution-deposited minerals with very large cross-section values. Most measurements of cross-section values for geologic materials have been made for hydrocarbon production applications. Hence, the specimen materials are sedimentary and clean in the sense that they are not altered by geothermal fluids. This investigation was undertaken to measure cross-section values from a sequence of igneous materials obtained from a single hole drilled in an active hydrothermal system. 3 refs., 1 fig.

Lysne, P.

1990-01-01T23:59:59.000Z

135

GALLIUM ARSENIDE SEMICONDUCTOR-BASED NEUTRON DETECTOR  

NEUTRON DETECTOR BENEFITS Portable, ... High Flux Isotope Reactor and Spallation Neutron Source. Several Homeland Security. LINKS TO ONLINE ...

136

Reflected Neutron Effects in Multiplicity Measurements of Bare HEU Assemblies  

SciTech Connect

In a passive multiplicity characterization of highly enriched uranium (HEU) assemblies, fission chains are initiated by the characteristically fast neutrons from spontaneous fission of {sup 238}U and {sup 235}U as well as cosmic-ray spallation neutrons. Active interrogation of HEU uses other physical mechanisms for starting chains by inducing fission from high-energy neutrons, high-energy gamma-rays, delayed neutrons, or thermal neutrons. In all cases a contribution to the initiation of fission chains is the reflection of neutrons that initially escape the assembly and re-enter it after undergoing some scattering. The reflected neutron flux is geometry dependent and a combination of fast and thermal energies. The reflected thermal neutron contribution occurs hundreds of microseconds after the beginning of the fission chain and can be distinguished from the cosmic-ray spallation neutrons unrelated to fission chains, resulting in an HEU detection signature with high signal-to-noise. However, the reflected thermal neutron flux can be eliminated with an efficient thermal neutron absorber to investigate reflected neutron effects. In this paper, active and passive multiplicity measurements with HEU oxide assemblies of up to 16 kg of fuel pins and HEU metal assemblies of up to five 18 kg storage castings are reported. Each case demonstrates the differences in HEU signature when a borated thermal neutron absorber is present and shows the various detectable signatures with 3He proportional counters, the standard detector for differential die-way and neutron multiplicity measurements, and liquid scintillators, a detector capable of operating on the timescale of fission chains.

McConchie, Seth M [ORNL; Hausladen, Paul [ORNL; Mihalczo, John T [ORNL

2010-01-01T23:59:59.000Z

137

Neutron source  

DOE Patents (OSTI)

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.

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

1975-10-21T23:59:59.000Z

138

Neutron dosimetry  

DOE Patents (OSTI)

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.

Quinby, Thomas C. (Kingston, TN)

1976-07-27T23:59:59.000Z

139

Construction of Thermal Neutron Calibration Fields Using a Graphite Pile and Americium-Beryllium Neutron Sources at KAERI  

Science Conference Proceedings (OSTI)

Neutron Measurements / Special Issue on the 11th International Conference on Radiation Shielding and the 15th Topical Meeting of the Radiation Protection and Shielding Division (Part 2) / Radiation Protection

B. H. Kim; S. M. Jun; J. S. Kim; K. S. Lim; J. L. Kim

140

Thermal expansion and decomposition of jarosite: a high-temperature neutron diffraction study  

DOE Green Energy (OSTI)

The structure of deuterated jarosite, KFe{sub 3}(SO{sub 4}){sub 2}(OD){sub 6}, was investigated using time-of-flight neutron diffraction up to its dehydroxylation temperature. Rietveld analysis reveals that with increasing temperature, its c dimension expands at a rate {approx}10 times greater than that for a. This anisotropy of thermal expansion is due to rapid increase in the thickness of the (001) sheet of [Fe(O,OH){sub 6}] octahedra and [SO{sub 4}] tetrahedra with increasing temperature. Fitting of the measured cell volumes yields a coefficient of thermal expansion, a = a{sub 0} + a{sub 1} T, where a{sub 0} = 1.01 x 10{sup -4} K{sup -1} and a{sub 1} = -1.15 x 10{sup -7} K{sup -2}. On heating, the hydrogen bonds, O1{hor_ellipsis}D-O3, through which the (001) octahedral-tetrahedral sheets are held together, become weakened, as reflected by an increase in the D{hor_ellipsis}O1 distance and a concomitant decrease in the O3-D distance with increasing temperature. On further heating to 575 K, jarosite starts to decompose into nanocrystalline yavapaiite and hematite (as well as water vapor), a direct result of the breaking of the hydrogen bonds that hold the jarosite structure together.

Xu, Hongwu [Los Alamos National Laboratory; Zhao, Yusheng [Los Alamos National Laboratory; Vogel, Sven C [Los Alamos National Laboratory; Hickmott, Donald D [Los Alamos National Laboratory; Daemen, Luke L [Los Alamos National Laboratory; Hartl, Monika A [Los Alamos National Laboratory

2009-01-01T23:59:59.000Z

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


141

A search for thermally emitting isolated neutron stars in the 2XMMp catalogue  

E-Print Network (OSTI)

The relatively large number of nearby radio-quiet and thermally emitting isolated neutron stars (INSs) discovered in the ROSAT All-Sky Survey, dubbed the ``Magnificent Seven'' (M7), suggests that they belong to a formerly neglected major component of the overall INS population. So far, attempts to discover similar INSs beyond the solar vicinity failed to confirm any reliable candidate. The EPIC cameras onboard the XMM-Newton satellite allow to efficiently search for new thermally emitting INSs. We used the 2XMMp catalogue to select sources with no catalogued candidate counterparts and with X-ray spectra similar to those of the M7, but seen at greater distances and thus undergoing higher interstellar absorptions. Identifications in more than 170 astronomical catalogues and visual screening allowed to select fewer than 30 good INS candidates. In order to rule out alternative identifications, we obtained deep ESO-VLT and SOAR optical imaging for the X-ray brightest candidates. We report here on the optical follo...

Pires, Adriana M; Janot-Pacheco, Eduardo

2009-01-01T23:59:59.000Z

142

Thermal neutron capture cross section of gadolinium by pile-oscillation measurements in MINERVE  

SciTech Connect

Natural gadolinium is used as a burnable poison in most LWR to account for the excess of reactivity of fresh fuels. For an accurate prediction of the cycle length, its nuclear data and especially its neutron capture cross section needs to be known with a high precision. Recent microscopic measurements at Rensselaer Polytechnic Inst. (RPI) suggest a 11% smaller value for the thermal capture cross section of {sup 157}Gd, compared with most of evaluated nuclear data libraries. To solve this inconsistency, we have analyzed several pile-oscillation experiments, performed in the MINERVE reactor. They consist in the measurement of the reactivity variation involved by the introduction in the reactor of small-samples, containing different mass amounts of natural gadolinium. The analysis of these experiments is done through the exact perturbation theory, using the PIMS calculation tool, in order to link the reactivity effect to the thermal capture cross section. The measurement of reactivity effects is used to deduce the 2200 m.s-1 capture cross section of {sup nat}Gd which is (49360 {+-} 790) b. This result is in good agreement with the JEFF3.1.1 value (48630 b), within 1.6% uncertainty at 1{sigma}, but is strongly inconsistent with the microscopic measurements at RPI which give (44200 {+-} 500) b. (authors)

Leconte, P.; Di-Salvo, J.; Antony, M.; Pepino, A. [CEA, DEN, DER, Cadarache, F-13108 Saint-Paul-Lez-Durance (France); Hentati, A. [International School in Nuclear Engineering, Cadarache, F-13108 Saint-Paul-Lez-Durance (France)

2012-07-01T23:59:59.000Z

143

A coupled neutronic/thermal-hydraulic scheme between COBAYA3 and SUBCHANFLOW within the NURESIM simulation platform  

SciTech Connect

Multi-scale, multi-physics problems reveal significant challenges while dealing with coupled neutronic/thermal-hydraulic solutions. Current generation of codes applied to Light Water Reactors (LWR) are based on 3D neutronic nodal methods coupled with one or two phase flow thermal-hydraulic system or sub-channel codes. In addition, spatial meshing and temporal schemes are crucial for the proper description of the non-symmetrical core behavior in case of transient and accidents e.g. reactivity insertion accidents. This paper describes the coupling approach between the 3D neutron diffusion code COBAYA3 and the sub-channel code SUBCHANFLOW within SALOME. The coupling is done inside the SALOME open source platform that is characterized by a powerful pre- and post-processing capabilities and a novel functionality for mapping of the neutronic and thermal hydraulic domains. The peculiar functionalities of SALOME and the steps required for the code integration and coupling are presented. The validation of the coupled codes is done based on two benchmarks the PWR MOX/UO{sub 2} RIA and the TMI-1 MSLB benchmark. A discussion of the prediction capability of COBAYA3/SUBCHANFLOW compared to other coupled solutions will be provided too. (authors)

Calleja, M.; Stieglitz, R.; Sanchez, V.; Jimenez, J.; Imke, U. [Karlsruhe Inst. of Technology KIT, Inst. for Neutron Physics and Reactor Technology INR, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany)

2012-07-01T23:59:59.000Z

144

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

DOE Green Energy (OSTI)

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.

J.T. Birkholzer

2005-02-14T23:59:59.000Z

145

Effects of thermal aging and neutron irradiation on the mechanical properties of three-wire stainless steel weld overlay cladding  

SciTech Connect

Thermal aging of three-wire series-arc stainless steel weld overlay cladding at 288{degrees}C for 1605 h resulted in an appreciable decrease (16%) in the Charpy V-notch (CVN) upper-shelf energy (USE), but the effect on the 41-J transition temperature shift was very small (3{degrees}C). The combined effect of aging and neutron irradiation at 288{degrees}C to a fluence of 5 x 10{sup 19} neutrons/cm{sup 2} (> 1 MeV) was a 22% reduction in the USE and a 29{degrees}C shift in the 41-J transition temperature. The effect of thermal aging on tensile properties was very small. However, the combined effect of irradiation and aging was an increase in the yield strength (6 to 34% at test temperatures from 288 to {minus}125{degrees}C) but no apparent change in ultimate tensile strength or total elongation. Neutron irradiation reduced the initiation fracture toughness (J{sub Ic}) much more than did thermal aging alone. Irradiation slightly decreased the tearing modulus, but no reduction was caused by thermal aging alone. Other results from tensile, CVN, and fracture toughness specimens showed that the effects of thermal aging at 288 or 343{degrees}C for 20,000 h each were very small and similar to those at 288{degrees}C for 1605 h. The effects of long-term thermal exposure time (50,000 h and greater) at 288{degrees}C will be investigated as the specimens become available in 1996 and beyond.

Haggag, F.M.; Nanstad, R.K.

1997-05-01T23:59:59.000Z

146

A Combined Neutronic-Thermal Hydraulic Model of CERMET NTR Reactor  

Science Conference Proceedings (OSTI)

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.

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

2011-02-01T23:59:59.000Z

147

Neutron Science Facilities Operating Status | ORNL Neutron Sciences  

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

Neutron Science Facilities Operating Status High Flux Isotope Reactor The reactor is currently operating at 100% power for fuel cycle 449. Spallation Neutron Source SNS is shutdown...

148

Lattice Calculation of Thermal Properties of Low-Density Neutron Matter with Pionless NN Effective Field Theory  

E-Print Network (OSTI)

Thermal properties of low-density neutron matter are investigated by determinantal quantum Monte Carlo lattice calculations on 3+1 dimensional cubic lattices. Nuclear effective field theory (EFT) is applied using the pionless single- and two-parameter neutron-neutron interactions, determined from the $^1S_0$ scattering length and effective range. The determination of the interactions and the calculations of neutron matter are carried out consistently by applying EFT power counting rules. The thermodynamic limit is taken by the method of finite-size scaling, and the continuum limit is examined in the vanishing lattice filling limit. The $^1S_0$ pairing gap at $T \\approx 0$ is computed directly from the off-diagonal long-range order of the spin pair-pair correlation function, and is found to be approximately 30% smaller than BCS calculations with the conventional nucleon-nucleon potentials. The critical temperature $T_c$ of the normal-to-superfluid phase transition and the pairing temperature scale $T^\\ast$ are determined, and the temperature-density phase diagram is constructed. The physics of low-density neutron matter is clearly identified as being a BCS-Bose-Einstein condensation crossover.

T. Abe; R. Seki

2007-08-19T23:59:59.000Z

149

Mixed field dosimetry of epithermal neutron beams for boron neutron capture therapy at the MITR-II research reactor  

SciTech Connect

During the past several years, there has been growing interest in Boron Neutron Capture Therapy (BNCT) using epithermal neutron beams. The dosimetry of these beams is challenging. The incident beam is comprised mostly of epithermal neutrons, but there is some contamination from photons and fast neutrons. Within the patient, the neutron spectrum changes rapidly as the incident epithermal neutrons scatter and thermalize, and a photon field is generated from neutron capture in hydrogen. In this paper, a method to determine the doses from thermal and fast neutrons, photons, and the B-10([ital n],[alpha])Li-7 reaction is presented. The photon and fast neutron doses are measured with ionization chambers, in realistic phantoms, using the dual chamber technique. The thermal neutron flux is measured with gold foils using the cadmium difference technique; the thermal neutron and B-10 doses are determined by the kerma factor method. Representative results are presented for a unilateral irradiation of the head. Sources of error in the method as applied to BNCT dosimetry, and the uncertainties in the calculated doses are discussed.

Rogus, R.D.; Harling, O.K.; Yanch, J.C. (Massachusetts Institute of Technology, Nuclear Reactor Laboratory, Cambridge, Massachusetts 02139 (United States))

1994-10-01T23:59:59.000Z

150

NON-THERMAL RESPONSE OF THE CORONA TO THE MAGNETIC FLUX DISPERSAL IN THE PHOTOSPHERE OF A DECAYING ACTIVE REGION  

Science Conference Proceedings (OSTI)

We analyzed Solar Dynamics Observatory line-of-sight magnetograms for a decaying NOAA active region (AR) 11451 along with co-temporal Extreme-Ultraviolet Imaging Spectrometer (EIS) data from the Hinode spacecraft. The photosphere was studied via time variations of the turbulent magnetic diffusivity coefficient, {eta}(t), and the magnetic power spectrum index, {alpha}, through analysis of magnetogram data from the Helioseismic and Magnetic Imager (HMI). These measure the intensity of the random motions of magnetic elements and the state of turbulence of the magnetic field, respectively. The time changes of the non-thermal energy release in the corona was explored via histogram analysis of the non-thermal velocity, v {sub nt}, in order to highlight the largest values at each time, which may indicate an increase in energy release in the corona. We used the 10% upper range of the histogram of v {sub nt} (which we called V {sup upp} {sub nt}) of the coronal spectral line of Fe XII 195 A. A 2 day time interval was analyzed from HMI data, along with the EIS data for the same field of view. Our main findings are the following. (1) The magnetic turbulent diffusion coefficient, {eta}(t), precedes the upper range of the v {sub nt} with the time lag of approximately 2 hr and the cross-correlation coefficient of 0.76. (2) The power-law index, {alpha}, of the magnetic power spectrum precedes V {sup upp} {sub nt} with a time lag of approximately 3 hr and the cross-correlation coefficient of 0.5. The data show that the magnetic flux dispersal in the photosphere is relevant to non-thermal energy release dynamics in the above corona. The results are consistent with the nanoflare mechanism of the coronal heating, due to the time lags being consistent with the process of heating and cooling the loops heated by nanoflares.

Harra, L. K. [UCL-Mullard Space Science Laboratory, Holmbury St. Mary, Dorking, Surrey, RH5 6NT (United Kingdom); Abramenko, V. I. [Big Bear Solar Observatory, 40386 N. Shore Lane, Big Bear City, CA 92314 (United States)

2012-11-10T23:59:59.000Z

151

SIGNATURES OF PHOTON-AXION CONVERSION IN THE THERMAL SPECTRA AND POLARIZATION OF NEUTRON STARS  

SciTech Connect

Conversion of photons into axions under the presence of a strong magnetic field can dim the radiation from magnetized astrophysical objects. Here we perform a detailed calculation aimed at quantifying the signatures of photon-axion conversion in the spectra, light curves, and polarization of neutron stars (NSs). We take into account the energy and angle dependence of the conversion probability and the surface thermal emission from NSs. The latter is computed from magnetized atmosphere models that include the effect of photon polarization mode conversion due to vacuum polarization. The resulting spectral models, inclusive of the general-relativistic effects of gravitational redshift and light deflection, allow us to make realistic predictions for the effects of photon to axion conversion on observed NS spectra, light curves, and polarization signals. We identify unique signatures of the conversion, such as an increase of the effective area of a hot spot as it rotates away from the observer line of sight. For a star emitting from the entire surface, the conversion produces apparent radii that are either larger or smaller (depending on axion mass and coupling strength) than the limits set by NS equations of state. For an emission region that is observed phase-on, photon-axion conversion results in an inversion of the plane of polarization with respect to the no-conversion case. While the quantitative details of the features that we identify depend on NS properties (magnetic field strength and temperature) and axion parameters, the spectral and polarization signatures induced by photon-axion conversion are distinctive enough to make NSs very interesting and promising probes of axion physics.

Perna, Rosalba [JILA and Department of Astrophysical and Planetary Science, University of Colorado at Boulder, 440 UCB, Boulder, CO 80304 (United States); Ho, Wynn C. G. [School of Mathematics, University of Southampton, Southampton, SO17 1BJ (United Kingdom); Verde, Licia; Jimenez, Raul [ICREA and ICC, University of Barcelona (IEEC-UB) (Spain); Van Adelsberg, Matthew [Center for Relativistic Astrophysics and School of Physics Georgia Institute of Technology, Atlanta, GA 30332 (United States)

2012-04-01T23:59:59.000Z

152

Modernization of the High Flux Isotope Reactor (HFIR) to Provide a Cold Neutron Source and Experimentation Facility  

Science Conference Proceedings (OSTI)

This paper discusses the installation of a cold neutron source at HFIR with respect to the project as a modernization of the facility. The paper focuses on why the project was required, the scope of the cold source project with specific emphasis on the design, and project management information.

Rothrock, Benjamin G [ORNL; Farrar, Mike B [ORNL

2009-01-01T23:59:59.000Z

153

Computation of neutron fluxes in clusters of fuel pins arranged in hexagonal assemblies (2D and 3D)  

Science Conference Proceedings (OSTI)

For computations of fluxes, we have used Carvik's method of collision probabilities. This method requires tracking algorithms. An algorithm to compute tracks (in 2D and 3D) has been developed for seven hexagonal geometries with cluster of fuel pins. This has been implemented in the NXT module of the code DRAGON. The flux distribution in cluster of pins has been computed by using this code. For testing the results, they are compared when possible with the EXCELT module of the code DRAGON. Tracks are plotted in the NXT module by using MATLAB, these plots are also presented here. Results are presented with increasing number of lines to show the convergence of these results. We have numerically computed volumes, surface areas and the percentage errors in these computations. These results show that 2D results converge faster than 3D results. The accuracy on the computation of fluxes up to second decimal is achieved with fewer lines. (authors)

Prabha, H.; Marleau, G. [Institut de Genie Nucleaire, Ecole Polytechnique de Montreal, Stn. CV, P.O. Box 6079, Montreal, QC H3C 3A7 (Canada)

2012-07-01T23:59:59.000Z

154

Cosmogenic neutron-capture-produced nuclides in stony meteorites  

SciTech Connect

The distribution of neutrons with energies below 15 MeV in spherical stony meteoroids is calculated using the ANISN neutron-transport code. The source distributions and intensities of neutrons are calculated using cross sections for the production of tritium. The meteoroid's radius and chemical composition strongly influence the total neutron flux and the neutron energy spectrum, while the location within a meteoroid only affects the relative neutron intensities. Meteoroids need to have radii of more than 50 g/cm/sup 2/ before they have appreciable fluxes of neutrons near thermal energies. Meteoroids with high hydrogen or low iron contents can thermalize neutrons better than chondrites. Rates for the production of /sup 60/Co, /sup 59/Ni, and /sup 36/Cl are calculated with evaluated neutron-capture cross sections and neutron fluxes determined for carbonaceous chondrites with high hydrogen contents, L-chondrites, and aubrites. For most meteoroids with radii < 300 g/cm/sup 2/, the production rates of these neutron-capture nuclides increase monotonically with depth. The highest calculated /sup 60/Co production rate in an ordinary chondrite is 375 atoms/(min g-Co) at the center of a meteoroid with a 250 g/cm/sup 2/ radius. The production rates calculated for spallogenic /sup 60/Co and /sup 59/Ni are greater than the neutron-capture rates for radii less than approx.50-75 g/cm/sup 2/. Only for very large meteoroids and chlorine-rich samples is the neutron-capture production of /sup 36/Cl important. The results of these calculations are compared with those of previous calculations and with measured activities in many meteorites. 44 refs., 15 figs., 1 tab.

Spergel, M.S.; Reedy, R.C.; Lazareth, O.W.; Levy, P.W.

1985-01-01T23:59:59.000Z

155

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  

Science Conference Proceedings (OSTI)

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.

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)

2010-12-15T23:59:59.000Z

156

Feasibility studies of an accelerator for the intense pulsed neutron source (IPNS)  

SciTech Connect

A proton linac plus synchrotron system was studied for the proposed Intense Pulsed Neutron Source (IPNS) at Argonne. An Alvarez H$sup -$ linac of 70 MeV and a high intensity fast cycling proton synchrotron to accelerate protons to 800 MeV will be the best choice to give a flux of 10$sup 16$ thermal neutron/sec cm$sup 2$ at the surface of moderator with a spallation neutron target of W or $sup 238$U. (auth)

Khoe, T.K.; Kimura, M.

1974-11-01T23:59:59.000Z

157

Literature survey of chemical analysis by thermal neutron induced capture gamma ray spectrometry  

DOE Green Energy (OSTI)

A brief discussion of the principles and techniques of chemical analysis by neutron capture gamma radiation is presented, and the widely scattered literature is collected into a single table arranged by element measured.

Gladney, E.S.

1979-09-01T23:59:59.000Z

158

Boosted Fast Flux Loop Alternative Cooling Assessment  

Science Conference Proceedings (OSTI)

The Gas Test Loop (GTL) Project was instituted to develop the means for conducting fast neutron irradiation tests in a domestic radiation facility. It made use of booster fuel to achieve the high neutron flux, a hafnium thermal neutron absorber to attain the high fast-to-thermal flux ratio, a mixed gas temperature control system for maintaining experiment temperatures, and a compressed gas cooling system to remove heat from the experiment capsules and the hafnium thermal neutron absorber. This GTL system was determined to provide a fast (E > 0.1 MeV) flux greater than 1.0E+15 n/cm2-s with a fast-to-thermal flux ratio in the vicinity of 40. However, the estimated system acquisition cost from earlier studies was deemed to be high. That cost was strongly influenced by the compressed gas cooling system for experiment heat removal. Designers were challenged to find a less expensive way to achieve the required cooling. This report documents the results of the investigation leading to an alternatively cooled configuration, referred to now as the Boosted Fast Flux Loop (BFFL). This configuration relies on a composite material comprised of hafnium aluminide (Al3Hf) in an aluminum matrix to transfer heat from the experiment to pressurized water cooling channels while at the same time providing absorption of thermal neutrons. Investigations into the performance this configuration might achieve showed that it should perform at least as well as its gas-cooled predecessor. Physics calculations indicated that the fast neutron flux averaged over the central 40 cm (16 inches) relative to ATR core mid-plane in irradiation spaces would be about 1.04E+15 n/cm2-s. The fast-to-thermal flux ratio would be in excess of 40. Further, the particular configuration of cooling channels was relatively unimportant compared with the total amount of water in the apparatus in determining performance. Thermal analyses conducted on a candidate configuration showed the design of the water coolant and Al-Hf alloy heat sink system is capable of maintaining all system components below their maximum temperature limits. The maximum temperature of this conduction cooling system, 224.2°C (435.6 °F) occurs in a small, localized region in the heat sink structure near the core mid-plane. The total coolant flow rate requirement for this configuration is 207 L/min (54.7 gpm). The calculated Flow Instability Ratio and Departure from Nucleate Boiling Ratio for this configuration under nominal conditions are 6.5 and 8.0, respectively, which safely exceed the minimum values of 2.0. Materials and fabrication issues inspection revealed that the neutron absorber would probably best be made from powdered Al3Hf mixed with aluminum powder and extruded or hot isostatically pressed. Although Al3Hf has not been specifically studied extensively, its mechanical and chemical properties should be very much like Al3Zr, which has been studied. Its behavior under irradiation should be very satisfactory, and resistance to corrosion will be investigated to a limited extent in planned miniplate irradiation tests in ATR. Pressurized water systems needed to effect heat removal are already available in the ATR complex, and mixed gas temperature control systems needed to trim experiment temperatures have been engineered and need only be fabricated and installed. In sum, it appears the alternately cooled configuration arrived at can be very successful. The cost estimate for this configuration indicates to

Glen R. Longhurst; Donna Post Guillen; James R. Parry; Douglas L. Porter; Bruce W. Wallace

2007-08-01T23:59:59.000Z

159

Wavelength-Shifting-Fiber Scintillation Detectors for Thermal Neutron Imaging at SNS  

Science Conference Proceedings (OSTI)

We have developed wavelength-Shifting-fiber Scintillator Detector (SSD) with 0.3 m2 area per module. Each module has 154 x 7 pixels and a 5 mm x 50 mm pixel size. Our goal is to design a large area neutron detector offering higher detection efficiency and higher count-rate capability for Time-Of-Flight (TOF) neutron diffraction in Spallation Neutron Source (SNS). A ZnS/6LiF scintillator combined with a novel fiber encoding scheme was used to record the neutron events. A channel read-out-card (CROC) based digital-signal processing electronics and position-determination algorithm was applied for neutron imaging. Neutron-gamma discrimination was carried out using pulse-shape discrimination (PSD). A sandwich flat-scintillator detector can have detection efficiency close to He-3 tubes (about 10 atm). A single layer flat-scintillator detector has count rate capability of 6,500 cps/cm2, which is acceptable for powder diffractometers at SNS.

Clonts, Lloyd G [ORNL; Cooper, Ronald G [ORNL; Crow, Lowell [ORNL; Diawara, Yacouba [ORNL; Ellis, E Darren [ORNL; Funk, Loren L [ORNL; Hannan, Bruce W [ORNL; Hodges, Jason P [ORNL; Richards, John D [ORNL; Riedel, Richard A [ORNL; Wang, Cai-Lin [ORNL

2012-01-01T23:59:59.000Z

160

Neutron Detection via the Cherenkov Effect  

SciTech Connect

We have incorporated neutron-absorbing elements in transparent, non-scintillating glasses and used the Cherenkov effect to convert neutron-induced beta-gamma radiation directly into light. Use of the Cherenkov effect requires glasses with a high index of refraction (to lower the threshold and increase the number of Cherenkov photons), and neutron absorbers resulting in radioactive products emitting high-energy beta or gamma radiation. In this paper, we present a brief description of the requirements for developing efficient Cherenkov-based neutron detectors, show the results of measurements of the response of representative samples to a thermal neutron flux, and give the results of a calculation of the expected response of a detector to a moderated fission spectrum.

Bell, Zane W [ORNL; Boatner, Lynn A [ORNL

2008-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "thermal neutron flux" from the National Library of EnergyBeta (NLEBeta).
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they are not comprehensive nor are they the most current set.
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to obtain the most current and comprehensive results.


161

Contribution of nano-scale effects to the total efficiency of converters of thermal neutrons on the basis of gadolinium foils  

E-Print Network (OSTI)

We study the influence of nano-scale layers of converters made from natural gadolinium and its 157 isotope into the total efficiency of registration of thermal neutrons. Our estimations show that contribution of low-energy Auger electrons with the runs about nanometers in gadolinium, to the total efficiency of neutron converters in this case is essential and results in growth of the total efficiency of converters. The received results are in good consent to the experimental data.

D. A. Abdushukurov; D. V. Bondarenko; Kh. Kh. Muminov; D. Yu. Chistyakov

2008-02-04T23:59:59.000Z

162

Thermal effects on the Fission Barrier of neutron-rich nuclei  

E-Print Network (OSTI)

We discuss the fission barrier height of neutron-rich nuclei in a r-process site at highly excited state, which is resulted from the beta-decay or the neutron-capture processes. We particularly investigate the sensitivity of the fission barrier height to the temperature, including the effect of pairing phase transition from superfluid to normal fluid phases. To this end, we use the finite-temperature Skyrme-Hartree-Fock-Bogolubov method with a zero-range pairing interaction. We also discuss the temperature dependence of the fission decay rate.

Futoshi Minato; Kouichi Hagino

2008-07-23T23:59:59.000Z

163

Directional Response of Microstructure Solid State Thermal Neutron Detectors Justin Dingleya  

E-Print Network (OSTI)

A STUDY OF THE SAL NEUTRON DETECTOR EFFICIENCY USING PHOTODISINTEGRATION OF THE DEUTERON A Thesis for the Degree of Master of Science in the Department of Physics and Engineering Physics University of Physics and Engineering Physics University of Saskatchewan Saskatoon, Saskatchewan S7N 0W0 i #12;Abstract

Danon, Yaron

164

Measurement of thermal neutron cross section and resonance integral for the {sup 170}Er(n,{gamma}){sup 171}Er reaction by using a {sup 55}Mn monitor  

Science Conference Proceedings (OSTI)

The thermal neutron cross section and the resonance integral of the reaction {sup 170}Er(n,{gamma}){sup 171}Er were measured by the Cd-ratio method using a {sup 55}Mn monitor as single comparator. Analytical grade MnO{sub 2} and Er{sub 2}O{sub 3} powder samples with and without a cylindrical 1 mm Cd shield box were irradiated in an isotropic neutron field obtained from three {sup 241}Am-Be neutron sources. The induced activities in the samples were measured with a 120.8% relative efficiency p-type HPGe detector. The correction factors for gamma-ray attenuation (F{sub g}), thermal neutron self-shielding (G{sub th}), and resonance neutron self-shielding (G{sub epi}) effects, and the epithermal neutron spectrum shape factor ({alpha}) were taken into account. The thermal neutron cross section for the (n,{gamma}) reaction in {sup 170}Er has been determined to be 8.00 {+-} 0.56 b, relative to that of the {sup 55}Mn monitor. However, some previously reported experimental results compared to the present result show a large discrepancy ranging from 8.3 to 86%. The present result is, in general, in good agreement with the recently measured values by 9%. According to the definition of Cd cut-off energy at 0.55 eV, the resonance integral obtained is 44.5 {+-} 4.0 b, which is determined relative to the reference integral value of the {sup 55}Mn monitor by using cadmium ratios. The existing experimental data for the resonance integral are distributed between 18 and 43 b. The present resonance integral value agrees only with the measurement of 43 {+-} 5 b by Gillette [Thermal Cross Section and Resonance Integral Studies, ORNL-4155, 15 (1967)] within uncertainty limits.

Yuecel, Haluk [Turkish Atomic Energy Authority (TAEK), Besevler Campus, 06100 Tandogan-Ankara (Turkey); Budak, M. Gueray; Karadag, Mustafa [Gazi University, Gazi Education Faculty, 06500 Teknikokullar-Ankara (Turkey)

2007-09-15T23:59:59.000Z

165

Measurement of Neutron Background at the Pyhasalmi mine for CUPP Project, Finland  

E-Print Network (OSTI)

A natural neutron flux is one of significant kind of background in high-sensitive underground experiments. Therefore, when scheduling a delicate underground measurements one needs to measure neutron background. Deep underground the most significant source of neutrons are the U-Th natural radioactive chains giving a fission spectrum with the temperature of 2-3 MeV. Another source is the U-Th alpha-reactions on light nuclei of mine rock giving neutrons with different spectra in the 1-15 MeV energy region. Normal basalt mine rocks contain 1 ppm g/g of U-238 and less. Deep underground those rocks produce natural neutron fluxes of 10^{-7} - 10^{-6} cm^{-2}s^{-1} above 1 MeV. To measure such a background one needs a special techniques. In the Institute for Nuclear Research, Moscow, the neutron spectrometer was developed and built which is sensitive to such a low neutron fluxes. At the end of 2001 the collection of neutron data at the Pyhasalmi mine was started for the CUPP project. During 2002 the background and rough energy spectra of neutron at underground levels 410, 660, 990 and 1410 m were measured. The result of the measurement of the neutron background at different levels of the Pyhasalmi mine is presented and discussed. Data analysis is performed in different energy ranges from thermal neutrons up to 25 MeV and above.

J. N. Abdurashitov; V. N. Gavrin; V. L. Matushko; A. A. Shikhin; V. E. Yants; J. Peltoniemi; T. Keranen

2006-07-20T23:59:59.000Z

166

NEUTRONIC REACTORS  

DOE Patents (OSTI)

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

Wigner, E.P.

1960-11-22T23:59:59.000Z

167

Data Management Practices | ORNL Neutron Sciences  

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

to data generated from neutron scattering experiments at the High Flux Isotope Reactor (HFIR) and the Spallation Neutron Source (SNS). Any changes to these guidelines will be...

168

What Can You Do With Neutrons?  

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

the globe, including the Spallation Neutron Source (SNS) and High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Today the number of active neutron users in...

169

Neutronics and Thermal Hydraulics Study for Using a Low-Enriched Uranium Core in the Advanced Test Reactor -- 2008 Final Report  

SciTech Connect

The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. 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 an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuel cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis was performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff versus effective full power days (EFPDs) between the HEU and the LEU cores. The MCNP ATR 1/8th core model was used to optimize the U 235 loading in the LEU core, such that the differences in K-eff and heat flux profiles between the HEU and LEU cores were minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the ATR reference HEU case study. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm (20 mil). In this work, the proposed LEU (U-10Mo) core conversion case with nominal fuel meat thickness of 0.330 mm (13 mil) and U-235 enrichment of 19.7 wt% is used to optimize the radial heat flux profile by varying the fuel meat thickness from 0.191 mm (7.0 mil) to 0.330 mm (13.0 mil) at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). A 0.8g of Boron-10, a burnable absorber, was added in the inner and outer plates to reduce the initial excess reactivity, and the peak to average ratio of the inner/outer heat flux more effectively. Because the B-10 (n,a) reaction will produce Helium-4 (He-4), which might degrade the LEU foil type fuel performance, an alternative absorber option is proposed. The proposed LEU case study will have 6.918 g of Cadmium (Cd) mixed with the LEU at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19) as a burnable absorber to achieve peak to average ratios similar to those for the ATR reference HEU case study.

G. S. Chang; M. A. Lillo; R. G. Ambrosek

2008-06-01T23:59:59.000Z

170

PHYSICAL REVIEW C 83, 064612 (2011) Advanced Monte Carlo modeling of prompt fission neutrons for thermal and fast neutron-induced  

E-Print Network (OSTI)

PHYSICAL REVIEW C 83, 064612 (2011) Advanced Monte Carlo modeling of prompt fission neutrons of the 6th All Union Conference on Neutron Physics, Kiev, 2­6 October 1983, p. 285, EXFOR entry 40871, A. F. Semenov, and B. I. Starostov, Proceedings of the 6th All Union Conference on Neutron Physics

Danon, Yaron

171

Beam characterization at the Neutron Radiography Reactor  

Science Conference Proceedings (OSTI)

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

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

2013-12-01T23:59:59.000Z

172

HFIR Technical Parameters | ORNL Neutron Sciences  

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

Reactor Technical Parameters Reactor Technical Parameters Overview HFIR Pool Layout HFIR pool layout. HFIR is a beryllium-reflected, light-water-cooled and -moderated, flux-trap type reactor that uses highly enriched uranium-235 as the fuel. The image on the right is a cutaway of the reactor which shows the pressure vessel, its location in the reactor pool, and some of the experiment facilities. The preliminary conceptual design of the reactor was based on the "flux trap" principle, in which the reactor core consists of four annular regions of fuel surrounding an unfueled moderating region or "island" (see cross section view). Such a configuration permits fast neutrons leaking from the fuel to be moderated in the island and thus produces a region of very high thermal-neutron flux at the center of the island. This reservoir of

173

Ultracold Neutron Production in a Pulsed Neutron Beam Line  

E-Print Network (OSTI)

We present the results of an Ultracold neutron (UCN) production experiment in a pulsed neutron beam line at the Los Alamos Neutron Scattering Center. The experimental apparatus allows for a comprehensive set of measurements of UCN production as a function of target temperature, incident neutron energy, target volume, and applied magnetic field. However, the low counting statistics of the UCN signal expected can be overwhelmed by the large background associated with the scattering of the primary cold neutron flux that is required for UCN production. We have developed a background subtraction technique that takes advantage of the very different time-of-flight profiles between the UCN and the cold neutrons, in the pulsed beam. Using the unique timing structure, we can reliably extract the UCN signal. Solid ortho-D$_2$ is used to calibrate UCN transmission through the apparatus, which is designed primarily for studies of UCN production in solid O$_2$. In addition to setting the overall detection efficiency in the apparatus, UCN production data using solid D$_2$ suggest that the UCN upscattering cross-section is smaller than previous estimates, indicating the deficiency of the incoherent approximation widely used to estimate inelastic cross-sections in the thermal and cold regimes.

C. M. Lavelle; W. Fox; G. Manus; P. M. McChesney; D. J. Salvat; Y. Shin; M. Makela; C. Morris; A. Saunders; A. Couture; A. R. Young; C. -Y. Liu

2010-04-15T23:59:59.000Z

174

Improved Calculation of Thermal Fission Energy  

E-Print Network (OSTI)

Thermal fission energy is one of the basic parameters needed in the calculation of antineutrino flux for reactor neutrino experiments. It is useful to improve the precision of the thermal fission energy calculation for current and future reactor neutrino experiments, which are aimed at more precise determination of neutrino oscillation parameters. In this article, we give new values for thermal fission energies of some common thermal reactor fuel iso-topes, with improvements on two aspects. One is more recent input data acquired from updated nuclear databases. The other, which is unprecedented, is a consideration of the production yields of fission fragments from both thermal and fast incident neutrons for each of the four main fuel isotopes. The change in calculated antineutrino flux due to the new values of thermal fission energy is about 0.33%, and the uncertainties of the new values are about 30% smaller.

Ma, X B; Wang, L Z; Chen, Y X; Cao, J

2013-01-01T23:59:59.000Z

175

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

DOE Patents (OSTI)

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)

McCune, D.A.

1964-03-17T23:59:59.000Z

176

Neutronic analysis of a fusion hybrid reactor  

SciTech Connect

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

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

2012-07-01T23:59:59.000Z

177

Thermal-Electromagnetic Analysis of a Fault-Tolerant Dual Star Flux-Switching Permanent Magnet Motor for Critical  

E-Print Network (OSTI)

of a fault-tolerant dual star Flux-Switching Permanent Magnet (FSPM) motor. The analytical results in terms permanent magnet motors have attracted increasing attentions in safety critical applications such as Hybrid, the phases of the motors are magnetically and physically separated. Thus, the phase failures (short

Paris-Sud XI, Université de

178

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

DOE Patents (OSTI)

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)

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

1984-08-16T23:59:59.000Z

179

HFIR | High Flux Isotope Reactor | ORNL  

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

HFIR Working with HFIR Neutron imaging offers new tools for exploring artifacts and ancient technology Home | User Facilities | HFIR HFIR | High Flux Isotope Reactor SHARE The High...

180

Pair breaking and Coulomb effects in cold fission from thermal neutron induced fission of U 233, U 235 and Pu 239  

E-Print Network (OSTI)

In this paper, the distribution of mass and kinetic energy in the cold region of the thermal neutron induced fission of U 233, U 235 and Pu 239, respectively, is interpreted in terms of nucleon pair-breaking and the Coulomb interaction energy between complementary fragments (Coulomb effect). In order to avoid the erosive consequences of neutron emission, one studies the cold fission regions, corresponding to total kinetic energy (TE) close to the maximum available energy of the reaction (Q). Contrary to expected, in cold fission is not observed high odd-even effect in mass number distribution. Nevertheless, the measured values are compatible with higher odd-even effects on proton or neutron number distribution, respectively. In addition, in cold fission, the minimal total excitation energy (X) is correlated with the Coulomb energy excess, which is defined as the difference between C (the electrostatic interaction energy between complementary fragments taken as spherical in scission configuration) and Q. These...

Montoya, Modesto

2013-01-01T23:59:59.000Z

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


181

Dose measurements and calculations in the epithermal neutron beam at the Brookhaven Medical Research Reactor (BMRR)  

SciTech Connect

The characteristics of the epithermal neutron beam at BMRR were measured, calculated, and reported. This beam has already been used for animal irradiations. We anticipate that it will be used for clinical trials. Thermal and epithermal neutron flux densities distributions, and dose rate distributions, as a function of depth were measured in a lucite dog-head phantom. Monte Carlo calculations were performed and compared with the measured values. 2 refs., 4 figs., 1 tab.

Fairchild, R.G.; Greenberg, D.; Kamen, Y.; Fiarman, S. (Brookhaven National Lab., Upton, NY (USA). Medical Dept.); Benary, V. (Brookhaven National Lab., Upton, NY (USA). Medical Dept. Tel Aviv Univ. (Israel)); Kalef-Ezra, J. (Brookhaven National Lab., Upton, NY (USA). Medical Dept. Ioannina Univ. (Greece)); Wielopolski, L. (Brookhaven National Lab., Upton, NY (USA). Medical Dept. State Univ. of New

1990-01-01T23:59:59.000Z

182

High flux compact neutron generators  

E-Print Network (OSTI)

This work was supported by Sandia National Laboratory andThis Work was supported by Sandia National Laboratory and

Reijonen, J.; Lou, T.-P.; Tolmachoff, B.; Leung, K.-N.; Verbeke, J.; Vujic, J.

2001-01-01T23:59:59.000Z

183

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

SciTech Connect

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 code’s 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 project’s primary objective is to advance the state of the art for reactor analysis.

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

2013-11-29T23:59:59.000Z

184

Neutron Characterization for Additive Manufacturing  

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

such as the Spallation Neutron Source (SNS) and the High Flux Isotope Reactor (HFIR) shown in Fig. 1 to solve challenging problems in additive manu- facturing (AM)....

185

Research Highlights | ORNL Neutron Sciences  

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

Polytechnic Institute used small-angle neutron scattering (SANS) at the High-Flux Isotope Reactor at Oak Ridge National Laboratory to identify these early stage aggregates in...

186

Research Highlights | ORNL Neutron Sciences  

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

techniques with neutron scattering at the General-Purpose SANS instrument at the ORNL High Flux Isotope Reactor. The cell mimics were vesicles (hollow spheres) made of...

187

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

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.

Selby, H.D., E-mail: hds@lanl.go [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Mac Innes, M.R.; Barr, D.W.; Keksis, A.L.; Meade, R.A.; Burns, C.J.; Chadwick, M.B.; Wallstrom, T.C. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

2010-12-15T23:59:59.000Z

188

Thermal Gravitational Waves from Primordial Black Holes  

E-Print Network (OSTI)

Thermal gravitational waves can be generated in various sources such as, in the cores of stars, white dwarfs and neutron stars due to the fermion collisions in the dense degenerate Fermi gas. Such high frequency thermal gravitational waves can also be produced during the collisions in a gamma ray burst or during the final stages of the evaporation of primordial black holes. Here we estimate the thermal gravitational waves from primordial black holes and estimate the integrated energy of the gravitational wave emission over the entire volume of the universe and over Hubble time. We also estimate the gravitational wave flux from gamma ray bursts and jets.

C. Sivaram; Kenath Arun

2010-05-19T23:59:59.000Z

189

Neutron flux, spectrum, and dose equivalent measurements for a 4500-W(th) /sup 238/PuO/sub 2/ general purpose heat source  

SciTech Connect

The total emission rate is (4.5 +- 0.4) 10/sup 7/ n/s, and the average neutron energy is (1.64 +- 0.07) MeV. The factor for converting from neutron fluence to dose equivalent for this spectrum is (3.10 +- 0.24) 10/sup -5/ mRem/n-cm/sup -2/. The factor for converting from neutron fluence to tissue absorbed dose is (3.18 +- 0.26) 10/sup -6/ mRad/n-cm/sup -2/.

Anderson, M.E.

1985-05-09T23:59:59.000Z

190

HIGH FLUX ISOTOPE REACTOR PRELIMINARY DESIGN STUDY  

SciTech Connect

A comparison of possible types of research reactors for the production of transplutonium elements and other isotopes indicates that a flux-trap reactor consisting of a beryllium-reflecteds light-water-cooled annular fuel region surrounding a light-water island provides the required thermal neutron fluxes at minimum cost. The preliminary desigu of such a reactor was carried out on the basis of a parametric study of the effect of dimensions of the island and fuel regions heat removal rates, and fuel loading on the achievable thermal neutmn fluxes in the island and reflector. The results indicate that a 12- to 14-cm- diam. island provides the maximum flux for a given power density. This is in good agreement with the US8R critical experiments. Heat removal calculations indicate that average power densities up to 3.9 Mw/liter are achievable with H/ sub 2/O-cooled, platetype fuel elements if the system is pressurized to 650 psi to prevent surface boiling. On this basis, 100 Mw of heat can be removed from a 14-cm-ID x 36-cm-OD x 30.5-cm-long fuel regions resulting in a thermal neutron flux of 3 x 10/sup 15/ in the island after insertion of 100 g of Cm/sup 244/ or equivalent. The resulting production of Cf/sup 252/ amounts to 65 mg for a 1 1/2- year irradiation. Operation of the reactor at the more conservative level of 67 Mw, providing an irradiation flux of 2 x 10/sup 15/ in the islands will result in the production of 35 mg of Cf/sup 252/ per 18 months from 100 g of Cm/sup 244/. A development program is proposed to answer the question of the feasibility of the higher power operation. In addition to the central irradiation facility for heavyelement productions the HFIR contains ten hydraulic rabbit tubes passing through the beryllium reflector for isotope production and four beam holes for basic research, Preliminary estimates indicate that the cost of the facility, designed for an operating power level of 100 Mw, will be approximately 2 million. (auth)

Lane, J.A.; Cheverton, R.D.; Claiborne, G.C.; Cole, T.E.; Gambill, W.R.; Gill, J.P.; Hilvety, N.; McWherther, J.R.; Vroom, D.W.

1959-03-20T23:59:59.000Z

191

Design, construction, and characterization of a facility for neutron capture gamma ray analysis of sulfur in coal using californium-252  

SciTech Connect

A study of neutron capture gamma ray analysis of sulfur in coal using californium-252 as a neutron source is reported. Both internal and external target geometries are investigated. The facility designed for and used in this study is described. The external target geometry is found to be inappropriate because of the low thermal neutron flux at the sample location, which must be outside the biological shielding. The internal target geometry is found to have a sufficient thermal neutron flux, but an excessive gamma ray background. A water filled plastic facility, rather than the paraffin filled steel one used in this study, is suggested as a means of increasing flexibility and decreasing the beackground in the internal target geometry.

Layfield, J.R.

1980-03-01T23:59:59.000Z

192

Pointwise Energy Solution of the Boltzmann Transport Equation for Thermal Neutrons - Final Report - 07/01/1999 - 06/30/2001  

Science Conference Proceedings (OSTI)

In July of 1999 Louisiana State University (LSU) was awarded a two year research grant by the D.O.E. NEER program to develop a methodology for neutron transport calculations using pointwise (PW) nuclear data in the thermal energy range, and to implement the method into the CENTRM transport code being developed at LSU for Oak Ridge National Laboratory (ORNL). This work has extended CENTRM's current epithermal PW calculation to encompass the thermal range, providing a continuous-energy deterministic transport code that can address problems that may not be adequately treated using multigroup methods. The new version of the CENTRM code was completed, and provided to ORNL for inclusion in the next release of the SCALE code system. The new thermal calculation developed by the NEER project is a significant improvement in the CENTRM capability, and should have an impact on criticality and shipping cask analysis done by numerous organizations who use this code system.

Williams, Mark L.

2001-06-30T23:59:59.000Z

193

Neutronic design of a fission converter-based epithermal neutron beam for neutron capture therapy  

SciTech Connect

To meet the needs for neutron capture theory (NCT) irradiations, a high-intensity, high-quality fusion converter-based epithermal neutron beam has been designed for the MITR-II research reactor. This epithermal neutron beam, capable of delivering treatments in a few minutes with negligible background contamination from fast neutrons and photons, will be installed in the present thermal column and hohlraum of the 5-MW MITR-II research reactor. Spent or fresh MITR-II fuel elements will be used to fuel the converter. With a fission converter power of {approximately}80 kW using spent fuel, epithermal fluxes (1 eV < E < 10 keV) in excess of 10{sup 10} n/cm{sup 2} {center_dot} s are achievable at the target position with negligible photon and fast neutron contamination, i.e., <2 {times} 10{sup {minus}11}cGy-cm{sup 2}/n. With the currently available {sup 10}B delivery compound boronophenylalanine-fructose, average therapeutic ratios of {approximately}5 can be achieved using this beam for brain irradiations with deep effective penetration ({approximately}9.5 cm) and high dose rates of up to 400 to 600 RBE cGy/min. If NCT becomes an accepted therapy, fission converter-based beams constructed at existing reactors could meet a large fraction of the projected requirements for intense, low-background epithermal neutron beams at a relatively low cost. The results of an extensive set of neutronic design studies investigating all components of the beam are presented. These detailed studies can be useful as guidance for others who may wish to use the fission converter approach to develop epithermal beams for NCT.

Kiger, W.S. III; Sakamoto, S.; Harling, O.K. [Massachusetts Inst. of Tech., Cambridge, MA (United States)

1999-01-01T23:59:59.000Z

194

Delayed neutron measurements for Th-232, Np-237, Pu-239, Pu-241 and depleted uranium  

E-Print Network (OSTI)

The neutron emission rates from five very pure actinide samples (Th-232, Np-237, Pu-239, Pu-241 and depleted uranium) were measured following equilibrium irradiation in fast and thermal neutron fluxes. The relative abundances (alphas) for the first four groups were calculated from the delayed neutron emission (counts vs. time) data using Keepin's 6-group decay constants (lambdas) for Th-232, Pu-239 and depleted uranium (both fast and thermal neutron induced fissions). The relative abundances (alphas) for the first five groups were calculated for the fast neutron induced fission of Np-237 using the 7-group lambdas obtained by Charlton (1997). The relative abundances for the first five groups were also calculated using the 7-group lambdas proposed by Loaiza and Haskin (2000), the 8-group lambdas proposed by Campbell and Spriggs (1998) and the 8-group lambdas proposed by Piksaikin (2000) for all of the samples (fast neutron induced fission only for Th-232 and Np-237, fast and thermal neutron induced fission for the remainder). Fission product yield and delayed neutron emission probability data from the ENDF-349 and JEF 2.2 nuclear data libraries were also used to simulate neutron emission data from the samples. The calculated neutron yield curves were used to obtain group relative abundances for each of the five actinide samples (fast neutron induced fission only for Th-232 and Np-237, fast and thermal neutron induced fission for the remainder) based on each set of proposed lambdas. The relative abundances obtained from the experiments and calculations are compared and the differences are noted and discussed.

Stone, Joseph C.

2001-01-01T23:59:59.000Z

195

Neutron Scattering Science User ...  

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

Proposals for beam time at Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR) and Spallation Neutron Source (SNS) will be accepted via the web-based proposal system...

196

Using a Borated Panel to Form a Dual Neutron-Gamma Detector  

SciTech Connect

A borated polyethylene plane placed between a neutron source and a gamma spectrometer is used to form a dual neutron-gamma detection system. The polyethylene thermalizes the source neutrons so that they are captured by {sup 10}B to produce a flux of 478 keV gamma-rays that radiate from the plane. This results in a buildup of count rate in the detector over that from a disk of the same diameter as the detector crystal (same thickness as the panel). Radiation portal systems are a potential application of this technique.

Scott Wilde; Raymond Keegan

2008-06-20T23:59:59.000Z

197

ORNL neutron facilities deliver neutrons  

Science Conference Proceedings (OSTI)

The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) resumed full power operations on May 16, 2007. There were three experiment cycles of 23 to 25 days in FY2007 and another six are proposed for FY2008 beginning in November 2007. During FY 2007, the High Flux Isotope Reactor delivered 1178 operating hours to users. Commissioning of two SANS instruments is under way and these instruments will join the user program in 2008. The Neutron Scattering Science Advisory Committee endorsed language encouraging development of the science case for two instruments proposed for HFIR.

Ekkebus, Allen E [ORNL

2008-01-01T23:59:59.000Z

198

A neutronic feasibility study for LEU conversion of the Budapest research reactor.  

SciTech Connect

A neutronic feasibility study for conversion of the Budapest Research Reactor (BRR) from HEU to LEU fuel was performed at Argonne National Laboratory in cooperation with the KFKI Atomic Energy Research Institute in Hungary. Comparisons were made of the reactor performance with the current HEU (36%) fuel and with a proposed LEU (19.75%) fuel. Cycle lengths, thermal neutron fluxes, and rod worths were calculated in equilibrium-type cores for each type of fuel. Relative to the HEU fuel, the LEU fuel has up to a 50% longer fuel cycle length, but a 7-10% smaller thermal neutron flux in the experiment locations. The rod worths are smaller with the LEU fuel, but are still large enough to easily satisfy the BRR shutdown margin criteria. Irradiation testing of four VVR-M2 LEU fuel assemblies that are nearly the same as the proposed BRR LEU fuel assemblies is currently in progress at the Petersburg Nuclear Physics Institute.

Pond, R. B.

1998-10-16T23:59:59.000Z

199

FABRICATION OF NEUTRON SOURCES  

DOE Patents (OSTI)

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.

Birden, J.H.

1959-01-20T23:59:59.000Z

200

Neutronic reactor thermal shield  

DOE Patents (OSTI)

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

Lowe, Paul E. (Blue Ash, OH)

1976-06-15T23:59:59.000Z

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


201

THERMAL NEUTRON MODERNIZATION PROGRAM  

Science Conference Proceedings (OSTI)

... tube liner and shutter that will operate inside the biological shield, and will ... permanent face plate has been designed so that the shielding along the ...

202

Neutron beam characterization at the Neutron Radiography Reactor (NRAD)  

Science Conference Proceedings (OSTI)

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

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

1990-01-01T23:59:59.000Z

203

NEUTRON TRANSPORT THEORY EIGENVALUES NEUTRON FLUX FAST NEUTRONS  

E-Print Network (OSTI)

COMPARATIVE EVALUATIONSToda correspondencia en relaci6n con este trabajo debe dirigirse a1 Servicio de ~ocumentacio'n Biblioteca: y Publicaciones, Junta de ~nergia Nuclear, Ciudad Uni- ':* versitaria, Madrid-3, ESPATA. Las solicitudes de ejemplares deben dirigirse a este mismo Servicio. Los descriptores se han seleccionado del Thesauro del INIS para describir las materias que contiene este informe con vistas a su recuperacio'n. Para m & detalles cog satese el informe IAEA-INIS-12 (INIS: Manual de Indizaci6n) y IAEA-INIS-13 (INIS: Thesauro) publicado por el Organism~ Internacional de Energfa ~t6mica. i Se autoriza la reproduccio'n de 10s resiimenes analiticos que aparecen en esta publicaci6n. Submitted to the 2oth Meeting of the NUCLEAR ENERGY AGENCY COMMITTEE ON REACTOR PHYSICS, Petten (Nederlands), June 1977. Este trkbajo se ha recibido para su impresi6n en Junio de 1977.

Guillermo Velarde; Carolina Ahnert; Jose Maria Aragones; Guillermo Velarde; Carolina Ahnert; Jose Maria Aragones; Thermal Neutrons

1977-01-01T23:59:59.000Z

204

Temperature and heat flux datasets of a complex object in a fire plume for the validation of fire and thermal response codes.  

SciTech Connect

It is necessary to improve understanding and develop temporally- and spatially-resolved integral scale validation data of the heat flux incident to a complex object in addition to measuring the thermal response of said object located within the fire plume for the validation of the SIERRA/FUEGO/SYRINX fire and SIERRA/CALORE codes. To meet this objective, a complex calorimeter with sufficient instrumentation to allow validation of the coupling between FUEGO/SYRINX/CALORE has been designed, fabricated, and tested in the Fire Laboratory for Accreditation of Models and Experiments (FLAME) facility. Validation experiments are specifically designed for direct comparison with the computational predictions. Making meaningful comparison between the computational and experimental results requires careful characterization and control of the experimental features or parameters used as inputs into the computational model. Validation experiments must be designed to capture the essential physical phenomena, including all relevant initial and boundary conditions. This report presents the data validation steps and processes, the results of the penlight radiant heat experiments (for the purpose of validating the CALORE heat transfer modeling of the complex calorimeter), and the results of the fire tests in FLAME.

Jernigan, Dann A.; Blanchat, Thomas K.

2010-09-01T23:59:59.000Z

205

Pair breaking and Coulomb effects in cold fission from thermal neutron induced fission of U 233, U 235 and Pu 239  

E-Print Network (OSTI)

In this paper, the distribution of mass and kinetic energy in the cold region of the thermal neutron induced fission of U 233, U 235 and Pu 239, respectively, is interpreted in terms of nucleon pair-breaking and the Coulomb interaction energy between complementary fragments (Coulomb effect). In order to avoid the erosive consequences of neutron emission, one studies the cold fission regions, corresponding to total kinetic energy (TE) close to the maximum available energy of the reaction (Q). Contrary to expected, in cold fission is not observed high odd-even effect in mass number distribution. Nevertheless, the measured values are compatible with higher odd-even effects on proton or neutron number distribution, respectively. In addition, in cold fission, the minimal total excitation energy (X) is correlated with the Coulomb energy excess, which is defined as the difference between C (the electrostatic interaction energy between complementary fragments taken as spherical in scission configuration) and Q. These Coulomb effects increase with the asymmetry of the charge fragmentations. In sum, the experimental data on cold fission suggest that scission configurations explore all the possibilities permitted by the available energy for fission.

Modesto Montoya

2013-04-30T23:59:59.000Z

206

Method and Apparatus for Detecting Neutrons - Available ...  

Summary. Pacific Northwest National Laboratory has developed a unique detection capability with the ability to identify thermal neutrons, which ...

207

Neutron imaging of alkali metal heat pipes  

Science Conference Proceedings (OSTI)

High-temperature heat pipes are two-phase, capillary driven heat transfer devices capable of passively providing high thermal fluxes. Such a device using a liquid-metal coolant can be used as a solution for successful thermal management on hypersonic flight vehicles. Imaging of the liquid-metal coolant inside will provide valuable information in characterizing the detailed heat and mass transport. Neutron imaging possesses an inherent advantage from the fact that neutrons penetrate the heat pipe metal walls with very little attenuation, but are significantly attenuated by the liquid metal contained inside. Using the BT-2 beam line at the National Institute of Standards and Technology (NIST) in Gaithersburg, Maryland, preliminary efforts have been conducted on a nickel-sodium heat pipe. The contrast between the attenuated beam and the background is calculated to be approximately 3%. This low contrast requires sacrifice in spatial or temporal resolution so efforts have since been concentrated on lithium (Li) which has a substantially larger neutron attenuation cross section. Using the CG-1D beam line at the High Flux Isotope Reactor (HFIR) of Oak Ridge National Laboratory (ORNL) in Oak Ridge, Tennessee, the first neutron images of high-temperature molybdenum (Mo)-Li heat pipes have been achieved. The relatively high neutron cross section of Li allows for the visualization of the Li working fluid inside the heat pipes. The evaporator region of a gravity assisted cylindrical heat pipe prototype 25 cm long was imaged from start-up to steady state operation up to approximately 900 C. In each corner of the square bore inside, the capillary action raises the Li meniscus above the bulk Li pool in the evaporator region. As the operational temperature changes, the meniscus shapes and the bulk meniscus height also changes. Furthermore, a three-dimensional tomographic image is also reconstructed from the total of 128 projection images taken 1.4o apart in which the Li had already cooled and solidified.

Kihm, Ken [University of Tennessee, Knoxville (UTK); Kirchoff, Eric [University of Tennessee, Knoxville (UTK); Golden, Matt [University of Tennessee, Knoxville (UTK); Rosenfeld, J. [Thermacore Inc.; Rawal, S. [Lockheed Martin Space Systems Company; Pratt, D. [United States Air Force Research Laboratory, Wright-Patterson Air Force Base; Bilheux, Hassina Z [ORNL; Walker, Lakeisha MH [ORNL; Voisin, Sophie [ORNL; Hussey, Dan [NIST Center for Neutron Research (NCRN), Gaithersburg, MD

2013-01-01T23:59:59.000Z

208

Ultracold-neutron production in a pulsed-neutron beam line  

Science Conference Proceedings (OSTI)

We present the results of an ultracold neutron (UCN) production experiment in a pulsed-neutron beam line at the Los Alamos Neutron Scattering Center. The experimental apparatus allows for a comprehensive set of measurements of UCN production as a function of target temperature, incident neutron energy, target volume, and applied magnetic field. However, the low counting statistics of the UCN signal can be overwhelmed by the large background associated with the scattering of the primary cold-neutron flux that is required for UCN production. We have developed a background subtraction technique that takes advantage of the very different time-of-flight profiles between the UCN and the cold neutrons, in the pulsed beam. Using the unique timing structure, we can reliably extract the UCN signal. Solid ortho-{sup 2}H{sub 2} is used to calibrate UCN transmission through the apparatus, which is designed primarily for studies of UCN production in solid O{sub 2}. In addition to setting the overall detection efficiency in the apparatus, UCN production data using solid {sup 2}H{sub 2} suggest that the UCN upscattering cross section is smaller than previous estimates, indicating the deficiency of the incoherent approximation widely used to estimate inelastic cross sections in the thermal and cold regimes.

Lavelle, C. M.; Liu, C.-Y.; Fox, W.; Manus, G.; McChesney, P. M.; Salvat, D. J.; Shin, Y.; Makela, M.; Morris, C.; Saunders, A.; Couture, A.; Young, A. R. [Physics Department, Indiana University, Bloomington, Indiana 47408 (United States); Physics Division, P25, Los Alamos National Laboratory, Los Alamos, New Mexico 87544 (United States); LANSCE Division, Nuclear Science, Los Alamos National Laboratory, Los Alamos, New Mexico 87544 (United States); Physics Department, North Carolina State University, Raleigh, North Carolina 27695 (United States)

2010-07-15T23:59:59.000Z

209

Aerial Neutron Detection of Cosmic-Ray Interactions with the Earth's Surface  

SciTech Connect

We have demonstrated the ability to measure the neutron flux produced by the cosmic-ray interaction with nuclei in the ground surface using aerial neutron detection. High energy cosmic-rays (primarily muons with GeV energies) interact with the nuclei in the ground surface and produce energetic neutrons via spallation. At the air-surface interface, the neutrons produced by spallation will either scatter within the surface material, become thermalized and reabsorbed, or be emitted into the air. The mean free path of energetic neutrons in air can be hundreds of feet as opposed to a few feet in dense materials. As such, the flux of neutrons escaping into the air provides a measure of the surface nuclei composition. It has been demonstrated that this effect can be measured at long range using neutron detectors on low flying helicopters. Radiological survey measurements conducted at Government Wash in Las Vegas, Nevada, have shown that the neutron background from the cosmic-soil interactions is repeatable and directly correlated to the geological data. Government Wash has a very unique geology, spanning a wide variety of nuclide mixtures and formations. The results of the preliminary measurements are presented.

Richard Maurer

2008-09-18T23:59:59.000Z

210

Aerial Neutron Detection of Cosmic-Ray Interactions with the Earth's Surface  

SciTech Connect

We have demonstrated the ability to measure the neutron flux produced by the cosmic-ray interaction with nuclei in the ground surface using aerial neutron detection. High energy cosmic-rays (primarily muons with GeV energies) interact with the nuclei in the ground surface and produce energetic neutrons via spallation. At the air-surface interface, the neutrons produced by spallation will either scatter within the surface material, become thermalized and reabsorbed, or be emitted into the air. The mean free path of energetic neutrons in air can be hundreds of feet as opposed to a few feet in dense materials. As such, the flux of neutrons escaping into the air provides a measure of the surface nuclei composition. It has been demonstrated that this effect can be measured at long range using neutron detectors on low flying helicopters. Radiological survey measurements conducted at Government Wash in Las Vegas, Nevada, have shown that the neutron background from the cosmic-soil interactions is repeatable and directly correlated to the geological data. Government Wash has a very unique geology, spanning a wide variety of nuclide mixtures and formations. The results of the preliminary measurements are presented.

Richard Maurer

2008-09-18T23:59:59.000Z

211

Proceedings of the Oak Ridge National Laboratory/Brookhaven National Laboratory workshop on neutron scattering instrumentation at high-flux reactors  

SciTech Connect

For the first three decades following World War II, the US, which pioneered the field of neutron scattering research, enjoyed uncontested leadership in the field. By the mid-1970's, other countries, most notably through the West European consortium at Institut Laue-Langevin (ILL) in Grenoble, France, had begun funding neutron scattering on a scale unmatched in this country. By the early 1980's, observers charged with defining US scientific priorities began to stress the need for upgrading and expansion of US research reactor facilities. The conceptual design of the ANS facility is now well under way, and line-item funding for more advanced design is being sought for FY 1992. This should lead to a construction request in FY 1994 and start-up in FY 1999, assuming an optimal funding profile. While it may be too early to finalize designs for instruments whose construction is nearly a decade removed, it is imperative that we begin to develop the necessary concepts to ensure state-of-the-art instrumentation for the ANS. It is in this context that this Instrumentation Workshop was planned. The workshop touched upon many ideas that must be considered for the ANS, and as anticipated, several of the discussions and findings were relevant to the planning of the HFBR Upgrade. In addition, this report recognizes numerous opportunities for further breakthroughs on neutron instrumentation in areas such as improved detection schemes (including better tailored scintillation materials and image plates, and increased speed in both detection and data handling), in-beam monitors, transmission white beam polarizers, multilayers and supermirrors, and more. Each individual report has been cataloged separately.

McBee, M.R. (ed.); Axe, J.D.; Hayter, J.B.

1990-07-01T23:59:59.000Z

212

Breast Tissue Imaging | ORNL Neutron Sciences  

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

the hydrogen-sensitive neutron imaging capabilities at the High Flux Isotope Reactor (HFIR) to image healthy and cancerous breast tissue specimens. Working with Hassina Bilheux,...

213

Acknowledgement Statement for User Publications | ORNL Neutron...  

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

with this required statement: Part of the Research conducted at ORNL's High Flux Isotope Reactor andor Spallation Neutron Source, as appropriate was sponsored by the...

214

Research Reactors Division | ORNL Neutron Sciences  

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

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

215

Research Reactors Division | Neutron Science | ORNL  

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

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

216

ATRC Neutron Detector Testing Quick Look Report  

SciTech Connect

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.

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

2013-08-01T23:59:59.000Z

217

Gas Flux Sampling | Open Energy Information  

Open Energy Info (EERE)

Gas Flux Sampling Gas Flux Sampling Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Technique: Gas Flux Sampling Details Activities (26) Areas (20) Regions (0) NEPA(0) Exploration Technique Information Exploration Group: Field Techniques Exploration Sub Group: Field Sampling Parent Exploration Technique: Gas Sampling Information Provided by Technique Lithology: Stratigraphic/Structural: High flux can be indicative of conduits for fluid flow. Hydrological: Thermal: Anomalous flux is associated with active hydrothermal activity. Dictionary.png Gas Flux Sampling: Gas flux sampling measures the flow of volatile gas emissions from a specific location and compares it to average background emissions. Anomalously high gas flux can be an indication of hydrothermal activity.

218

Startup of the Fission Converter Epithermal Neutron Irradiation Facility at the MIT Reactor  

Science Conference Proceedings (OSTI)

A new epithermal neutron irradiation facility, based on a fission converter assembly placed in the thermal column outside the reactor core, has been put into operation at the Massachusetts Institute of Technology Research Reactor (MITR). This facility was constructed to provide a high-intensity, forward-directed beam for use in neutron capture therapy with an epithermal flux of [approximately equal to]10{sup 10} n/cm{sup 2}.s at the medical room entrance with negligible fast neutron and gamma-ray contamination. The fission converter assembly consists of 10 or 11 MITR fuel elements placed in an aluminum tank and cooled with D{sub 2}O. Thermal-hydraulic criteria were established based on heat deposition calculations. Various startup tests were performed to verify expected neutronic and thermal-hydraulic behavior. Flow testing showed an almost flat flow distribution across the fuel elements with <5% bypass flow. The total reactivity change caused by operation of the facility was measured at 0.014 {+-} 0.002% {delta}K/K. Thermal power produced by the facility was measured to be 83.1 {+-} 4.2 kW. All of these test results satisfied the thermal-hydraulic safety criteria. In addition, radiation shielding design measurements were made that verified design calculations for the neutronic performance.

Newton, Thomas H. Jr.; Riley, Kent J.; Binns, Peter J.; Kohse, Gordon E.; Hu Linwen; Harling, Otto K. [Massachusetts Institute of Technology (United States)

2002-08-15T23:59:59.000Z

219

Highlights from Research Conducted at ARCS | ORNL Neutron Sciences  

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

New neutron studies support magnetism's role in superconductors Advances in unconventional iron-based superconductors Startling thermal energy behavior revealed by neutron...

220

Neutron capture therapy with deep tissue penetration using capillary neutron focusing  

DOE Patents (OSTI)

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.

Peurrung, Anthony J. (Richland, WA)

1997-01-01T23:59:59.000Z

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


221

Determining Reactor Neutrino Flux  

E-Print Network (OSTI)

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

Cao, Jun

2011-01-01T23:59:59.000Z

222

Determining Reactor Neutrino Flux  

E-Print Network (OSTI)

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

Jun Cao

2011-01-12T23:59:59.000Z

223

Transport-theory-equivalent diffusion coefficients for node-homogenized neutron diffusion problems in CANDU lattices.  

E-Print Network (OSTI)

??Calculation of the neutron flux in a nuclear reactor core is ideally performed by solving the neutron transport equation for a detailed-geometry model using several… (more)

Patel, Amin

2010-01-01T23:59:59.000Z

224

Top neutron scientists named to positions at ORNL | ornl.gov  

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

Department of Energy's Spallation Neutron Source (SNS) and High Flux Isotope Reactor (HFIR), has filled two high-level administrative positions with leaders in the neutron...

225

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

Science Conference Proceedings (OSTI)

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

Ott, Larry J [ORNL; McDuffee, Joel Lee [ORNL

2011-01-01T23:59:59.000Z

226

Radiation Damage Study in Natural Zircon Using Neutrons Irradiation  

Science Conference Proceedings (OSTI)

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.

Lwin, Maung Tin Moe; Amin, Yusoff Mohd.; Kassim, Hasan Abu [Department of Physics, University of Malaya, 50603 Kuala Lumpur (Malaysia); Mohamed, Abdul Aziz [Materials Technology Group, Industrial Technology Division, Malaysian Nuclear Agency Bangi, 43000 Kajang, Selangor Darul Ehsan (Malaysia); Karim, Julia Abdul [Reactor Physics Section, Nuclear Power Division, Malaysian Nuclear Agency Bangi, 43000 Kajang, Selangor Darul Ehsan (Malaysia)

2011-03-30T23:59:59.000Z

227

PRELIMINARY SOLUTION CRITICAL EXPERIMENTS FOR THE HIGH-FLUX ISOTOPE REACTOR  

DOE Green Energy (OSTI)

The design of the High-Flux Isotope Reactor (HFIR) was supported by a series of preliminary experiments performed at the Oak Ridge Critical Experiments Facility in 1960. The experiments yielded results describing directly some of the expected performance characteristics of the reactor and strengthened the calculational methods used in its design. The critical assembly, like the reactor, was of a flux-trap type in which a central 6-in.-dia column of H/sub 2/O was surrounded by an annulus of fissile material and, in turn, by an annular neutron reflector. The fuel region contained a solution of enriched uranyl nitrate in a mixture of H/sub 2/O and D/sub 2/O and the reflector was a composite of two annuli, the inner one of D/sub 2/O surrounded by one of H/sub 2/O. In most experiments the ends of the assembly were reflected by H/sub 2/O. Important results evaluate the absolute thermal-neutron flux to be expected in the design reactor and describe the flux distributions within this type of assembly. It was also observed that the cadmium ratio along the axis of the assembly was about 100, showing that a highly thermal-neutron flux was truly developed in the trap. It was shown that reduction of the hydrogen density in the central water column to about 80% of its normal value increased the reactivity about 6% and that further hydrogen density reduction decreased the reactivity as the effect of the loss of neutron moderation dominated the effect of the increased coupling across the central column. These considerations are of importance to the safety of the reactor. Additional experiments gave values of the usual critical dimensions and explored the effects on both the dimensions and the flux distributions of changing the concentration of the uranyl nitrate solution, of changing the composition of the solvent, and of adding neutron-absorbing materials to the D/ sub 2/O reflector. These changes were made to alter the neutron properties of the fuel solution over a range including those expected in the reactor itself. (auth)

Fox, J.K.; Gilley, L.W.; Magnuson, D.W.

1963-06-12T23:59:59.000Z

228

Improved Calculation of Thermal Fission Energy  

E-Print Network (OSTI)

Thermal fission energy is one of the basic parameters needed in the calculation of antineutrino flux for reactor neutrino experiments. It is useful to improve the precision of the thermal fission energy calculation for current and future reactor neutrino experiments, which are aimed at more precise determination of neutrino oscillation parameters. In this article, we give new values for thermal fission energies of some common thermal reactor fuel isotopes, with improvements on three aspects. One is more recent input data acquired from updated nuclear databases. the second one is a consideration of the production yields of fission fragments from both thermal and fast incident neutrons for each of the four main fuel isotopes. The last one is more carefully calculation of the average energy taken away by antineutrinos in thermal fission with the comparison of antineutrino spectrum from different models. The change in calculated antineutrino flux due to the new values of thermal fission energy is about 0.32%, and the uncertainties of the new values are about 50% smaller.

X. B. Ma; W. L. Zhong; L. Z. Wang; Y. X. Chen; J. Cao

2012-12-29T23:59:59.000Z

229

Research Highlights | ORNL Neutron Sciences  

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

particles of her milk using small-angle neutron scattering at ORNL's High Flux Isotope Reactor (HFIR). Casein micelles, a family of related phosphorus-containing proteins, make up...

230

Optical heat flux gauge  

DOE Patents (OSTI)

A heat flux gauge comprising first and second thermographic phosphor layers separated by a layer of a thermal insulator. 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 figs.

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

1989-06-07T23:59:59.000Z

231

Fluxes at experiment facilities in HEU and LEU designs for the FRM-II.  

SciTech Connect

An Alternative LEU Design for the FRM-II proposed by the RERTR Program at Argonne National Laboratory (ANL) has a compact core consisting of a single fuel element that uses LEU silicide fuel with a uranium density of 4.5 g/cm{sup 3} and has a power level of 32 MW. Both the HEU design by the Technical University of Munich (TUM) and the alternative LEU design by ANL have the same fuel lifetime(50 days) and the same neutron flux performance (8 x 10{sup 14} n/cm{sup 2}-s in the reflector). LEU silicide fuel with 4.5 g/cm{sup 3} has been thoroughly tested and is fully-qualified, licensable, and available now for use in a high flux reactor such as the FRM-II. Several issues that were raised by TUM have been addressed in Refs. 1-3. The conclusions of these analyses are summarized below. This paper addresses four additional issues that have been raised in several forums, including Ref 4: heat generation in the cold neutron source (CNS), the gamma and fast neutron fluxes which are components of the reactor noise in neutron scattering experiments in the experiment hall of the reactor, a fuel cycle length difference, and the reactivity worth of the beam tubes and other experiment facilities. The results show that: (a) for the same thermal neutron flux, the neutron and gamma heating in the CNS is smaller in the LEU design than in the HEU design, and cold neutron fluxes as good or better than those of the HEU design can be obtained with the LEU desin; (b) the gamma and fast neutron components of the reactor noise in the experiment hall are about the same in both designs; (c) the fuel cycle length is 50 days for both designs; and (d) the absolute value of the reactivity worth of the beam tubes and other experiment facilities is smaller in the LEU design, allowing its fuel cycle length to be increased to 53 or 54 days. Based on the excellent results for the Alternative LEU Design that were obtained in all analyses, the RERTR Program reiterates its conclusion that there are no major technical issues regarding use of LEU fuel instead of HEU fuel in the FRM-II and that it is definitely feasible to use LEU fuel in the FRM-II without compromising the safety or performance of the facility.

Hanan, N. A.

1998-01-16T23:59:59.000Z

232

Neutron proton crystallography station (PCS)  

SciTech Connect

The PCS (Protein Crystallography Station) at Los Alamos Neutron Science Center (LANSCE) is a unique facility in the USA that is designed and optimized for detecting and collecting neutron diffraction data from macromolecular crystals. PCS utilizes the 20 Hz spallation neutron source at LANSCE to enable time-of-flight measurements using 0.6-7.0 {angstrom} neutrons. This increases the neutron flux on the sample by using a wavelength range that is optimal for studying macromolecular crystal structures. The diagram below show a schematic of PCS and photos of the detector and instrument cave.

Fisher, Zoe [Los Alamos National Laboratory; Kovalevsky, Andrey [Los Alamos National Laboratory; Johnson, Hannah [Los Alamos National Laboratory; Mustyakimov, Marat [Los Alamos National Laboratory

2009-01-01T23:59:59.000Z

233

Evaluation of the neutron cross sections of /sup 235/U in the thermal energy region. Final report  

SciTech Connect

The objective of this work has been to improve the knowledge of the thermal cross sections of the fissile nuclei as a step toward providing a standard data base for the nuclear industry. The methodology uses a form of the Adler-Adler multilevel-fission theory and Breit-Wigner multilevel-scattering theory. It incorporates these theories in a general nonlinear least-squares (LSQ) fitting program SIGLEARNThe analysis methodology in this work was applied to the thermal data on /sup 235/U. A reference data file has been developed which includes most of the known data of interest. The first important result of this work is the assessment of the shape uncertainties of the partial cross sections. The results of our studies lead to the following values and error estimates for /sup 235/U g factors in a thermal (20.44/sup 0/C) energy spectrum: g/sub f/ = 0.97751 (+-0.11%); g/sub ..gamma../ = 0.98230 (+-0.14%). A second important result of this study is the development of a recommended set of 2200 m/s (0.0253 eV) values of the parameters and the probable range of further adjustment which might be made. The analysis also provides the result of a common interpretation of energy-dependent absolute cross-section data of different measurements to yield a consistent set of experimental 0.0253 eV values with rigorous error estimates. It also provides normalization factors for relative fission and capture cross sections on a common basis with rigorous error estimates. The results of these analyses provide a basis for deciding what new measurements would be most beneficial. The most important of these would be improved direct capture data in the thermal region.

Leonard, B.R. Jr.; Kottwitz, D.A.; Thompson, J.K.

1976-02-01T23:59:59.000Z

234

Activation Measurements for Thermal Neutrons, U.S. Measurements of 36Cl in Mineral Samples from Hiroshima and Nagasaki; and Measurement of 63 Ni in Copper Samples From Hiroshima by Accelerator Mass Spectrometry  

SciTech Connect

The present paper presents the {sup 36}Cl measurement effort in the US. A large number of {sup 36}Cl measurements have been made in both granite and concrete samples obtained from various locations and distances in Hiroshima and Nagasaki. These measurements employed accelerator mass spectrometry (AMS) to quantify the number of atoms of {sup 36}Cl per atom of total Cl in the sample. Results from these measurements are presented here and discussed in the context of the DS02 dosimetry reevaluation effort for Hiroshima and Nagasaki atomic-bomb survivors. The production of {sup 36}Cl by bomb neutrons in mineral samples from Hiroshima and Nagasaki was primarily via the reaction {sup 35}Cl(n,{gamma}){sup 36}Cl. This reaction has a substantial thermal neutron cross-section (43.6 b at 0.025 eV) and the product has a long half-life (301,000 y). hence, it is well suited for neutron-activation detection in Hiroshima and Nagasaki using AMS more than 50 years after the bombings. A less important reaction for bomb neutrons, {sup 39}K(n,{alpha}){sup 36}Cl, typically produces less than 10% of the {sup 36}Cl in mineral samples such as granite and concrete, which contain {approx} 2% potassium. In 1988, only a year after the publication of the DS86 final report (Roesch 1987), it was demonstrated experimentally that {sup 36}Cl measured using AMS should be able to detect the thermal neutron fluences at the large distances most relevant to the A-bomb survivor dosimetry. Subsequent measurements in mineral samples from both Hiroshima and Nagasaki validated the experimental findings. The potential utility of {sup 36}Cl as a thermal neutron detector in Hiroshima was first presented by Haberstock et al. who employed the Munich AMS facility to measure {sup 36}Cl/Cl ratios in a gravestone from near the hypocenter. That work subsequently resulted in an expanded {sup 36}Cl effort in Germany that paralleled the US work. More recently, there have also been {sup 36}Cl measurements made by a Japanese group. The impetus for the extensive {sup 36}Cl and other neutron activation measurements was the recognized need to validate the neutron component of the dose in Hiroshima. Although this was suggested at the time of the DS86 Final Report, where it was stated that the calculated neutron doses for survivors could possibly be wrong, the paucity of neutron validation measurements available at that time prevented adequate resolution of this matter. It was not until additional measurements and data evaluations were made that it became clear that more work was required to better understand the discrepancies observed for thermal neutrons in Hiroshima. This resulted in a large number of additional neutron activation measurements in Hiroshima and Nagasaki by scientists in the US, Japan, and Germany. The results presented here for {sup 36}Cl, together with measurements made by other scientists and for other isotopes, now provide a much improved measurement basis for the validation of neutrons in Hiroshima.

Tore Straume; Alfredo A. Marchetti; Stephen D. Egbert; James A. Roberts; Ping Men; Shoichiro Fujita; Kiyoshi Shizuma; Masaharu Hoshi; G. Rugel; W. Ruhm; G. Korschinek; J. E. McAninch; K. L. Carroll; T. Faestermann; K. Knie; R. E. Martinelli; A. Wallner; C. Wallner

2005-01-14T23:59:59.000Z

235

FABRICATION OF NEUTRON SOURCES  

DOE Patents (OSTI)

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.

Birden, J.H.

1959-04-21T23:59:59.000Z

236

Neutronics studies for a long-wavelength target station at SNS.  

DOE Green Energy (OSTI)

The Spallation Neutron Source (SNS), under construction at Oak Ridge National Laboratory, will be the premier facility for neutron scattering studies in the United States. From the outset the SNS can achieve additional flexibility and accommodate a broader range of scientific investigation than would be possible with only the High Power Target Station by utilizing two target stations, each operating under a separate set of conditions and optimized for a certain class of instruments. A second target station, termed the Long-Wavelength Target Station (LWTS), would operate at a lower pulse rate (e.g., 10 vs. 60 Hz) and utilize very cold moderators to emphasize low-energy (long wavelength) neutrons. The LWTS concept discussed here obtains the highest low-energy fluxes possible for neutron scattering instruments by using a heavy-water-cooled solid tungsten target with two moderators in slab geometry and one in a front wing position. The primary focus has been on solid methane moderators, with liquid methane and hydrogen also considered. We used MCNPX to conduct a series of optimization and sensitivity studies to help determine the optimal neutronic parameters of the LWTS. We compared different options based on the thermal and epithermal fluxes as determined by fitting the spectral intensity of the moderators with a Maxwellian peak and a modified Westcott function. The primary parameters are the moderator positions and composition and the target size. We report results for spectral intensity, pulse shapes, high-energy neutron emission, heating profiles in the target, and target activation.

Micklich, B. J.; Iverson, E. B.; Carpenter, J. M.

2001-09-21T23:59:59.000Z

237

Statistically based uncertainty analysis for ranking of component importance in the thermal-hydraulic safety analysis of the Advanced Neutron Source Reactor  

SciTech Connect

The Analytic Hierarchy Process (AHP) has been used to help determine the importance of components and phenomena in thermal-hydraulic safety analyses of nuclear reactors. The AHP results are based, in part on expert opinion. Therefore, it is prudent to evaluate the uncertainty of the AHP ranks of importance. Prior applications have addressed uncertainty with experimental data comparisons and bounding sensitivity calculations. These methods work well when a sufficient experimental data base exists to justify the comparisons. However, in the case of limited or no experimental data the size of the uncertainty is normally made conservatively large. Accordingly, the author has taken another approach, that of performing a statistically based uncertainty analysis. The new work is based on prior evaluations of the importance of components and phenomena in the thermal-hydraulic safety analysis of the Advanced Neutron Source Reactor (ANSR), a new facility now in the design phase. The uncertainty during large break loss of coolant, and decay heat removal scenarios is estimated by assigning a probability distribution function (pdf) to the potential error in the initial expert estimates of pair-wise importance between the components. Using a Monte Carlo sampling technique, the error pdfs are propagated through the AHP software solutions to determine a pdf of uncertainty in the system wide importance of each component. To enhance the generality of the results, study of one other problem having different number of elements is reported, as are the effects of a larger assumed pdf error in the expert ranks. Validation of the Monte Carlo sample size and repeatability are also documented.

Wilson, G.E.

1992-01-01T23:59:59.000Z

238

The Segregation of Aerosols by Cloud-Nucleating Activity. Part I: Design, Construction, and Testing of A High-Flux Thermal Diffusion Cloud Chamber for Mass Separation  

Science Conference Proceedings (OSTI)

We describe a thermal diffusion cloud chamber operated in series with an aerodynamic dichotomous separator that can segregate aerosol particles by their abilities to nucleate cloud droplets. The apparatus takes advantage of compensating gradients ...

Lee Harrison; Halstead Harrison

1985-04-01T23:59:59.000Z

239

CRTF Real-Time Aperture Flux system  

SciTech Connect

The Real-Time Aperture Flux system (TRAF) is a test measurement system designed to determine the input power/unit area (flux density) during solar experiments conducted at the Central Receiver Test Facility, Sandia National Laboratories, Albuquerque, New Mexico. The RTAF is capable of using both thermal sensors and photon sensors to determine the flux densities in the RTAF measuring plane. These data are manipulated in various ways to derive input power and flux density distribution to solar experiments.

Davis, D.B.

1980-01-01T23:59:59.000Z

240

CONTROL SYSTEM FOR NEUTRONIC REACTORS  

DOE Patents (OSTI)

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)

Crever, F.E.

1962-05-01T23:59:59.000Z

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


241

Conference on New Frontiers in Neutron Macromolecular Crystallography  

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

systems being studied by x-ray diffraction. The advent of the Spallation Neutron Source (SNS) with over an order of magnitude increase in neutron flux, in combination with advances...

242

Comparison of different experimental and analytical measures of the thermal annealing response of neutron-irradiated RPV steels  

Science Conference Proceedings (OSTI)

The thermal annealing response of several materials as indicated by Charpy transition temperature (TT) and upper-shelf energy (USE), crack initiation toughness, K{sub Jc}, predictive models, and automated-ball indentation (ABI) testing are compared. The materials investigated are representative reactor pressure vessel (RPV) steels (several welds and a plate) that were irradiated for other tasks of the Heavy-Section Steel Irradiation (HSSI) Program and are relatively well characterized in the unirradiated and irradiated conditions. They have been annealed at two temperatures, 343 and 454 C (650 and 850 F) for varying lengths of time. The correlation of the Charpy response and the fracture toughness, ABI, and the response predicted by the annealing model of Eason et al. for these conditions and materials appears to be reasonable. The USE after annealing at the temperature of 454 C appears to recover at a faster rate than the TT, and even over-recovers (i.e., the recovered USE exceeds that of the unirradiated material).

Iskander, S.K.; Sokolov, M.A.; Nanstad, R.K. [Oak Ridge National Lab., TN (United States). Metals and Ceramics Div.

1997-05-01T23:59:59.000Z

243

Glossary Term - Neutron Emission  

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

Neutron Previous Term (Neutron) Glossary Main Index Next Term (Niobe) Niobe Neutron Emission After neutron emission, an atom contains one less neutron. Neutron emission is one...

244

Investigating Microscopic Heat Transport with Neutron Scattering  

Science Conference Proceedings (OSTI)

In-Situ Neutron Diffraction and Crystal Plasticity Modeling of a-Uranium · In-Situ Studies of the ... Thermal Residual Stresses and Strains in Depleted Uranium.

245

Continuous Reactors Measure some of the neutrons all of the ...  

Science Conference Proceedings (OSTI)

... reactors use compact cores and highly enriched fuel (over ... U-235) in order to achieve high neutron fluxes ... the use of intermediate enrichment (20-50 ...

2009-11-29T23:59:59.000Z

246

Neutron Diffraction Studies of Intercritically Austempered Ductile Irons  

Science Conference Proceedings (OSTI)

... a function of applied stress were determined using neutron diffraction at the NRSF2 at the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory.

247

Neutron Imaging Explored as Complementary Technique for Improving...  

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

the hydrogen-sensitive neutron imaging capabilities at the High Flux Isotope Reactor (HFIR) to image healthy and cancerous breast tissue specimens. Working with Hassina Bilheux,...

248

Education | ORNL Neutron Sciences  

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

Education banner Education banner Sunil Sinha A Chat with Sunil Sinha, Distinguished Professor of Physics at the University of California-San Diego and speaker at the recent CNMS-SNS Research Forum more... The purpose of the Spallation Neutron Source and the High Flux Isotope Reactor is to facilitate neutron scattering as an integral tool for scientific research and technological development across many scientific and engineering domains within the scientific, academic,and industrial communities. Coupled with this role is a recognized need to inspire, educate, and facilitate the next generation of users and hence foster enhanced use of the unique neutron scattering facilities at ORNL. This is the central theme of the education activities within the Neutron Sciences Directorate (NScD).

249

The advanced neutron source research and development plan  

Science Conference Proceedings (OSTI)

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.

Selby, D.L.

1995-08-01T23:59:59.000Z

250

High Flux Isotope Reactor | ORNL Neutron Sciences  

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

Home Facilities HFIR How to Work with HFIR How to Work with HFIR HFIR Workflow Please contact the experiment interface or coordinator for additional information and...

251

Thermal-hydraulic simulation of natural convection decay heat removal in the High Flux Isotope Reactor using RELAP5 and TEMPEST: Part 1, Models and simulation results  

Science Conference Proceedings (OSTI)

A study was conducted to examine decay heat removal requirements in the High Flux Isotope Reactor (HFIR) following shutdown from 85 MW. The objective of the study was to determine when forced flow through the core could be terminated without causing the fuel to melt. This question is particularly relevant when a station blackout caused by an external event is considered. Analysis of natural circulation in the core, vessel upper plenum, and reactor pool indicates that 12 h of forced flow will permit a safe shutdown with some margin. However, uncertainties in the analysis preclude conclusive proof that 12 h is sufficient. As a result of the study, two seismically qualified diesel generators were installed in HFIR. 9 refs., 4 figs.

Morris, D.G.; Wendel, M.W.; Chen, N.C.J.; Ruggles, A.E.; Cook, D.H.

1989-01-01T23:59:59.000Z

252

Neutron capture therapy with deep tissue penetration using capillary neutron focusing  

DOE Patents (OSTI)

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.

Peurrung, A.J.

1997-08-19T23:59:59.000Z

253

Fission fragment driven neutron source  

DOE Patents (OSTI)

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.

Miller, Lowell G. (Idaho Falls, ID); Young, Robert C. (Idaho Falls, ID); Brugger, Robert M. (Columbia, MO)

1976-01-01T23:59:59.000Z

254

Shift register neutron coincidence module  

SciTech Connect

A neutron coincidence module was designed using multistage shift registers to produce the coincidence gates and a crystal controlled oscillator with variable clock outputs to change the gate lengths. The advantage of this system over the conventional, thermal-neutron coincidence gates is a decrease in deadtime by more than an order of magnitude. (auth)

Stephens, M.M.; Swansen, J.E.; East, L.V.

1975-11-01T23:59:59.000Z

255

Thermal-hydraulic simulation of natural convection decay heat removal in the High Flux Isotope Reactor (HFIR) using RELAP5 and TEMPEST: Part 2, Interpretation and validation of results  

SciTech Connect

The RELAP5/MOD2 code was used to predict the thermal-hydraulic behavior of the HFIR core during decay heat removal through boiling natural circulation. The low system pressure and low mass flux values associated with boiling natural circulation are far from conditions for which RELAP5 is well exercised. Therefore, some simple hand calculations are used herein to establish the physics of the results. The interpretation and validation effort is divided between the time average flow conditions and the time varying flow conditions. The time average flow conditions are evaluated using a lumped parameter model and heat balance. The Martinelli-Nelson correlations are used to model the two-phase pressure drop and void fraction vs flow quality relationship within the core region. Systems of parallel channels are susceptible to both density wave oscillations and pressure drop oscillations. Periodic variations in the mass flux and exit flow quality of individual core channels are predicted by RELAP5. These oscillations are consistent with those observed experimentally and are of the density wave type. The impact of the time varying flow properties on local wall superheat is bounded herein. The conditions necessary for Ledinegg flow excursions are identified. These conditions do not fall within the envelope of decay heat levels relevant to HFIR in boiling natural circulation. 14 refs., 5 figs., 1 tab.

Ruggles, A.E.; Morris, D.G.

1989-01-01T23:59:59.000Z

256

Neutron Radiography  

Science Conference Proceedings (OSTI)

Table 8   Characteristics of neutron radiography at various neutron-energy ranges...Good discrimination between materials and ready availability

257

Neutron Sources  

Science Conference Proceedings (OSTI)

Table 1   Characteristics of neutron radiography at various neutron-energy ranges...Good discrimination between materials, and ready

258

A neutronic feasibility study for LEU conversion of the Brookhaven Medical Research Reactor (BMRR).  

SciTech Connect

A neutronic feasibility study for converting the Brookhaven Medical Research Reactor from HEU to LEU fuel was performed at Argonne National Laboratory in cooperation with Brookhaven National Laboratory. Two possible LEU cores were identified that would provide nearly the same neutron flux and spectrum as the present HEU core at irradiation facilities that are used for Boron Neutron Capture Therapy and for animal research. One core has 17 and the other has 18 LEU MTR-type fuel assemblies with uranium densities of 2.5g U/cm{sup 3} or less in the fuel meat. This LEU fuel is fully-qualified for routine use. Thermal hydraulics and safety analyses need to be performed to complete the feasibility study.

Hanan, N. A.

1998-01-14T23:59:59.000Z

259

High heat flux engineering in solar energy applications  

DOE Green Energy (OSTI)

Solar thermal energy systems can produce heat fluxes in excess of 10,000 kW/m{sup 2}. This paper provides an introduction to the solar concentrators that produce high heat flux, the receivers that convert the flux into usable thermal energy, and the instrumentation systems used to measure flux in the solar environment. References are incorporated to direct the reader to detailed technical information.

Cameron, C.P.

1993-07-01T23:59:59.000Z

260

Theory of cooling neutron stars versus observations  

E-Print Network (OSTI)

We review current state of neutron star cooling theory and discuss the prospects to constrain the equation of state, neutrino emission and superfluid properties of neutron star cores by comparing the cooling theory with observations of thermal radiation from isolated neutron stars.

Yakovlev, D G; Kaminker, A D; Potekhin, A Yu

2007-01-01T23:59:59.000Z

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


261

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

Science Conference Proceedings (OSTI)

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.

Baumann, N.P.; Owais, M.

1980-01-01T23:59:59.000Z

262

Neutronic Characterization of the Megapie Target  

E-Print Network (OSTI)

The MEGAPIE project is one of the key experiments towards the feasibility of Accelerator Driven Systems. On-line operation and post-irradiation analysis will provide the scientific community with unique data on the behavior of a liquid spallation target under realistic irradiation conditions. A good neutronics performance of such a target is of primary importance towards an intense neutron source, where an extended liquid metal loop requires some dedicated verifications related to the delayed neutron activity of the irradiated PbBi. In this paper we report on the experimental characterization of the MEGAPIE neutronics in terms of the prompt neutron (PN) flux inside the target and the delayed neutron (DN) flux on the top of it. For the PN measurements, a complex detector, made of 8 microscopic fission chambers, has been built and installed in the central part of the target to measure the absolute neutron flux and its spatial distribution. Moreover, integral information on the neutron energy distribution as a function of the position along the beam axis could be extracted, providing integral constraints on the neutron production models implemented in transport codes such as MCNPX. For the DN measurement, we used a standard 3He counter and we acquired data during the start-up phase of the target irradiation in order to take sufficient statistics at variable beam power. Experimental results obtained on the PN flux characteristics and their comparison with MCNPX simulations are presented, together with a preliminary analysis of the DN decay time spectrum.

Stefano Panebianco; Olivier Bringer; Pavel Bokov; Sebastien Chabod; Frederic Chartier; Emmeric Dupont; Diane Dore; Xavier Ledoux; Alain Letourneau; Ludovic Oriol; Aurelien Prevost; Danas Ridikas; Jean-Christian Toussaint

2007-10-31T23:59:59.000Z

263

Advanced Neutron Source (ANS) Project progress report, FY 1994  

SciTech Connect

The President`s budget request for FY 1994 included a construction project for the Advanced Neutron Source (ANS). However, the budget that emerged from the Congress did not, and so activities during this reporting period were limited to continued research and development and to advanced conceptual design. A significant effort was devoted to a study, requested by the US Department of Energy (DOE) and led by Brookhaven National Laboratory, of the performance and cost impacts of reducing the uranium fuel enrichment below the baseline design value of 93%. The study also considered alternative core designs that might mitigate those impacts. The ANS Project proposed a modified core design, with three fuel elements instead of two, that would allow operation with only 50% enriched uranium and use existing fuel technology. The performance penalty would be 15--20% loss of thermal neutron flux; the flux would still just meet the minimum design requirement set by the user community. At the time of this writing, DOE has not established an enrichment level for ANS, but two advisory committees have recommended adopting the new core design, provided the minimum flux requirements are still met.

Campbell, J.H.; King-Jones, K.H. [eds.; Selby, D.L.; Harrington, R.M. [Oak Ridge National Lab., TN (United States); Thompson, P.B. [Martin Marietta Energy Systems, Inc., Oak Ridge, TN (United States). Central Engineering Services

1995-01-01T23:59:59.000Z

264

Portable Neutron Sensors for Emergency Response Operations  

SciTech Connect

This article presents the experimental work performed in the area of neutron detector development at the Remote Sensing Laboratory–Andrews 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.

,

2012-06-24T23:59:59.000Z

265

DETERMINATION OF SPECIFIC NEUTRONIC REACTIVITY  

DOE Patents (OSTI)

A method is given for production-line determination of the specific neutronic reactivity of such objects as individual nuclear fuel or neutron absorber elements and is notable for rapidity and apparatus simplicity. The object is incorporated in a slightly sub-critical chain fission reactive assembly having a discrete neutron source, thereby establishing a K/sub eff/ within the crucial range of 0.95 to 0.995. The range was found to afford, uniquely, flux- transient damped response in a niatter of seconds simultaneously with acceptable analytical sensitivity. The resulting neutron flux measured at a situs spaced from both object and source within the assembly serves as a calibrable indication of said reactivity.

Dessauer, G.

1960-05-10T23:59:59.000Z

266

PEBBLE: a two-dimensional steady-state pebble bed reactor thermal hydraulics code  

SciTech Connect

This report documents the local implementation of the PEBBLE code to treat the two-dimensional steady-state pebble bed reactor thermal hydraulics problem. This code is implemented as a module of a computation system used for reactor core history calculations. Given power density data, the geometric description in (RZ), and basic heat removal conditions and thermal properties, the coolant properties, flow conditions, and temperature distributions in the pebble fuel elements are predicted. The calculation is oriented to the continuous fueling, steady state condition with consideration of the effect of the high energy neutron flux exposure and temperature history on the thermal conductivity. The coolant flow conditions are calculated for the same geometry as used in the neutronics calculation, power density and fluence data being used directly, and temperature results are made available for subsequent use.

Vondy, D.R.

1981-09-01T23:59:59.000Z

267

THERMAL HYDRAULICS KEYWORDS: neutron activation,  

E-Print Network (OSTI)

, where the energy generated is determined from measurements of heat balance. The lat- ter includes by standard methods of radiation transport, in particular with Monte Carlo methods. The fluid dynamic part are equivalent regarding their ability to account for the ef- fect of fluid dynamics on the detector time

Pázsit, Imre

268

SHUG Chairman's Message | ORNL Neutron Sciences Users  

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

shall be the Spallation Neutron Source (SNS) and High Flux Isotope Reactor (HFIR) User Group, "SHUG." II. PURPOSE The purpose of the SHUG is to: Provide a formal and...

269

SINGLE CRYSTAL NEUTRON DIFFRACTION.  

SciTech Connect

Single-crystal neutron diffraction measures the elastic Bragg reflection intensities from crystals of a material, the structure of which is the subject of investigation. A single crystal is placed in a beam of neutrons produced at a nuclear reactor or at a proton accelerator-based spallation source. Single-crystal diffraction measurements are commonly made at thermal neutron beam energies, which correspond to neutron wavelengths in the neighborhood of 1 Angstrom. For high-resolution studies requiring shorter wavelengths (ca. 0.3-0.8 Angstroms), a pulsed spallation source or a high-temperature moderator (a ''hot source'') at a reactor may be used. When complex structures with large unit-cell repeats are under investigation, as is the case in structural biology, a cryogenic-temperature moderator (a ''cold source'') may be employed to obtain longer neutron wavelengths (ca. 4-10 Angstroms). A single-crystal neutron diffraction analysis will determine the crystal structure of the material, typically including its unit cell and space group, the positions of the atomic nuclei and their mean-square displacements, and relevant site occupancies. Because the neutron possesses a magnetic moment, the magnetic structure of the material can be determined as well, from the magnetic contribution to the Bragg intensities. This latter aspect falls beyond the scope of the present unit; for information on magnetic scattering of neutrons see Unit 14.3. Instruments for single-crystal diffraction (single-crystal diffractometers or SCDs) are generally available at the major neutron scattering center facilities. Beam time on many of these instruments is available through a proposal mechanism. A listing of neutron SCD instruments and their corresponding facility contacts is included in an appendix accompanying this unit.

KOETZLE,T.F.

2001-03-13T23:59:59.000Z

270

Glossary Term - Neutron  

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

Neutrino Previous Term (Neutrino) Glossary Main Index Next Term (Neutron Emission) Neutron Emission Neutron A Neutron Neutrons are uncharged particles found within atomic nuclei....

271

Neutron Imaging of Archaeological Bronzes  

Science Conference Proceedings (OSTI)

This article presents the initial results of 2-D and 3-D neutron imaging of bronze artifacts using the CG-1D prototype beamline at the High Flux Isotope Reactor (HFIR) located at the Oak Ridge National Laboratory (ORNL). Neutron imaging is a non-destructive technique capable of producing unprecedented three-dimensional information on archaeomaterials, including qualitative, quantitative, and visual data on impurities, composition change, voids, and c

Ryzewski, Krysta [Wayne State University, Detroit; Herringer, Susan [Brown University; Bilheux, Hassina Z [ORNL; Walker, Lakeisha MH [ORNL; Sheldon, Brian [Brown University; Voisin, Sophie [ORNL; Bilheux, Jean-Christophe [ORNL; Finocchiaro, Vincenzo [University of Messina, Messina, Italy

2013-01-01T23:59:59.000Z

272

FLUX SENSOR EVALUATIONS AT THE ATR CRITICAL FACILITY  

Science Conference Proceedings (OSTI)

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 ATRC’s 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.

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

2010-11-01T23:59:59.000Z

273

About Neutrons  

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

Neutron Basics Neutron Basics A neutron is one of the fundamental particles that make up matter. This uncharged particle exists in the nucleus of a typical atom, along with its positively charged counterpart, the proton. Protons and neutrons each have about the same mass, and both can exist as free particles away from the nucleus. In the universe, neutrons are abundant, making up more than half of all visible matter. Find Out What a Neutron Is Youtube icon Properties of Neutrons How Can Neutrons Be Used for Research? Image of glucose movement in plants Neutron imaging techniques have been able to determine the precise movement of glucose in plants. This knowledge can help scientists better understand how biomass can be efficiently converted into fuel. Neutrons have many properties that make them ideal for certain types of

274

The relevance of particle flux monitors in accelerator-based activation analysis  

SciTech Connect

One of the most critical parameters in activation analysis is the flux density of the activating radiation, its spatial distribution in particular. The validity of the basic equation for calculating the activity induced to the exposed item depends upon the fulfilment of several conditions, the most relevant of them being equal doses of incident activating radiation received by the unknown sample, the calibration material and the reference material, respectively. This requirement is most problematic if accelerator-produced radiation is used for activation. Whilst nuclear research reactors usually are equipped with exposure positions that provide fairly homogenous activation fields for thermal neutron activation analysis accelerator-generated particle beams (neutrons, photons, charged particles) usually exhibit axial and, in particular, sharp radial flux gradients. Different experimental procedures have been developed to fulfil the condition mentioned above. In this paper, three variants of the application of flux monitors in photon activation analysis are discussed (external monitor, additive and inherent internal monitor). Experiments have indicated that the latter technique yields highest quality of the analytical results.

Segebade, Chr.; Maimaitimin, M.; Sun Zaijing [Idaho Accelerator Centre, Idaho State University, 1500 Alvin Ricken Drive, Pocatello, ID 83201 (United States)

2013-04-19T23:59:59.000Z

275

Neutron beam imaging at neutron spectrometers at Dhruva  

SciTech Connect

A low efficiency, 2-Dimensional Position Sensitive Neutron Detector based on delay line position encoding is developed. It is designed to handle beam flux of 10{sup 6}-10{sup 7} n/cm{sup 2}/s and for monitoring intensity profiles of neutron beams. The present detector can be mounted in transmission mode, as the hardware allows maximum neutron transmission in sensitive region. Position resolution of 1.2 mm in X and Y directions, is obtained. Online monitoring of beam images and intensity profile of various neutron scattering spectrometers at Dhruva are presented. It shows better dynamic range of intensity over commercial neutron camera and is also time effective over the traditionally used photographic method.

Desai, Shraddha S.; Rao, Mala N. [Solid State Physics Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

2012-06-05T23:59:59.000Z

276

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

DOE Patents (OSTI)

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.

Gold, Raymond (Richland, WA); McElroy, William N. (Richland, WA)

1988-01-01T23:59:59.000Z

277

Variable control of neutron albedo in toroidal fusion devices  

DOE Patents (OSTI)

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.

Jassby, Daniel L. (Princeton, NJ); Micklich, Bradley J. (Princeton, NJ)

1986-01-01T23:59:59.000Z

278

Monte Carlo Modelling of the Electron Spectra of 235U- and 239Pu- Films, Irradiated by Thermal Neutrons, Due to All Possible Mechanisms Excluding b-Decay. Comparison With Experiment  

E-Print Network (OSTI)

The electron energy spectra, not connected to b-decay, of 235U- and 239Pu-films, irradiated by thermal neutrons, obtained by a Monte Carlo method is presented in the given work. The modelling was performed with the help of a computer code MCNP4C (Monte Carlo Neutron Photon transport code system), allowing to carry out the computer experiments on joint transport of neutrons, photons and electrons. The experiment geometry and the parameters of an irradiation were the same, as in [11] (the thickness of a foil varied only). As a result of computer experiments, the electron spectra was obtained for the samples of 235U, 239Pu and uranium dioxide of 93 % enrichment representing a set of films of 22 mm in diameter and different thickness: 0,001 mm, 0,005 mm, 0,02 mm, 0,01 mm, 0,1 mm, 1,0 mm; and also for the uranium dioxide film of 93 % enrichment (diameter 22 mm and thickness 0,01mm), located between two protective 0,025 mm aluminium disks (the conditions of experiment in [11]) and the electron spectrum was fixed at the output surface of a protective disk. The comparative analysis of the experimental [11] and calculated b--spectra is carried out.

V. D. Rusova; V. N. Pavlovychb; V. A. Tarasova; S. V. Iaroshenkob; D. A. Litvinova

2004-07-05T23:59:59.000Z

279

Monte Carlo Modelling of the Electron Spectra of 235U- and 239Pu- Films, Irradiated by Thermal Neutrons, Due to All Possible Mechanisms Excluding b-Decay. Comparison With Experiment  

E-Print Network (OSTI)

The electron energy spectra, not connected to b-decay, of 235U- and 239Pu-films, irradiated by thermal neutrons, obtained by a Monte Carlo method is presented in the given work. The modelling was performed with the help of a computer code MCNP4C (Monte Carlo Neutron Photon transport code system), allowing to carry out the computer experiments on joint transport of neutrons, photons and electrons. The experiment geometry and the parameters of an irradiation were the same, as in [11] (the thickness of a foil varied only). As a result of computer experiments, the electron spectra was obtained for the samples of 235U, 239Pu and uranium dioxide of 93 % enrichment representing a set of films of 22 mm in diameter and different thickness: 0,001 mm, 0,005 mm, 0,02 mm, 0,01 mm, 0,1 mm, 1,0 mm; and also for the uranium dioxide film of 93 % enrichment (diameter 22 mm and thickness 0,01mm), located between two protective 0,025 mm aluminium disks (the conditions of experiment in [11]) and the electron spectrum was fixed at...

Rusova, V D; Tarasova, V A; Iaroshenkob, S V; Litvinova, D A

2004-01-01T23:59:59.000Z

280

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  

Science Conference Proceedings (OSTI)

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.

Pritychenko, B. [National Nuclear Data Center, Brookhaven National Laboratory, Upton, NY 11973-5000 (United States)] [National Nuclear Data Center, Brookhaven National Laboratory, Upton, NY 11973-5000 (United States); Mughabghab, S.F. [National Nuclear Data Center, Brookhaven National Laboratory, Upton, NY 11973-5000 (United States)] [National Nuclear Data Center, Brookhaven National Laboratory, Upton, NY 11973-5000 (United States)

2012-12-15T23:59:59.000Z

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


281

ENERGY DISTRIBUTION OF FAST NEUTRON BEAM  

DOE Green Energy (OSTI)

Experimental techniques are described for the spectral measurement of a collimated fast-neutron beam. A H/sub 2-/ filled cloud chamber, proton-recording nuclear plates, and threshold fission foils were used as neutron detectors in the measurements. As an application of these techniques, the energy distribution and absolute flux of the fast neutron beam emerging from the Los Alamos fast reactor was measured from 0.1 to 18 Mev. (D.E.B.)

Nereson, N.; Allison, E.; Carlson, J.; Norwood, P.; Squires, D.

1951-02-15T23:59:59.000Z

282

National Solar Thermal Test Facility  

SciTech Connect

This is a brief report about a Sandia National Laboratory facility which can provide high-thermal flux for simulation of nuclear thermal flash, measurements of the effects of aerodynamic heating on radar transmission, etc

Cameron, C.P.

1989-12-31T23:59:59.000Z

283

Post irradiation experiment analysis using the APOLLO2 deterministic tool. Validation of JEFF-3.1.1 thermal and epithermal actinides neutron induced cross sections through MELUSINE experiments  

Science Conference Proceedings (OSTI)

Two different experiments performed in the 8 MWth MELUSINE experimental power pool reactor aimed at analyzing 1 GWd/t spent fuel pellets doped with several actinides. The goal was to measure the averaged neutron induced capture cross section in two very different neutron spectra (a PWR-like and an under-moderated one). This paper summarizes the combined deterministic APOLLO2-stochastic TRIPOLI4 analysis using the JEFF-3.1.1 European nuclear data library. A very good agreement is observed for most of neutron induced capture cross section of actinides and a clear underestimation for the {sup 241}Am(n,{gamma}) as an accurate validation of its associated isomeric ratio are emphasized. Finally, a possible huge resonant fluctuation (factor of 2.7 regarding to the 1=0 resonance total orbital momenta) is suggested for isomeric ratio. (authors)

Bernard, D.; Fabbris, O. [CEA, DEN, SPRC, Laboratoire d'Etudes de Physique, F-13108 Saint Paul Lez Durance (France)

2012-07-01T23:59:59.000Z

284

The Upscattering of Ultracold Neutrons from the polymer $[C_6 H_{12}]_n$  

E-Print Network (OSTI)

It is generally accepted that the main cause of ultracold neutron (UCN) losses in storage traps is the upscattering to the thermal energy range by hydrogen adsorbed on the surface of the trap walls. However, the data on which this conclusion is based are poor and contradictory. Here, we report a measurement, performed at the Los Alamos National Laboratory UCN source, of the average energy of the flux of upscattered neutrons after the interaction of UCN with hydrogen bound in semicrystalline polymer PMP (tradename TPX), [C$_{6}$H$_{12}$]$_n$. Our analysis, performed with the MCNP code based on the application of the neutron scattering law to UCN upscattered by bound hydrogen in semicrystalline polyethylene, [C$_{2}$H$_{4}$]$_n$, leads us to a flux average energy value of 26$\\pm3$ meV in contradiction with previously reported experimental values of 10 to 13 meV and in agreement with the theoretical models of neutron heating implemented in the MCNP code.

Sharapov, E I; Makela, M; Saunders, A; Adamek, Evan R; Broussard, L J; Cude-Woods, C B; Fellers, Deion E; Geltenbort, Peter; Hartl, M; Hasan, S I; Hickerson, K P; Hogan, G; Holley, A T; Lavelle, C M; Liu, Chen-Yu; Mendenhall, M P; Ortiz, J; Pattie, R W; Ramsey, J; Salvat, D J; Seestrom, S J; Shaw, E; Sjue, Sky; Sondheim, W E; VornDick, B; Wang, Z; Womack, T L; Young, A R; Zeck, Bryan A; Phillips, D G

2013-01-01T23:59:59.000Z

285

The Upscattering of Ultracold Neutrons from the polymer $[C_6 H_{12}]_n$  

E-Print Network (OSTI)

It is generally accepted that the main cause of ultracold neutron (UCN) losses in storage traps is the upscattering to the thermal energy range by hydrogen adsorbed on the surface of the trap walls. However, the data on which this conclusion is based are poor and contradictory. Here, we report a measurement, performed at the Los Alamos National Laboratory UCN source, of the average energy of the flux of upscattered neutrons after the interaction of UCN with hydrogen bound in semicrystalline polymer PMP (tradename TPX), [C$_{6}$H$_{12}$]$_n$. Our analysis, performed with the MCNP code based on the application of the neutron scattering law to UCN upscattered by bound hydrogen in semicrystalline polyethylene, [C$_{2}$H$_{4}$]$_n$, leads us to a flux average energy value of 26$\\pm3$ meV in contradiction with previously reported experimental values of 10 to 13 meV and in agreement with the theoretical models of neutron heating implemented in the MCNP code.

E. I. Sharapov; C. L. Morris; M. Makela; A. Saunders; Evan R. Adamek; L. J. Broussard; C. B. Cude-Woods; Deion E Fellers; Peter Geltenbort; M. Hartl; S. I. Hasan; K. P. Hickerson; G. Hogan; A. T. Holley; C. M. Lavelle; Chen-Yu Liu; M. P. Mendenhall; J. Ortiz; R. W. Pattie Jr.; J. Ramsey; D. J. Salvat; S. J. Seestrom; E. Shaw; Sky Sjue; W. E. Sondheim; B. VornDick; Z. Wang; T. L. Womack; A. R. Young; Bryan A. Zeck; D. G. Phillips

2013-08-12T23:59:59.000Z

286

SCIENCE HIGHLIGHTS 2008 ANNUAL REPORT ORNL NEUTRON SCIENCES neutrons.ornl.gov  

E-Print Network (OSTI)

of the campus, High Flux Isotope Reactor (HFIR), Conference Center and short walk to the Spallation Neutron nearby Reservations can be made 24/7 by calling 865-576-8101 Map of ORNL Campus #12;Maps of SNS, HFIR

287

Evaluation of the thermal-neutron constants for /sup 233/U, /sup 235/U, /sup 239/Pu and /sup 241/Pu  

Science Conference Proceedings (OSTI)

A consistent set of best values of the 2200 meter/second neutron cross sections, Westcott g-factors, and fission neutron yields for /sup 233/U, /sup 235/U, /sup 239/Pu and /sup 241/Pu are presented. A least squares fitting program, LSF, is used to obtain the best fit and to estimate the sensitivity of these fissile parameters to the quoted uncertainties in experimental data. The half-lives of the uranium and plutonium nuclides have been evaluated and these have been used to reassess the significant experimental data. The latest revision of the spontaneous fission neutron yield anti nu, of /sup 252/Cf and the foil thickness corrections to the fission neutron yield ratios of fissile nuclei to /sup 252/Cf are included. These lead to greater consistency in the data used for anti nu (/sup 252/Cf). Similarly, the /sup 234/U half-life as revised leads to improved consistency in the /sup 235/U fission cross section. Comparison is made with the values from ENDF/B-V and other evaluations.

Stehn, J.R.; Divadeenam, M.; Holden, N.E.

1982-01-01T23:59:59.000Z

288

Micro/Nano-Scale Phase Change Systems for Thermal Management and Solar Energy Conversion Applications  

E-Print Network (OSTI)

Heat Exchangers,” Applied Thermal Engineering, 25 (1), pp.Raad P. E. , 2008, “Thermal Challenges in Next-GenerationAssessment of High-Heat-Flux Thermal Management Schemes,”

Coso, Dusan

2013-01-01T23:59:59.000Z

289

Research Highlights | ORNL Neutron Sciences  

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

phase behavior in carbon pores phase behavior in carbon pores Neutrons measure phase behavior in pores at angstrom size Compelling new methods for assessing carbon pores for hydrogen storage in fuel cells Research Contact: Yuri Melnichenko Jan. 2012, Written by Agatha Bardoel Yuri Melnichenko and Lilin He GP-SANS instrument scientist Yuri Melnichenko (left) and postdoctoral associate Lilin He. Researchers have measured the phase behavior of green house gases in pores at the angstrom level, using small-angle neutron scattering (SANS) at the Oak Ridge National Laboratory's High Flux Isotope Reactor. Yuri Melnichenko, an instrument scientist on the General-Purpose Small-Angle Neutron Scattering (GP-SANS) Diffractometer at ORNL's High Flux Isotope Reactor, his postdoctoral associate Lilin He and collaborators

290

Fast flux locked loop  

DOE Patents (OSTI)

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

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

2002-09-10T23:59:59.000Z

291

Contacts | ORNL Neutron Sciences  

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

Science Points of Contact Science Points of Contact Name Research Area Doug Abernathy Wide Angular-Range Chopper Spectrometer (ARCS). Atomic-scale dynamics at thermal and epithermal energies Ke An Engineering Materials Diffractometer (VULCAN). Residual stress, deformation mechanism of materials, phase transitions/transformation, and in situ/operando neutron diffraction in material systems (e.g., working batteries). John Ankner Liquids Reflectometer (LR). Density profiles normal to the surface at liquid surfaces and liquid interfaces Bryan Chakoumakos Nuclear and magnetic crystal structure systematics and structure-property relationships among inorganic materials, powder and single-crystal neutron and x-ray diffraction methods Leighton Coates Macromolecular Neutron Diffractometer (MaNDi). Protein crystallography, biological structure and function

292

Optimized Structures for Low-Profile Phase Change Thermal Spreaders.  

E-Print Network (OSTI)

??Thin, low-profile phase change thermal spreaders can provide cooling solutions for some of today's most pressing heat flux dissipation issues. These thermal issues are only… (more)

Sharratt, Stephen A.

2012-01-01T23:59:59.000Z

293

X-ray and Neutron Studies of Fluids in Confinement  

Science Conference Proceedings (OSTI)

In-Situ Neutron Diffraction and Crystal Plasticity Modeling of a-Uranium · In-Situ Studies of the ... Thermal Residual Stresses and Strains in Depleted Uranium.

294

What Neutrons Tell Us about Magnetic Shape Memory Materials?  

Science Conference Proceedings (OSTI)

In-Situ Neutron Diffraction and Crystal Plasticity Modeling of a-Uranium · In-Situ Studies of the ... Thermal Residual Stresses and Strains in Depleted Uranium.

295

Neutron Log | Open Energy Information  

Open Energy Info (EERE)

Neutron Log Neutron Log Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Technique: Neutron Log Details Activities (4) Areas (4) Regions (0) NEPA(0) Exploration Technique Information Exploration Group: Downhole Techniques Exploration Sub Group: Well Log Techniques Parent Exploration Technique: Well Log Techniques Information Provided by Technique Lithology: if used in conjunction with other logs, this technique can provide information on the rock type and the porosity Stratigraphic/Structural: Corelation of rock units Hydrological: Estimate of formation porosity Thermal: 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 hydrogen can be used to identify zones of higher porosity.

296

Fiber/Matrix Interfacial Thermal Conductance Effect on the Thermal Conductivity of SiC/SiC Composites  

SciTech Connect

SiC/SiC composites used in fusion reactor applications are subjected to high heat fluxes and require knowledge and tailoring of their in-service thermal conductivity. Accurately predicting the thermal conductivity of SiC/SiC composites as a function of temperature will guide the design of these materials for their intended use, which will eventually include the effects of 14-MeV neutron irradiations. This paper applies an Eshelby-Mori-Tanaka approach (EMTA) to compute the thermal conductivity of unirradiated SiC/SiC composites. The homogenization procedure includes three steps. In the first step EMTA computes the homogenized thermal conductivity of the unidirectional (UD) SiC fiber embraced by its coating layer. The second step computes the thermal conductivity of the UD composite formed by the equivalent SiC fibers embedded in a SiC matrix, and finally the thermal conductivity of the as-formed SiC/SiC composite is obtained by averaging the solution for the UD composite over all possible fiber orientations using the second-order fiber orientation tensor. The EMTA predictions for the transverse thermal conductivity of several types of SiC/SiC composites with different fiber types and interfaces are compared to the predicted and experimental results by Youngblood et al.

Nguyen, Ba Nghiep; Henager, Charles H.

2013-04-20T23:59:59.000Z

297

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

E-Print Network (OSTI)

The objective of this research is to fully characterize the effectiveness of the Texas A&M University Nuclear Science Center Reactor (TAMU NSCR) neutrons for bacterial sterilization, and to assess the secondary gamma flux produced when neutrons collide with nuclei in biological materials. Sterilization of bacteria by exposure to gamma rays and charged particles is fairly well understood. Exposure to neutrons and gamma rays from fission as a means of sterilization has not to date been adequately characterized. The lack of data on the relationship between biological detriment resulting from thermal or fast neutron exposures and absorbed doses as applied in countermeasures to weapons of mass destruction (WMD) is the primary motivation for this investigation of neutron doses to endospores. Bacillus thuringiensis (Bt) spores were irradiated after producing and sampling them using standard microbiological procedures. Irradiation was accomplished using neutrons and gamma rays from the 1-MW TRIGA reactor at the TAMU NSCR using a reactor power of 100 kilowatts (kW). The combination of neutron and gamma-ray absorbed dose provided an effective means of sterilization of these types of spores; it yielded a 100-percent kill for the first study. Survival curves have been developed, from subsequent experiments, for these energy dependent neutron interactions with biological materials using a combination of radiation dosimetry, microbiological culture techniques, and computer modeling (Monte Carlo Neutral Particle history modeling - MCNP). Survival curves indicate a D?? value of 321.08 Gy. Additional work is needed to investigate the specific bacteria used in biological weapons in order to understand agent-specific radiation sensitivity. Once this is done, more effective and meaningful experiments can be conducted in order to tailor the neutron source strength to the robustness of the threat.

Hearnsberger, David Wayne

2001-01-01T23:59:59.000Z

298

Precision Measurement Of The Neutron's Beta Asymmetry Using Ultra-Cold Neutrons  

Science Conference Proceedings (OSTI)

A measurement of A{beta}, the correlation between the electron momentum and neutron (n) spin (the beta asymmetry) in n beta-decay, together with the n lifetime, provides a method for extracting fundamental parameters for the charged-current weak interaction of the nucleon. In particular when combined with decay measurements, one can extract the Vud element of the CKM matrix, a critical element in CKM unitarity tests. By using a new SD2 super-thermal source at LANSCE, large fluxes of UCN (ultra-cold neutrons) are expected for the UCNA project. These UCN will be 100% polarized using a 7 T magnetic field, and directed into the {beta} spectrometer. This approach, together with an expected large reduction in backgrounds, will result in an order of magnitude reduction in the critical systematic corrections associated with current n {beta}-asymmetry measurements. This paper will give an overview of the UCNA A{beta} measurement as well as an update on the status of the experiment.

Makela, M. [Los Alamos National Lab., P.O. Box 1663, Los Alamos, NM 87545 (United States); Back, H. O. [North Carolina State University Raleigh, NC 27695 (United States); Melconian, D. [University of Washington, Department of Physics, Box 351560 Seattle, WA 98195 (United States); Plaster, B. [California Institute of Technology, Kellogg Radiation Lab, Pasadena, CA 91125 (United States)

2006-07-11T23:59:59.000Z

299

Neutron Sources  

Science Conference Proceedings (OSTI)

... for Neutron Reaction Rate Measurements, JA Grundl, V. Spiegel, CM Eisenhauer, HT Heaton II, DM Gilliam (NBS), and J. Bigelow (ORNL), Nucl. ...

2013-07-27T23:59:59.000Z

300

Tracking heat flux sensors for concentrating solar applications  

DOE Patents (OSTI)

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.

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

2013-06-11T23:59:59.000Z

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


301

Plasma momentum meter for momentum flux measurements  

DOE Patents (OSTI)

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.

Zonca, Fulvio (Rome, IT); Cohen, Samuel A. (Hopewell, NJ); Bennett, Timothy (Princeton, NJ); Timberlake, John R. (Allentown, NJ)

1993-01-01T23:59:59.000Z

302

Plasma momentum meter for momentum flux measurements  

DOE Patents (OSTI)

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 {minus}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 an intense magnetic field and pulsed plasma environment.

Zonca, F.; Cohen, S.A.; Bennett, T.; Timberlake, J.R.

1991-12-31T23:59:59.000Z

303

Plasma momentum meter for momentum flux measurements  

DOE Patents (OSTI)

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 [minus]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 an intense magnetic field and pulsed plasma environment.

Zonca, F.; Cohen, S.A.; Bennett, T.; Timberlake, J.R.

1991-01-01T23:59:59.000Z

304

Optimization of Boron Neutron Capture Therapy for the Treatment of Undifferentiated Thyroid Cancer  

Science Conference Proceedings (OSTI)

Purpose: To analyze the possible increase in efficacy of boron neutron capture therapy (BNCT) for undifferentiated thyroid carcinoma (UTC) by using p-boronophenylalanine (BPA) plus 2,4-bis ({alpha},{beta}-dihydroxyethyl)-deutero-porphyrin IX (BOPP) and BPA plus nicotinamide (NA) as a radiosensitizer of the BNCT reaction. Methods and Materials: Nude mice were transplanted with a human UTC cell line (ARO), and after 15 days they were treated as follows: (1) control, (2) NCT (neutrons alone), (3) NCT plus NA (100 mg/kg body weight [bw]/day for 3 days), (4) BPA (350 mg/kg bw) + neutrons, (5) BPA + NA + neutrons, and (6) BPA + BOPP (60 mg/kg bw) + neutrons. The flux of the mixed (thermal + epithermal) neutron beam was 2.8 x 10{sup 8} n/cm{sup 2}/sec for 83.4 min. Results: Neutrons alone or with NA caused some tumor growth delay, whereas in the BPA, BPA + NA, and BPA + BOPP groups a 100% halt of tumor growth was observed in all mice at 26 days after irradiation. When the initial tumor volume was 50 mm{sup 3} or less, complete remission was found with BPA + NA (2 of 2 mice), BPA (1 of 4), and BPA + BOPP (7 of 7). After 90 days of complete regression, recurrence of the tumor was observed in BPA + NA (2 of 2) and BPA + BOPP (1 of 7). The determination of apoptosis in tumor samples by measurements of caspase-3 activity showed an increase in the BNCT (BPA + NA) group at 24 h (p < 0.05 vs. controls) and after the first week after irradiation in the three BNCT groups. Terminal transferase dUTP nick end labeling analysis confirmed these results. Conclusions: Although NA combined with BPA showed an increase of apoptosis at early times, only the group irradiated after the combined administration of BPA and BOPP showed a significantly improved therapeutic response.

Dagrosa, Maria Alejandra; Thomasz, Lisa M.Sc. [Department of Radiobiology (Constituyentes Atomic Center), National Atomic Energy Commission of Argentina, Buenos Aires (Argentina); Longhino, Juan [Nuclear Reactor RA-6 (Bariloche Atomic Center), National Atomic Energy Commission of Argentina, Buenos Aires (Argentina); Perona, Marina [Department of Radiobiology (Constituyentes Atomic Center), National Atomic Energy Commission of Argentina, Buenos Aires (Argentina); Calzetta, Osvaldo; Blaumann, Herman [Nuclear Reactor RA-6 (Bariloche Atomic Center), National Atomic Energy Commission of Argentina, Buenos Aires (Argentina); Rebagliati, Raul Jimenez [Department of Chemistry (Constituyentes Atomic Center), National Atomic Energy Commission of Argentina, Buenos Aires (Argentina); Cabrini, Romulo [Department of Radiobiology (Constituyentes Atomic Center), National Atomic Energy Commission of Argentina, Buenos Aires (Argentina); Kahl, Steven [Department of Pharmaceutical Chemistry, University of California, San Francisco, CA (United States); Juvenal, Guillermo Juan [Department of Radiobiology (Constituyentes Atomic Center), National Atomic Energy Commission of Argentina, Buenos Aires (Argentina); Pisarev, Mario Alberto [Department of Radiobiology (Constituyentes Atomic Center), National Atomic Energy Commission of Argentina, Buenos Aires (Argentina); Department of Human Biochemistry, School of Medicine, University of Buenos Aires, Buenos Aires (Argentina)], E-mail: pisarev@cnea.gov.ar

2007-11-15T23:59:59.000Z

305

Thermal conductivity of dense quark matter and cooling of stars  

E-Print Network (OSTI)

The thermal conductivity of the color-flavor locked phase of dense quark matter is calculated. The dominant contribution to the conductivity comes from photons and Nambu-Goldstone bosons associated with breaking of baryon number which are trapped in the quark core. Because of their very large mean free path the conductivity is also very large. The cooling of the quark core arises mostly from the heat flux across the surface of direct contact with the nuclear matter. As the thermal conductivity of the neighboring layer is also high, the whole interior of the star should be nearly isothermal. Our results imply that the cooling time of compact stars with color-flavor locked quark cores is similar to that of ordinary neutron stars.

Igor A. Shovkovy; Paul J. Ellis

2002-04-11T23:59:59.000Z

306

Neutronic reactor  

DOE Patents (OSTI)

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

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

1976-08-17T23:59:59.000Z

307

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

SciTech Connect

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).

Gray Chang

2012-03-01T23:59:59.000Z

308

U.S. Department of Energy Categorical Exclusion ...  

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

Installation of a High Flux Thermal Neutron Source Savannah River Site AikenAikenSouth Carolina A new High Flux Thermal Neutron Source (HFTNS), along with auxiliary equipment...

309

Neutron-driven gamma-ray laser  

DOE Patents (OSTI)

A lasing cylinder emits laser radiation at a gamma-ray wavelength of 0.87 {angstrom} when subjected to an intense neutron flux of about 400 eV neutrons. A 250 {angstrom} thick layer of Be is provided between two layers of 100 {angstrom} 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.

Bowman, C.D.

1989-03-28T23:59:59.000Z

310

Neutron-driven gamma-ray laser  

DOE Patents (OSTI)

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.

Bowman, Charles D. (Los Alamos, NM)

1990-01-01T23:59:59.000Z

311

Ultracold Neutron Production in a Pulsed Neutron Beam Line  

E-Print Network (OSTI)

We present the results of an Ultracold neutron (UCN) production experiment in a pulsed neutron beam line at the Los Alamos Neutron Scattering Center. The experimental apparatus allows for a comprehensive set of measurements of UCN production as a function of target temperature, incident neutron energy, target volume, and applied magnetic field. However, the low counting statistics of the UCN signal expected can be overwhelmed by the large background associated with the scattering of the primary cold neutron flux that is required for UCN production. We have developed a background subtraction technique that takes advantage of the very different time-of-flight profiles between the UCN and the cold neutrons, in the pulsed beam. Using the unique timing structure, we can reliably extract the UCN signal. Solid ortho-D$_2$ is used to calibrate UCN transmission through the apparatus, which is designed primarily for studies of UCN production in solid O$_2$. In addition to setting the overall detection efficiency in the...

Lavelle, C M; Manus, G; McChesney, P M; Salvat, D J; Shin, Y; Makela, M; Morris, C; Saunders, A; Couture, A; Young, A R; Liu, C -Y

2010-01-01T23:59:59.000Z

312

Neutron scattering workshop promotes high-pressure research ...  

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

long-term goals in these areas closer to reality, Oak Ridge National Laboratory (ORNL), home of the Spallation Neutron Source (SNS) and the High Flux Isotope Reactor, is hosting...

313

Long-Range Neutron Detection  

Science Conference Proceedings (OSTI)

A neutron detector designed for detecting neutron sources at distances of 50 to 100 m has been constructed and tested. This detector has a large surface area (1 m{sup 2}) to enhance detection efficiency, and it contains a collimator and shielding to achieve direction sensitivity and reduce background. An unusual feature of the detector is that it contains no added moderator, such as polyethylene, to moderate fast neutrons before they reach the {sup 3}He detector. As a result, the detector is sensitive mainly to thermal neutrons. The moderator-free design reduces the weight of the detector, making it more portable, and it also aids in achieving directional sensitivity and background reduction. Test results show that moderated fission-neutron sources of strength about 3 x 10{sup 5} n/s can be detected at a distance out to 70 m in a counting time of 1000 s. The best angular resolution of the detector is obtained at distances of 30 m or less. As the separation .distance between the source and detector increases, the contribution of scattered neutrons to the measured signal increases with a resultant decrease in the ability to detect the direction to a distant source. Applications for which the long-range detector appears to be suitable include detecting remote neutron sources (including sources in moving vehicles) and monitoring neutron storage vaults for the intrusion of humans and the effects they make on the detected neutron signal. Also, the detector can be used to measure waste for the presence of transuranic material in the presence of high gamma-ray background. A test with a neutron source (3 x 10{sup 5} n/s) in a vehicle showed that the detector could readily measure an increase in count rate at a distance of 10 m for vehicle speeds up to 35 mph (the highest speed tested). These results. indicate that the source should be detectable at this distance at speeds up to 55 mph.

AJ Peurrung; DC Stromswold; RR Hansen; PL Reeder; DS Barnett

1999-11-24T23:59:59.000Z

314

Modeling high-energy cosmic ray induced terrestrial muon flux: A lookup table  

E-Print Network (OSTI)

On geological timescales, the Earth is likely to be exposed to an increased flux of high energy cosmic rays (HECRs) from astrophysical sources such as nearby supernovae, gamma ray bursts or by galactic shocks. Typical cosmic ray energies may be much higher than the ~ 1 GeV flux which normally dominates. These high-energy particles strike the Earth's atmosphere initiating an extensive air shower. As the air shower propagates deeper, it ionizes the atmosphere by producing charged secondary particles. Secondary particles such as muons and thermal neutrons produced as a result of nuclear interactions are able to reach the ground, enhancing the radiation dose. Muons contribute 85% to the radiation dose from cosmic rays. This enhanced dose could be potentially harmful to the biosphere. This mechanism has been discussed extensively in literature but has never been quantified. Here, we have developed a lookup table that can be used to quantify this effect by modeling terrestrial muon flux from any arbitrary cosmic ray spectra with 10 GeV - 1 PeV primaries. This will enable us to compute the radiation dose on terrestrial planetary surfaces from a number of astrophysical sources.

Dimitra Atri; Adrian L. Melott

2010-11-19T23:59:59.000Z

315

NXS 2010 - Neutron Scattering School  

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

2-26, 2010 2-26, 2010 Argonne National Laboratory, Argonne, IL Oak Ridge National Laboratory, Oak Ridge, TN NXS2010 Travel Airport Shuttles Departure Flights Schedule Participants Lectures Lecturers Lecture Notes/Videos Experiments Schedule, Desc, Groups Student Presentations ANL Facilities APS Facility ANL Map ANL Visitor's Guide ORNL Facilities HFIR Facility SNS Facility HFIR/SNS Map Access Requirements ANL ORNL Rad Worker Training Study Guide Wireless Networks ANL ORNL Safety & Security Rules ANL ORNL NSSA New Initiatives NSSA Weblink Contacts ANL ORNL 12th National School on Neutron & X-ray Scattering 2009 Neutron Scattering School participants 2010 National School Participants Students share their thoughts about NXS 2010. Purpose: The main purpose of the National School on Neutron and X-ray Scattering is to educate graduate students on the utilization of major neutron and x-ray facilities. Lectures, presented by researchers from academia, industry, and national laboratories, will include basic tutorials on the principles of scattering theory and the characteristics of the sources, as well as seminars on the application of scattering methods to a variety of scientific subjects. Students will conduct four short experiments at Argonne's Advanced Photon Source and Oak Ridge's Spallation Neutron Source and High Flux Isotope Reactor facilities to provide hands-on experience for using neutron and synchrotron sources.

316

Quantitative interpretation of pulsed neutron capture logs: Part 1 --Fast numerical simulation  

E-Print Network (OSTI)

NEUTRON CAPTURE LOGS IN THINLY-BEDDED FORMATIONS Jordan G. Mimoun and Carlos Torres-Verdín, The University to capture neutrons. The lower the neutron energy, the more likely capture phenomena will take place; hence neutrons at thermal energies are the most likely to be absorbed. Consequently, monitoring the population

Torres-Verdín, Carlos

317

Tailoring the Neutron Spectrum from a 14-MeV Neutron Generator to Approximate a Spontaneous-Fission Spectrum  

SciTech Connect

Many applications of neutrons for non-invasive measurements began with isotopic sources such as AmBe or Cf-252. Political factors have rendered AmBe undesirable in the United States and other countries, and the supply of Cf-252 is limited and significantly increasing in price every few years. Compact and low-power deuterium-tritium (DT) electronic neutron generators can often provide sufficient flux, but the 14-MeV neutron spectrum is much more energetic (harder) than an isotopic neutron source. A series of MCNP simulations were run to examine the extent to which the 14-MeV DT neutron spectrum could be softened through the use of high-Z and low-Z materials. Some potential concepts of operation require a portable neutron generator system, so the additional weight of extra materials is also a trade-off parameter. Using a reference distance of 30 cm from the source, the average neutron energy can be lowered to be less than that of either AmBe or Cf-252, while obtaining an increase in flux at the reference distance compared to a bare neutron generator. This paper discusses the types and amounts of materials used, the resulting neutron spectra, neutron flux levels, and associated photon production.

James Simpson; David Chichester

2011-06-01T23:59:59.000Z

318

Neutron Cross Section Measurements at the Spallation Neutron Source  

Science Conference Proceedings (OSTI)

With the prospect of construction of the Spallation Neutron Source (SNS) at ORNL, and the fantastic high neutron flux, new, up to now impossible, experiments seem to be feasible in the fields of applied nuclear physics and astrophysics. These experiments will supply crucial neutron-induced cross section data for radionuclides, which are badly needed by many applied physics programs. The SNS will be uniquely suited for measuring the cross sections of interest to nuclear criticality safety, accelerator transmutation of nuclear waste (ATW), and heavy element nucleosynthesis for astrophysics. Because the sample sizes required at current facilities are usually too large for practical measurements, scarce information of these cross sections is available. Using the high neutron flux at the SNS will allow these measurements to be made with samples about 40 times smaller than at the next best facility. The large reduction in sample size at the SNS will result in orders of magnitude reduction in background from the radioactive samples and make them much easier to produce; hence, a much wider range of samples will be accessible for measurement at the SNS than at any other facility.

Guber, K.H.

2001-08-24T23:59:59.000Z

319

A Convective Transport Theory for Surface Fluxes  

Science Conference Proceedings (OSTI)

For a boundary layer in free convection where turbulent thermal structures communicate information between the surface and the interior of the mixed layer, it is hypothesized that the surface momentum flux can be parameterized by u*2 = bDwBMML, ...

Roland B. Stull

1994-01-01T23:59:59.000Z

320

NEUTRONIC REACTOR  

DOE Patents (OSTI)

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

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

1961-11-21T23:59:59.000Z

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


321

NEUTRON SOURCES  

DOE Patents (OSTI)

A neutron source is obtained without employing any separate beryllia receptacle, as was formerly required. The new method is safer and faster, and affords a source with both improved yield and symmetry of neutron emission. A Be container is used to hold and react with Pu. This container has a thin isolating layer that does not obstruct the desired Pu--Be reaction and obviates procedures previously employed to disassemble and remove a beryllia receptacle. (AEC)

Richmond, J.L.; Wells, C.E.

1963-01-15T23:59:59.000Z

322

Neutron range spectrometer  

DOE Patents (OSTI)

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.

Manglos, S.H.

1988-03-10T23:59:59.000Z

323

ORNL Neutron Sciences  

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

ORNL's Neutron Science Future: Integrating Neutron Scattering Across the Laboratory Greg Smith, HFIR Center for Neutron Scattering Upgrade Status and Scientific Opportunities...

324

Geometrical vector flux sinks and ideal flux concentrators  

SciTech Connect

The description of ideal flux concentrators as shapes that do not disturb the geometrical vector flux field is extended to all the known types of ideal flux concentrators. This is accomplished, in part, by the introduction of vector flux sinks.

Greenman, P.

1981-06-01T23:59:59.000Z

325

Neutron Instruments Added at Oak Ridge  

Science Conference Proceedings (OSTI)

The neutron scattering facilities at Oak Ridge National Laboratory continue their development as new instruments are commissioned and join the user program at the Spallation Neutron Source and High Flux Isotope Reactor. More than 640 proposals were received for beam time during the January-May 2011 period on SNS and HFIR instruments with about half either being accepted or identified as alternates. The proposal call for the period June-December 2011, announced at http://neutrons.ornl.gov, will close February 23, 2011.

Ekkebus, Allen E [ORNL

2011-01-01T23:59:59.000Z

326

Variable control of neutron albedo in toroidal fusion devices  

DOE Patents (OSTI)

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.

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

1983-06-01T23:59:59.000Z

327

Californium-252: neutron source for industry and medicine  

SciTech Connect

From eleventh conference on radioisotopes; Tokyo, Japan (13 Nov 1973). The history, production, and availability of /sup 252/Cf and its many potential uses are discussed. Applications in life sciences, education chemical analysis, exploration for natural resources, industrial process control, neutron radiography, nondestructive inspection, and neutron flux enhancement are described. (TFD)

Reinig, W.C.; Permar, P.H.; Cornman, W.R.

1973-01-01T23:59:59.000Z

328

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

SciTech Connect

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.

Gray S. Chang

2011-05-01T23:59:59.000Z

329

Science Opportunities at ORNL's Neutron Sources  

Science Conference Proceedings (OSTI)

The Neutron Sciences Directorate at Oak Ridge National Laboratory (ORNL) operates two of the world's most advanced neutron scattering research facilities: the Spallation Neutron Source (SNS) and the High Flux Isotope Reactor (HFIR). Our vision is to provide unprecedented capabilities for understanding structure and properties across the spectrum of biology, chemistry, physics, and engineering, and to stay at the leading edge of neutron science by developing new instruments, tools, and services. This talk will provide an update on the operations of the two research facilities and highlight the significant research that is emerging. For example, scientists from ORNL are at the forefront of research on a new class of iron-based superconductors based on experiments performed at the Triple-Axis Spectrometer at HFIR and ARCS at SNS. The complementary nature of neutron and x-ray techniques will be discussed to spark discussion among attendees.

Anderson, Ian [ORNL, SNS

2010-02-03T23:59:59.000Z

330

MFE/ACT: a TRS-80 code for calculating neutron activation  

Science Conference Proceedings (OSTI)

The MFE/ACT code, written to run on the TRS-80, can be used to calculate the neutron activation of materials used in fission and fusion reactors. Input data include the specific isotopes to be calculated, the neutron fluxes, the neutron cross sections, and the nuclear decay half-lives.

Dorn, D.W.

1982-10-01T23:59:59.000Z

331

The New Munich Neutron Source FRM II: Overview and Uses for Biological Studies  

E-Print Network (OSTI)

Neutron Physics at NIST M. Arif 8th UCN Workshop St. Petersburg ­ Moscow, Russia June 11-21, 2011 #12;NCNR Guide Hall 20 MW Reactor #12;Neutron Physics at the NCNR Beam Flux n cm-2 s-1 Peak Wavelength Facility Low Scatter Neutron Dosimeter Calibration Facility #12;December 31, 2012 Physics Physics Physics

Doster, Wolfgang

332

Lithium-6 filter for a fission converter-based Boron Neutron Capture Therapy irradiation facility beam  

E-Print Network (OSTI)

(cont.) A storage system was designed to contain the lithium-6 filter safely when it is not in use. A mixed field dosimetry method was used to measure the photon, thermal neutron and fast neutron dose. The measured advantage ...

Gao, Wei, Ph. D.

2005-01-01T23:59:59.000Z

333

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

SciTech Connect

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

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

2010-08-19T23:59:59.000Z

334

Forecasting neutron star temperatures: predictability and variability  

E-Print Network (OSTI)

It is now possible to model thermal relaxation of neutron stars after bouts of accretion during which the star is heated out of equilibrium by nuclear reactions in its crust. Major uncertainties in these models can be encapsulated in modest variations of a handful of fudge parameters that change the crustal thermal conductivity, specific heat, and heating rates. Observations of thermal relaxation constrain these fudge parameters and allow us to predict longer term variability in terms of the neutron star core temperature. We demonstrate this explicitly by modeling ongoing thermal relaxation in the neutron star XTE J1701-462. Its future cooling, over the next 5 to 30 years, is strongly constrained and depends mostly on its core temperature, uncertainties in crust physics having essentially been pinned down by fitting to the first three years of observations.

Dany Page; Sanjay Reddy

2013-07-17T23:59:59.000Z

335

Underground particle fluxes in the Soudan mine.  

E-Print Network (OSTI)

This is a summary of our knowledge of the underground particle fluxes in the vicinity of Soudan 2 and of the future MINOS detector. It includes a brief description of the measured muon fluxes and of the gamma ray spectra deduced from measurements of 238 U, 232 Th and 40 K concentrations in the rock. Counting rates in gaseous and scintillation detectors are estimated. Some data on what is known about the chemical composition of the local rocks are included; these are relevant to an understanding of the underground muon rates and also to a calculation of low energy neutron fluxes. 1 Introduction As plans for the MINOS detector and for the excavation of a new detector hall progress, some people have begun asking what is known of the fluxes of various particles underground. The muon flux is relevant for possibly calibrating and certainly for monitoring the long term behavior of the detector. It will likely be the determining factor in the eventual trigger rate if the MINOS det...

Keith Ruddick; Keith Ruddick; Th

1996-01-01T23:59:59.000Z

336

Neutron Repulsion  

E-Print Network (OSTI)

Earth is connected gravitationally, magnetically and electrically to its heat source - a neutron star that is obscured from view by waste products in the photosphere. Neutron repulsion is like the hot filament in an incandescent light bulb. Excited neutrons are emitted from the solar core and decay into hydrogen that glows in the photosphere like a frosted light bulb. Neutron repulsion was recognized in nuclear rest mass data in 2000 as the overlooked source of energy, the keystone of an arch that locked together these puzzling space-age observations: 1.) Excess 136Xe accompanied primordial helium in the stellar debris that formed the solar system (Fig. 1); 2.) The Sun formed on the supernova core (Fig. 2); 3.) Waste products from the core pass through an iron-rich mantle, selectively carrying lighter elements and lighter isotopes of each element into the photosphere (Figs. 3-4); and 4.) Neutron repulsion powers the Sun and sustains life (Figs. 5-7). Together these findings offer a framework for understanding how: a.) The Sun generates and releases neutrinos, energy and solar-wind hydrogen and helium; b.) An inhabitable planet formed and life evolved around an ordinary-looking star; c.) Continuous climate change - induced by cyclic changes in gravitational interactions of the Sun's energetic core with planets - has favored survival by adaptation.

Oliver K. Manuel

2011-02-08T23:59:59.000Z

337

Rigorous Simulation of X-Ray Thermal Diffuse Scattering  

Science Conference Proceedings (OSTI)

In-Situ Neutron Diffraction and Crystal Plasticity Modeling of a-Uranium · In-Situ Studies of the ... Thermal Residual Stresses and Strains in Depleted Uranium.

338

Intergranular Thermal Residual Strain in Rolled and Texture-free ? ...  

Science Conference Proceedings (OSTI)

In this study, the intergranular thermal residual strains are determined from ... rolled and texture-free ?-uranium measured by neutron diffraction during cooling.

339

Californium Neutron Irradiation Facility  

Science Conference Proceedings (OSTI)

Californium Neutron Irradiation Facility. Summary: ... Cf irradiation facility (Photograph by: Neutron Physics Group). Lead Organizational Unit: pml. Staff: ...

2013-07-23T23:59:59.000Z

340

Neutron Physics Group  

Science Conference Proceedings (OSTI)

... spectrum and fluencies is essential for several ... Neutron Interferometer and Optics Facility performed a ... other neutron scattering facilities depends on ...

2011-10-24T23:59:59.000Z

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


341

Neutronic evaluation of LEU 30-20 fuel for the Texas A&M Nuclear Science Center Reactor  

E-Print Network (OSTI)

A neutronic evaluation of the Texas A&M University Nuclear Science Center Reactor (TAMU NSCR) using the General Atomic Company (GA) low enrichment uranium (LEU 30-20 fuel was performed to determine the feasibility of this type of fuel. To perform this evaluation, the WIMSD4m transport code and DIF3D diffusion code were utilized. These codes were provided by Argonne National Laboratory (ANL). WIMSD4m was used to calculate macroscopic cross-sections for the various core materials and DIF3D was used to calculate the effective multiplication factor and thermal neutron flux for various core configurations. In order to benchmark these codes and the core model used to evaluate the proposed LEU 30-20 core, the current FLIP core was first modeled. Various neutronic parameters, such as excess reactivity, shutdown margin, critical rod height and control rod worths were calculated using the core model and codes, and the results compared with actual experimental results for the FLIP core. Once the core model and codes were validated, the core model was modified using LEU 30-20 fuel and an optimum core configuration obtained which satisfied certain design criteria, including values for excess reactivity, shutdown margin and thermal neutron flux. In addition to modeling the LEU 30-20 core, a control rod model was developed to generate effective macroscopic cross-sections for the control rod materials in the core, based on previous work in this area performed at ANL. The results of this analysis indicate the feasibility of converting the TAMU NSCR to GA LEU 30-20 fuel.

Bigler, Mark Andrew

1996-01-01T23:59:59.000Z

342

Oceanic Heat Flux Calculation  

Science Conference Proceedings (OSTI)

The authors review the procedure for the direct calculation of oceanic heat flux from hydrographic measurements and set out the full “recipe” that is required.

Sheldon Bacon; Nick Fofonoff

1996-12-01T23:59:59.000Z

343

Chapter 13 - NEUTRON AREA DETECTORS 1. NEUTRON ...  

Science Conference Proceedings (OSTI)

... The neutron peak corresponds to both reaction products being entirely absorbed in the ... 6. A fission chamber is a very low efficiency neutron detector ...

2009-11-29T23:59:59.000Z

344

Neutron Generators for Spent Fuel Assay  

SciTech Connect

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.

Ludewigt, Bernhard A

2010-12-30T23:59:59.000Z

345

Application of Neutron Imaging to InvestigateFlow Through Fractures for EGS  

SciTech Connect

This paper will describe ongoing efforts at Oak Ridge National Laboratory to develop a unique experimental capability for investigating flow through porous and fractured geological media using neutron imaging techniques. This capability is expected to support numerous areas of investigation associated with flow processes relevant to EGS including, but not limited to: experimental visualization and measurement of velocity profiles and other flow characteristics to better inform reduced-order modeling of flow through fractures; laboratory scale validation of flow models and simulators; and a 'real-time' tool for studying geochemical rock/fluid interactions by noninvasively measuring material effects such as precipitation and dissolution in EGS representative conditions. Neutron scattering and attenuation based techniques have many distinctive advantages over other radiographic imaging methods for studying certain types of physical processes because cold and thermal neutrons are more highly attenuated by materials with large Hydrogen compositions while they more easily penetrate higher Z materials, such as those used in structural applications. Experiments exploiting this behavior may therefore be devised to study flow behaviors in samples even when thick pressure vessel walls and large sample masses are present. The objective of this project is to develop an experimental setup and methodology for taking EGS representative core samples with engineered fractures and fracture features, subjecting them to a triaxial stress state at EGS representative temperatures (up to 300 degrees C), and pumping high pressure fluid through the sample while imaging and measuring fluid flow characteristics using high flux neutron beams. This effort will take advantage of signature Oak Ridge National Laboratory facilities, including the Spallation Neutron Source and High Flux Isotope Reactor, as well as its core expertise in Neutron Science. Towards this end, a geothermal pressure test cell and flow system has been developed that can accommodate 1.5 diameter, 6 long core samples and apply a radial confining pressure up to 10,000 psi with fluid flow pressures up to 5,000 psi. This cell has been specially designed to optimize the transmission of neutrons and permit improved imaging of the interior of the sample of interest. Proof of principle measurements of the system have been performed and will be discussed in this paper. Techniques for injecting fluid contrast agents to permit visualization and quantification of flow profiles are also being developed and will be described along with future development plans.

Polsky, Yarom [ORNL; Anovitz, Lawrence {Larry} M [ORNL; Bilheux, Hassina Z [ORNL; Bingham, Philip R [ORNL; Carmichael, Justin R [ORNL

2013-01-01T23:59:59.000Z

346

Materials Selection for the HFIR Cold Neutron Source  

DOE Green Energy (OSTI)

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.

Farrell, K.

2001-08-24T23:59:59.000Z

347

NEUTRONIC REACTOR  

DOE Patents (OSTI)

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

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

1958-04-15T23:59:59.000Z

348

Energy flux limitation by tame turbulence  

SciTech Connect

A quasi-linear theory of energy flux limitation by ion acoustic turbulence is presented. This distribution function is modelled by a Maxwellian plus an additional piece which carries a heat flux Q. By taking the fourth moment of the Vlasov equation one finds the anomalous thermal conductivity K approximately 3 v/sub e/ delta/sub De/ (e phi/T/sub e/)$sup -2$. Other moments treated self-consistently include anomalous ion heating, electron cooling, current generation and enhanced inverse bremsstrahlung due to the nonthermal ion fluctuations. (auth)

Manheimer, W.M.

1975-11-01T23:59:59.000Z

349

Concentration with uniform flux  

SciTech Connect

A modification of a parabolic cylinder concentrator is developed to procedure uniform flux. The controlling surface equation is given. A three-dimensional ray-trace technique is used to obtain the shape of the image at the focal plane of a thin slice of the mirror. Also, the concentration distribution for uniform flux is given. 1 references, 7 figures.

Not Available

1986-01-01T23:59:59.000Z

350

Summary report on four Oak Ridge sensors for enhancing nuclear safeguards neutron detectors  

SciTech Connect

The need for monitoring weapons grade Pu in nuclear facilities worldwide was addressed with four radiation detector technologies being developed at Y-12 and ORNL. This paper describes experimental results of 4 Oak Ridge Sensors for Enhancing Nuclear Safeguards (ORSENS) neutron detector technologies and includes the potential application, cost, and advantages for each. These are a {sup 6}LiF- ZnS(Ag) thermal neutron scintillator coupled to a wavelength-shifting optical fiber, a CdWO{sub 4} based scintillating thermal neutron detector, a rhodium silicon thermal neutron detector, and a proton- recoil fast neutron detector.

Williams, J.A.; Clark, R.L.; Hutchinson, D.P.; Miller, V.C.; Ramsey, J.A. [Oak Ridge National Lab., TN (United States); Bell, Z.W.; Hiller, J.M.; Wallace, S.A. [Oak Ridge Y-12 Plant, TN (United States)

1997-08-01T23:59:59.000Z

351

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

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

247: Construction and Operation of the Spallation Neutron 247: Construction and Operation of the Spallation Neutron Source EIS-0247: Construction and Operation of the Spallation Neutron Source SUMMARY 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 scattering research than is currently available. This source would assure the availability of a state-of-the-art neutron research facility in the United States in the decades ahead. This facility would be used to conduct research in areas such as materials science, condensed matter physics, the molecular structure of biological materials, properties of polymers and complex fluids, and magnetism. In addition to creating new scientific and

352

Design Analyses and Shielding of HFIR Cold Neutron Scattering Instruments  

Science Conference Proceedings (OSTI)

Research reactor geometries and special characteristics present unique dosimetry analysis and measurement issues. The introduction of a cold neutron moderator and the production of cold neutron beams at the Oak Ridge National Laboratory High Flux Isotope Reactor have created the need for modified methods and devices for analyzing and measuring low energy neutron fields (0.01 to 100 meV). These methods include modifications to an MCNPX version to provide modeling of neutron mirror reflection capability. This code has been used to analyze the HFIR cold neutron beams and to design new instrument equipment that will use the beams. Calculations have been compared with time-of-flight measurements performed at the start of the neutron guides and at the end of one of the guides. The results indicate that we have a good tool for analyzing the transport of these low energy beams through neutron mirror and guide systems for distance up to 60 meters from the reactor. (authors)

Gallmeier, F.X.; Selby, D.L.; Winn, B.; Stoica, D.; Jones, A.B.; Crow, L. [Neutron Sciences Directorate, Oak Ridge National Laboratory (United States)

2011-07-01T23:59:59.000Z

353

Neutron Imaging Reveals Internal Plant Hydraulic Dynamics  

SciTech Connect

Many terrestrial ecosystem processes are constrained by water availability and transport within the soil. Knowledge of plant water fluxes is thus critical for assessing mechanistic processes linked to biogeochemical cycles, yet resolution of root structure and xylem water transport dynamics has been a particularly daunting task for the ecologist. Through neutron imaging, we demonstrate the ability to non-invasively monitor individual root functionality and water fluxes within Zea mays L. (maize) and Panicum virgatum L. (switchgrass) seedlings growing in a sandy medium. Root structure and growth were readily imaged by neutron radiography and neutron computed tomography. Seedlings were irrigated with water or deuterium oxide and imaged through time as a growth lamp was cycled on to alter leaf demand for water. Sub-millimeter scale resolution reveals timing and magnitudes of root water uptake, redistribution within the roots, and root-shoot hydraulic linkages, relationships not well characterized by other techniques.

Warren, Jeffrey [ORNL; Bilheux, Hassina Z [ORNL; Kang, Misun [ORNL; Voisin, Sophie [ORNL; Cheng, Chu-Lin [ORNL; Horita, Jusuke [ORNL; Perfect, Edmund [ORNL

2013-01-01T23:59:59.000Z

354

Continued Neutron Star Crust Cooling of the 11 Hz X-Ray Pulsar in Terzan 5: A Challenge to Heating and Cooling Models?  

E-Print Network (OSTI)

The transient neutron star low-mass X-ray binary and 11 Hz X-ray pulsar IGR J17480-2446 in the globular cluster Terzan 5 exhibited an 11-week accretion outburst in 2010. Chandra observations performed within five months after the end of the outburst revealed evidence that the crust of the neutron star became substantially heated during the accretion episode and was subsequently cooling in quiescence. This provides the rare opportunity to probe the structure and composition of the crust. Here, we report on new Chandra observations of Terzan 5 that extend the monitoring to ~2.2 yr into quiescence. We find that the thermal flux and neutron star temperature have continued to decrease, but remain significantly above the values that were measured before the 2010 accretion phase. This suggests that the crust has not thermally relaxed yet, and may continue to cool. Such behavior is difficult to explain within our current understanding of heating and cooling of transiently accreting neutron stars. Alternatively, the q...

Degenaar, N; Brown, E F; Altamirano, D; Cackett, E M; Fridriksson, J; Homan, J; Heinke, C O; Miller, J M; Pooley, D; Sivakoff, G R

2013-01-01T23:59:59.000Z

355

International Training Program in Support of Safety Analysis: 3D S.UN.COP-Scaling, Uncertainty and 3D Thermal-Hydraulics/Neutron-Kinetics Coupled Codes Seminars  

Science Conference Proceedings (OSTI)

Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers and vendors, nuclear fuel companies, research organizations, consulting companies, and technical support organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the 'user effect' and stems from the limitations embedded in the codes as well as from the limited capability of the analysts to use the codes. Code user training and qualification is an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. The 3D S.UN.COP (Scaling, Uncertainty and 3D COuPled code calculations) seminars have been organized as follow-up of the proposal to IAEA for the Permanent Training Course for System Code Users [1]. Five seminars have been held at University of Pisa (2003, 2004), at The Pennsylvania State University (2004), at University of Zagreb (2005) and at the School of Industrial Engineering of Barcelona (2006). It was recognized that such courses represented both a source of continuing education for current code users and a mean for current code users to enter the formal training structure of a proposed 'permanent' stepwise approach to user training. The 3D S.UN.COP 2006 was successfully held with the attendance of 33 participants coming from 18 countries and 28 different institutions (universities, vendors, national laboratories and regulatory bodies). More than 30 scientists (coming from 13 countries and 23 different institutions) were involved in the organization of the seminar, presenting theoretical aspects of the proposed methodologies and holding the training and the final examination. A certificate (LA Code User grade) was released to participants that successfully solved the assigned problems. A sixth seminar will be organized in 2007 at the Texas A and M University involving more than 30 scientists between lecturers and code developers. (http://dimnp.ing.unipi.it/3dsuncop/2007). (authors)

Petruzzi, Alessandro; D'Auria, Francesco [University of Pisa, Lungarno Pacinotti, 43 - 56126 Pisa (Italy); Bajs, Tomislav [University of Zagreb, Trg marsala Tita 14, HR-10000 Zagreb (Croatia); Reventos, Francesc [School of Industrial Engineering, Technical University of Catalonia - UPC, Seccion de Ingeniera Nuclear, Av. Diagonal No. 647, Pabellon C, 08028 Barcelona (Spain)

2006-07-01T23:59:59.000Z

356

NEUTRONIC REACTORS  

DOE Patents (OSTI)

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

Anderson, H.L.

1958-10-01T23:59:59.000Z

357

Thermal Analysis of a Uranium Silicide Miniplate Irradiation Experiment  

Science Conference Proceedings (OSTI)

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.

Donna Post Guillen

2009-09-01T23:59:59.000Z

358

Scientific Upgrades at the Oak Ridge National Laboratory High Flux Isotope Reactor  

Science Conference Proceedings (OSTI)

The United States Department of Energy is sponsoring a number of projects that will provide scientific upgrades to the neutron science facilities associated with the High Flux Isotope Reactor (HFIR) located at Oak Ridge National Laboratory. Funding for the first upgrade project was initiated in 1996 and all presently identified upgrade projects are expected to be completed by the end of 2003. The upgrade projects include: (1) larger beam tubes, (2) a new monochromator drum for the HB-1 beam line, (3) a new HB-2 beam line system that includes one thermal guide and a new monochromator drum, (4) new instruments for the HB-2 beamline, (5) a new monochromator drum for the HB-3 beam line, (6) a supercritical hydrogen cold source system to be retrofitted into the HB-4 beam tube, (7) a 3.5 kW refrigeration system at 20 K to support the cold source and a new building to house it, (8) a new HB-4 beam line system composed of four cold neutron guides with various mirror coatings and associated shielding, (9) a number of new instruments for the cold beams including two new SANS instruments, and (10) construction of support buildings. This paper provides a short summary of these projects including their present status and schedule.

Selby, Douglas L [ORNL; Jones, Amy [ORNL; Crow, Lowell [ORNL

2012-01-01T23:59:59.000Z

359

Neutron reflecting supermirror structure  

DOE Patents (OSTI)

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.

Wood, James L. (Drayton Plains, MI)

1992-01-01T23:59:59.000Z

360

Neutron reflecting supermirror structure  

DOE Patents (OSTI)

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.

Wood, J.L.

1992-12-01T23:59:59.000Z

Note: This page contains sample records for the topic "thermal neutron flux" from the National Library of EnergyBeta (NLEBeta).
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they are not comprehensive nor are they the most current set.
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to obtain the most current and comprehensive results.


361

Advanced Neutron Source (ANS) Project progress report  

SciTech Connect

This report discusses the following topics on the advanced neutron source: quality assurance (QA) program; reactor core development; fuel element specification; corrosion loop tests and analyses; thermal-hydraulic loop tests; reactor control concepts; critical and subcritical experiments; material data, structural tests, and analysis; cold source development; beam tube, guide, and instrument development; hot source development; neutron transport and shielding; I C research and development; facility concepts; design; and safety.

McBee, M.R.; Chance, C.M. (eds.) (Oak Ridge National Lab., TN (USA)); Selby, D.L.; Harrington, R.M.; Peretz, F.J. (Oak Ridge National Lab., TN (USA))

1990-04-01T23:59:59.000Z

362

Computing Solar Absolute Fluxes  

E-Print Network (OSTI)

Computed color indices and spectral shapes for individual stars are routinely compared with observations for essentially all spectral types, but absolute fluxes are rarely tested. We can confront observed irradiances with the predictions from model atmospheres for a few stars with accurate angular diameter measurements, notably the Sun. Previous calculations have been hampered by inconsistencies and the use of outdated atomic data and abundances. I provide here a progress report on our current efforts to compute absolute fluxes for solar model photospheres. Uncertainties in the solar composition constitute a significant source of error in computing solar radiative fluxes.

Prieto, Carlos Allende

2007-01-01T23:59:59.000Z

363

Computing Solar Absolute Fluxes  

E-Print Network (OSTI)

Computed color indices and spectral shapes for individual stars are routinely compared with observations for essentially all spectral types, but absolute fluxes are rarely tested. We can confront observed irradiances with the predictions from model atmospheres for a few stars with accurate angular diameter measurements, notably the Sun. Previous calculations have been hampered by inconsistencies and the use of outdated atomic data and abundances. I provide here a progress report on our current efforts to compute absolute fluxes for solar model photospheres. Uncertainties in the solar composition constitute a significant source of error in computing solar radiative fluxes.

Carlos Allende Prieto

2007-09-14T23:59:59.000Z

364

Safety control circuit for a neutronic reactor  

DOE Patents (OSTI)

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

Ellsworth, Howard C. (Richland, WA)

2004-04-27T23:59:59.000Z

365

Design of the Mechanical Parts for the Neutron Guide System at HANARO  

SciTech Connect

The research reactor HANARO (High-flux Advanced Neutron Application ReactOr) in Korea will be equipped with a neutron guide system, in order to transport cold neutrons from the neutron source to the neutron scattering instruments in the neutron guide hall near the reactor building. The neutron guide system of HANARO consists of the in-pile plug assembly with in-pile guides, the primary shutter with in-shutter guides, the neutron guides in the guide shielding room with dedicated secondary shutters, and the neutron guides connected to the instruments in the neutron guide hall. Functions of the in-pile plug assembly are to shield the reactor environment from nuclear radiation and to support the neutron guides and maintain them precisely oriented. The primary shutter is a mechanical structure to be installed just after the in-pile plug assembly, which stops neutron flux on demand. This paper describes the design of the in-pile assembly and the primary shutter for the neutron guide system at HANARO. The design of the guide shielding assembly for the primary shutter and the neutron guides is also presented.

Shin, J. W.; Cho, Y. G.; Cho, S. J.; Ryu, J. S. [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

2008-03-17T23:59:59.000Z

366

Neutron Scattering Facilities 1982  

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

NEUTRON SOURCES NEUTRON SOURCES Types of Sources U.S. Sources Available for Users Plans for the Future The Neutron Scattering Society of America (NSSA) SNS/ANL School on Neutron and x-Ray Scattering, June 2011 Jim Rhyne Lujan Neutron Scattering Center Los Alamos National Lab. What do we need to do neutron scattering? * Neutron Source - produces neutrons * Diffractometer or Spectrometer - Allows neutrons to interact with sample - Sorts out discrete wavelengths by monochromator (reactor) or by time of flight (pulse source) - Detectors pick up neutrons scattered from sample * Analysis methods to determine material properties * Brain power to interpret results Sources of neutrons for scattering * Nuclear Reactor - Neutrons produced from fission of 235 U - Fission spectrum neutrons

367

Self-regulating neutron coincidence counter  

DOE Patents (OSTI)

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.

Baron, N.

1980-06-16T23:59:59.000Z

368

CRC handbook of fast neutron generators  

Science Conference Proceedings (OSTI)

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.

Csikai, G.J.

1987-01-01T23:59:59.000Z

369

Biology and Soft Matter | Neutron Sciences | ORNL  

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

Biology and Soft Matter Biology and Soft Matter SHARE Biology and Soft Matter This is a time of unprecedented opportunity for using neutrons in biological and soft matter research. The US Department of Energy (DOE) has invested in two forefront neutron user facilities, the accelerator-based Spallation Neutron Source (SNS) and the reactor-based High Flux Isotope Reactor (HFIR), at Oak Ridge National Laboratory (ORNL). Researchers have access to new instrumentation on some of the world's most intense neutron beam lines for studying the structure, function, and dynamics of complex systems. We anticipate that soft matter and biological sciences of tomorrow will require understanding, predicting, and manipulating complex systems to produce the new materials and products required to meet our nation's

370

Characterization of a ballistic supermirror neutron guide  

E-Print Network (OSTI)

We describe the beam characteristics of the first ballistic supermirror neutron guide H113 that feeds the neutron user facility for particle physics PF1B of the Institute Laue-Langevin, Grenoble (ILL). At present, the neutron capture flux density of H113 at its 20x6cm2 exit window is 1.35x10^10/cm^2/s, and will soon be raised to above 2x10^10/cm^2/s. Beam divergence is no larger than beam divergence from a conventional Ni coated guide. A model is developed that permits rapid calculation of beam profiles and absolute event rates from such a beam. We propose a procedure that permits inter-comparability of the main features of beams emitted from ballistic or conventional neutron guides.

H. Abele; D. Dubbers; H. Haese; M. Klein; A. Knoepfler; M. Kreuz; T. Lauer; B. Maerkisch; D. Mund; V. Nesvizhevsky; A. Petoukhov; C. Schmidt; M. Schumann; T. Soldner

2005-10-26T23:59:59.000Z

371

Monte Carlo next-event point flux estimation for RCP01  

SciTech Connect

Two next event point estimators have been developed and programmed into the RCP01 Monte Carlo program for solving neutron transport problems in three-dimensional geometry with detailed energy description. These estimators use a simplified but accurate flux-at-a-point tallying technique. Anisotropic scattering in the lab system at the collision site is accounted for by determining the exit energy that corresponds to the angle between the location of the collision and the point detector. Elastic, inelastic, and thermal kernel scattering events are included in this formulation. An averaging technique is used in both estimators to eliminate the well-known problem of infinite variance due to collisions close to the point detector. In a novel approach to improve the estimator`s efficiency, a Russian roulette scheme based on anticipated flux fall off is employed where averaging is not appropriate. A second estimator successfully uses a simple rejection technique in conjunction with detailed tracking where averaging isn`t needed. Test results show good agreement with known numeric solutions. Efficiencies are examined as a function of input parameter selection and problem difficulty.

Martz, R.L.; Gast, R.C.; Tyburski, L.J.

1991-12-31T23:59:59.000Z

372

Neutron Science In the News - 2014 | ORNL Neutron Sciences  

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

Neutron Science In the News - 2014 Neutron Science In the News - 2014 Because some media sources archive past articles and require a subscription for access, some of the links below might not be active. If a citation listed here is no longer available, please contact the newspaper or your library directly. January Multiphysics Simulations Transmuting Designs for Safer Nuclear Power Engineering.com 1/7 Like the rest of the US's nuclear research reactors, Oak Ridge National Lab's (ORNL) high flux isotope reactor (HFIR) is moving from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU). As such, the safety of the system must be assessed to incorporate the changes in fuel properties and the subsequently modified fuel plate. Due to the recent growth in multiphysics, fluid-structure dynamics

373

Photovoltaic roof heat flux  

E-Print Network (OSTI)

of ~24°C, indicating that heat conduction was small. T h i sday, indicating large heat conduction a n d storage. Control2.1.3 showed that conduction heat flux through the roof was

Samady, Mezhgan Frishta

2011-01-01T23:59:59.000Z

374

Thermal Hydraulic Effect of Fuel Plate Surface Roughness  

Science Conference Proceedings (OSTI)

This study presents surface roughness measurements characteristic of the pre-film layer applied to a typical Advanced Test Reactor (ATR) fuel plate. This data is used to estimate the friction factor for thermal hydraulic flow calculations of a Gas Test Loop (GTL) system proposed for incorporation into ATR to provide a fast neutron flux environment for the testing of nuclear fuels and materials. To attain the required neutron flux, the design includes booster fuel plates clad with the same aluminum alloy as the ATR driver fuel and cooled with water supplied by the ATR primary coolant pumps. The objectives of this study are to: (1) determine the surface roughness of the protective boehmite layer applied to the ATR driver fuel prior to reactor operations in order to specify the machining tolerances for the surface finish on simulated booster fuel plates in a GTL hydraulic flow test model, and (2) assess the consequent thermal hydraulic impact due to surface roughness on the coolability of the booster fuel with a similar pre-film layer applied. While the maximum roughness of this coating is specified to be 1.6 µm (63 microinches), no precise data on the actual roughness were available. A representative sample coupon autoclaved with the ATR driver fuel to produce the pre-film coating was analyzed using optical profilometry. Measurements yielded a mean surface roughness of 0.53 µm (21 microinches). Results from a sensitivity study show that a ±15% deviation from the mean measured surface finish would have a minimal effect on coolant temperature, coolant flow rate, and fuel temperature. However, frictional losses from roughnesses greater than 1.5 µm (~60 microinches) produce a marked decrease in flow rate, causing fuel and coolant temperatures to rise sharply.

Donna Post Guillen; Timothy S. Yoder

2008-09-01T23:59:59.000Z

375

Comparison of Calculated and Measured Neutron Fluence in Fuel/Cladding Irradiation Experiments in HFIR  

Science Conference Proceedings (OSTI)

A recently-designed thermal neutron irradiation facility has been used for a first series of irradiations of PWR fuel pellets in the high flux isotope reactor (HFIR) at Oak Ridge National Laboratory. Since June 2010, irradiations of PWR fuel pellets made of UN or UO{sub 2}, clad in SiC, have been ongoing in the outer small VXF sites in the beryllium reflector region of the HFIR, as seen in Fig. 1. HFIR is a versatile, 85 MW isotope production and test reactor with the capability and facilities for performing a wide variety of irradiation experiments. HFIR is a beryllium-reflected, light-water-cooled and -moderated, flux-trap type reactor that uses highly enriched (in {sup 235}U) uranium (HEU) as the fuel. The reactor core consists of a series of concentric annular regions, each about 2 ft (0.61 m) high. A 5-in. (12.70-cm)-diam hole, referred to as the flux trap, forms the center of the core. The fuel region is composed of two concentric fuel elements made up of many involute-shaped fuel plates: an inner element that contains 171 fuel plates, and an outer element that contains 369 fuel plates. The fuel plates are curved in the shape of an involute, which provides constant coolant channel width between plates. The fuel (U{sub 3}O{sub 8}-Al cermet) is nonuniformly distributed along the arc of the involute to minimize the radial peak-to-average power density ratio. A burnable poison (B{sub 4}C) is included in the inner fuel element primarily to reduce the negative reactivity requirements of the reactor control plates. A typical HEU core loading in HFIR is 9.4 kg of {sup 235}U and 2.8 g of {sup 10}B. The thermal neutron flux in the flux trap region can exceed 2.5 x 10{sup 15} n/cm{sup 2} {center_dot} s while the fast flux in this region exceeds 1 x 10{sup 15} n/cm{sup 2} {center_dot} s. The inner and outer fuel elements are in turn surrounded by a concentric ring of beryllium reflector approximately 1 ft (0.30 m) thick. The beryllium reflector consists of three regions: the removable reflector, the semi-permanent reflector, and the permanent reflector. It is surrounded by a water reflector of effectively infinite thickness. In the axial direction, the reactor is reflected by water above and below the reactor. The irradiation facilities, one for UN and the other for UO{sub 2} pellets, utilize a thin cylindrical hafnium shield approximately 4 cm in diameter surrounding the facility basket to reduce the thermal neutron flux sufficiently such that the linear power rating in the irradiated fuel pins will be similar to PWR operating conditions. The facilities each contain nine fuel pins, each comprising 10 fuel pellets, arranged as if three fuel rods.

Ellis, Ronald James [ORNL

2011-01-01T23:59:59.000Z

376

Advanced Neutron Source enrichment study  

SciTech Connect

A study has been performed of the impact on performance of using low enriched uranium (20% {sup 235}U) or medium enriched uranium (35% {sup 235}U) as an alternative fuel for the Advanced Neutron Source, which is currently designed to use uranium enriched to 93% {sup 235}U. Higher fuel densities and larger volume cores were evaluated at the lower enrichments in terms of impact on neutron flux, safety, safeguards, technical feasibility, and cost. The feasibility of fabricating uranium silicide fuel at increasing material density was specifically addressed by a panel of international experts on research reactor fuels. The most viable alternative designs for the reactor at lower enrichments were identified and discussed. Several sensitivity analyses were performed to gain an understanding of the performance of the reactor at parametric values of power, fuel density, core volume, and enrichment that were interpolations between the boundary values imposed on the study or extrapolations from known technology.

Bari, R.A.; Ludewig, H.; Weeks, J.R.

1994-12-31T23:59:59.000Z

377

The Effect of Neutron Irradiation on the Fracture Toughness of Graphite  

SciTech Connect

As part of our irradiated graphite recycle program a small quantity of PCEA grade graphite was irradiated in the High Flux Isotope Reactor (HFIR) at ORNL. The graphite will provide the raw material for future recycle experiments. The geometry of the irradiated graphite allowed us to study the effects of neutron irradiation on the Critical Stress Intensity Factor, KIc, of graphite. The specimens where irradiated in two groups of 6 at an irradiation temperature of 900 C in rabbit capsules to doses of 6.6 and 10.2 DPA, respectively. Following a full suite of pre-and post-irradiation examination, which included dimensions, mass, electrical resistivity, elastic constants, and thermal expansion (to 800 C) the samples were notched and tested to determine their KIc using the newly approved ATSM test method for SENB fracture toughness of graphite. Here we report the irradiation induced changes in the dimensions, elastic constants, resistivity, and coefficient of thermal expansion of PCEA graphite. Moreover, irradiation induced changes in the Critical Stress Intensity Factor, KIc, or fracture toughness, are reported and discussed. Very little work on the effect of neutron irradiation on the fracture toughness of graphite has previously be performed or reported.

Burchell, Timothy D [ORNL; Strizak, Joe P [ORNL

2012-01-01T23:59:59.000Z

378

Development of a three-dimensional two-fluid code with transient neutronic feedback for LWR applications  

E-Print Network (OSTI)

The development of a three-dimensional coupled neutronics/thermalhydraulics code for LWR safety analysis has been initiated. The transient neutronics code QUANDRY has been joined to the two-fluid thermal-hydraulics code ...

Griggs, D. P.

1981-01-01T23:59:59.000Z

379

SIMULATION OF CARGO CONTAINER INTERROGATION BY D-D NEUTRONS  

E-Print Network (OSTI)

or photons to induce fission in the nuclear material andearlier, the fission-induced signals from nuclear materialsnuclear materials are less than 1% (e.g. 0.685 % for thermal neutron fission

Lou, Tak Pui; Antolak, Arlyn

2007-01-01T23:59:59.000Z

380

22.54 Neutron Interactions and Applications, Spring 2002  

E-Print Network (OSTI)

Comprehensive treatment of neutron interactions in condensed matter at energies from thermal to MeV, focusing on aspects most relevant to radiation therapy, industrial imaging, and materials research applications. Comparative ...

Yip, Sidney

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


381

Packed rod neutron shield for fast nuclear reactors  

DOE Patents (OSTI)

A fast neutron nuclear reactor including a core and a plurality of vertically oriented neutron shield assemblies surrounding the core. Each assembly includes closely packed cylindrical rods within a polygonal metallic duct. The shield assemblies are less susceptible to thermal stresses and are less massive than solid shield assemblies, and are cooled by liquid coolant flow through interstices among the rods and duct.

Eck, John E. (Hempfield Township, Westmoreland County, PA); Kasberg, Alvin H. (Murrysville, PA)

1978-01-01T23:59:59.000Z

382

Flux lattices reformulated  

E-Print Network (OSTI)

We theoretically explore the optical flux lattices produced for ultra-cold atoms subject to laser fields where both the atom-light coupling and the effective detuning are spatially periodic. We analyze the geometric vector potential and the magnetic flux it generates, as well as the accompanying geometric scalar potential. We show how to understand the gauge-dependent Aharonov-Bohm singularities in the vector potential, and calculate the continuous magnetic flux through the elementary cell in terms of these singularities. The analysis is illustrated with a square optical flux lattice. We conclude with an explicit laser configuration yielding such a lattice using a set of five properly chosen beams with two counterpropagating pairs (one along the x axes and the other y axes), together with a single beam along the z axis. We show that this lattice is not phase-stable, and identify the one phase-difference that affects the magnetic flux. Thus armed with realistic laser setup, we directly compute the Chern number...

Juzeli?nas, G

2012-01-01T23:59:59.000Z

383

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

DOE Patents (OSTI)

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.

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

1999-01-01T23:59:59.000Z

384

Contact ORNL Neutron Sciences  

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

Sciences Organization Charts Neutron Sciences Directorate Associate Laboratory Director for Neutron Sciences, Kelly Beierschmitt Biology and Soft Matter Division Director, Paul...

385

NEUTRON SOURCE  

DOE Patents (OSTI)

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.

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

1959-01-13T23:59:59.000Z

386

NCNR thermal neutron prompt gamma facility  

Science Conference Proceedings (OSTI)

... plug, cube of lithium carbonate in resin surrounded by lead bricks, Aluminum cube (43 centimeters on a side) lithium 6 fluoride polymer plug ...

387

Spallation Neutron Source | ORNL Neutron Sciences  

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

Spallation Neutron Source SNS site, Spring 2012 The 80-acre SNS site is located on the east end of the ORNL campus and is about a three-minute drive from her sister neutron...

388

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

Science Conference Proceedings (OSTI)

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.

Dagrosa, Maria Alejandra, E-mail: dagrosa@cnea.gov.a [Department of Radiobiology, National Atomic Energy Commission, Buenos Aires (Argentina); National Research Council (Argentina); Crivello, Martin [Department of Radiobiology, National Atomic Energy Commission, Buenos Aires(Argentina); Perona, Marina [Department of Radiobiology, National Atomic Energy Commission, Buenos Aires (Argentina); National Research Council (Argentina); Thorp, Silvia; Santa Cruz, Gustavo Alberto [Department of Instrumentation and Control, National Atomic Energy Commission, Buenos Aires (Argentina); Pozzi, Emiliano [Argentina Reactor, National Atomic Energy Commission, Buenos Aires (Argentina); Casal, Mariana [Institute of Oncology 'Angel H. Roffo', University of Buenos Aires (Argentina); Thomasz, Lisa; Cabrini, Romulo [Department of Radiobiology, National Atomic Energy Commission, Buenos Aires (Argentina); Kahl, Steven [Department of Pharmaceutical Chemistry, University of California, San Francisco, CA (United States); Juvenal, Guillermo Juan [Department of Radiobiology, National Atomic Energy Commission, Buenos Aires (Argentina); National Research Council (Argentina); Pisarev, Mario Alberto [Department of Radiobiology, National Atomic Energy Commission, Buenos Aires (Argentina); National Research Council (Argentina); Department of Human Biochemistry, School of Medicine, University of Buenos Aires (Argentina)

2011-01-01T23:59:59.000Z

389

Science | ORNL Neutron Sciences  

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Neutron Science Neutron Science Neutron Scattering Science Neutrons are one of the fundamental particles that make up matter and have properties that make them ideal for certain types of research. In the universe, neutrons are abundant, making up more than half of all visible matter. Neutron scattering provides information about the positions, motions, and magnetic properties of solids. When a beam of neutrons is aimed at a sample, many neutrons will pass through the material. But some will interact directly with atomic nuclei and "bounce" away at an angle, like colliding balls in a game of pool. This behavior is called neutron diffraction, or neutron scattering. Using detectors, scientists can count scattered neutrons, measure their energies and the angles at which they scatter, and map their final position

390

Thermally Driven and Eddy-Driven Jet Variability in Reanalysis  

Science Conference Proceedings (OSTI)

Two important dynamical processes influence the extratropical zonal wind field: angular momentum transport by the thermally direct Hadley circulation (thermal-driving T) and momentum flux convergence by atmospheric waves (eddies) that develop in ...

Camille Li; Justin J. Wettstein

2012-03-01T23:59:59.000Z

391

DIFFERENTIAL SPECTRUM OF NEUTRONS AT CHACALTAYA-BOLIVIA  

SciTech Connect

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.

Mayta, R. [Carrera de Fisica, Universidad Mayor de San Andres, La Paz-Bolivia (Bolivia, Plurinational State of); Zanini, A. [Instituto Nazionale di Fisica Nucleare Sez. Torino Via P. Giuria n. 10125, Torino (Italy); Ticona, R.; Velarde, A. [Instituto de Investigaciones Fisicas, UMSA, La Paz-Bolivia (Bolivia, Plurinational State of)

2009-04-30T23:59:59.000Z

392

Mapping Heat Flux  

Science Conference Proceedings (OSTI)

An infrared camera technique designed for remote sensing of air–water heat flux has been developed. The technique uses the differential absorption of water between 3.817 and 4.514 microns. This difference causes each channel’s radiance to ...

Walt McKeown; Richard Leighton

1999-01-01T23:59:59.000Z

393

The theory of compact and efficient circular-pore MCP neutron collimators  

E-Print Network (OSTI)

centers, 3 mole % of nat Gd2O3, thermal neutrons of 25meV;centers, 3 mole % of nat Gd2O3, thermal neutrons of 25meV;0,1,2,3 mole % of nat Gd2O3. Fig. 7 .b L/D=250:1, 10um pores

Tremsin, A S; Feller, W B

2006-01-01T23:59:59.000Z

394

A Neutron Multiplicity Meter for Deep Underground Muon-Induced High Energy Neutron Measurements  

E-Print Network (OSTI)

We present the design of an instrument capable of measuring the high energy ($>$60 MeV) muon-induced neutron flux deep underground. The instrument is based on applying the Gd-loaded liquid-scintillator technique to measure the rate of high-energy neutrons underground based on the neutron multiplicity induced in a Pb target. We present design studies based on Monte Carlo simulations that show that an apparatus consisting of a Pb target of 200 cm by 200 cm area by 60 cm thickness covered by a 60 cm thick Gd-loaded liquid scintillator (0.5% Gd content) detector could measure, at a depth of 2000 meters of water equivalent, a rate of $70\\pm8$ (stat) events/year. Based on these studies, we also discuss the benefits of using a neutron multiplicity meter as a component of active shielding in such experiments.

R. Hennings-Yeomans; D. S. Akerib

2006-11-12T23:59:59.000Z

395

Downstream Heat Flux Profile vs. Midplane T Profile in Tokamaks  

SciTech Connect

The relationship between the midplane scrape-off-layer electron temperature profile and the parallel heat flux profile at the divertor in tokamaks is investigated. A model is applied which takes into account anisotropic thermal diffusion, in a rectilinear geometry with constant density. Eigenmode analysis is applied to the simplified problem with constant thermal diffusivities. A self-similar nonlinear solution is found for the more realistic problem with anisotropically temperature-dependent thermal diffusivities. Numerical solutions are developed for both cases, with spatially dependent heat flux emerging from the plasma. For both constant and temperature-dependent thermal diffusivities it is found that, below about one-half of its peak, the heat flux profile shape at the divertor, compared with the midplane temperature profile shape, is robustly described by the simplest two-point model. However the physical processes are not those assumed in the simplest two-point model, nor is the numerical coefficient relating q||div to Tmp ?||mp/L|| as predicted. For realistic parameters the peak in the heat flux, moreover, can be reduced by a factor of two or more from the two-point model scaling which fits the remaining profile. For temperature profiles in the SOL region above the x-point set by marginal stability, the heat flux profile to the divertor can be largely decoupled from the prediction of the two-point model. These results suggest caveats for data interpretation, and possibly favorable outcomes for divertor configurations with extended field lines.

Robert J. Goldston

2009-08-20T23:59:59.000Z

396

The resonance absorption probability function for neutron and multiplicative integral  

E-Print Network (OSTI)

The analytical approximations for the moderating neutrons flux density like Fermi spectra, widely used in reactor physics, involve the probability function for moderating neutron to avoid the resonant absorption obtained using some restrictive assumptions regarding the acceptable resonances width. By means of multiplicative integral (Volterra integral) theory for a commutative algebra an analytical expression for the probability function is obtained rigorously without any restrictive assumptions.

Rusov, V D; Kosenko, S I; Chernegenko, S A

2012-01-01T23:59:59.000Z

397

The resonance absorption probability function for neutron and multiplicative integral  

E-Print Network (OSTI)

The analytical approximations for the moderating neutrons flux density like Fermi spectra, widely used in reactor physics, involve the probability function for moderating neutron to avoid the resonant absorption obtained using some restrictive assumptions regarding the acceptable resonances width. By means of multiplicative integral (Volterra integral) theory for a commutative algebra an analytical expression for the probability function is obtained rigorously without any restrictive assumptions.

V. D. Rusov; V. A. Tarasov; S. I. Kosenko; S. A. Chernegenko

2012-08-05T23:59:59.000Z

398

NEUTRONIC REACTORS  

DOE Patents (OSTI)

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

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

1958-10-14T23:59:59.000Z

399

Determining Yankee Nuclear Power Station neutron activation  

Science Conference Proceedings (OSTI)

The Yankee nuclear power station located in Rowe, Massachusetts, permanently ceased power operations on February 26, 1992, after 31 yr of operation. Yankee has since initiated decommissioning planning activities. A significant component of these activities is a determination of the extent of radiological contamination of the Yankee site. Included in this effort was determination of the extent of neutron activation of plant components. This paper describes the determination of the neutron activation of the Yankee reactor vessel, associated internals, and surrounding structures. The Yankee reactor vessel is a 600-MW(thermal) stainless steel-lined, carbon steel vessel with stainless steel internal components designed by Westinghouse. The reactor vessel is surrounded and supported by a carbon steel neutron shield tank that was filled with chromated water during plant operation. A 5-ft-thick concrete biological shield wall surrounds the neutron shield tank. A project is under way to remove the reactor vessel internals from the reactor vessel.

Heider, K.J.; Morrissey, K.J. (Yankee Atomic Electric Co., Bolton, MA (United States))

1993-01-01T23:59:59.000Z

400

Neutron reflecting supermirror structure  

DOE Patents (OSTI)

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.

Wood, James L. (Drayton Plains, MI)

1992-01-01T23:59:59.000Z

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


401

Dim Isolated Neutron Stars, Cooling and Energy Dissipation  

E-Print Network (OSTI)

The cooling and reheating histories of dim isolated neutron stars(DINs) are discussed. Energy dissipation due to dipole spindown with ordinary and magnetar fields, and due to torques from a fallback disk are considered as alternative sources of reheating which would set the temperature of the neutron star after the initial cooling era. Cooling or thermal ages are related to the numbers and formation rates of the DINs and therefore to their relations with other isolated neutron star populations. Interaction with a fallback disk, higher multipole fields and activity of the neutron star are briefly discussed.

M. Ali Alpar

2006-09-07T23:59:59.000Z

402

ON THE NUMBER OF NEUTRONS FROM SOME FISSION FRAGMENTS  

SciTech Connect

The number of neutrons emitted by individual fragments from U/sup 235/ fission by thermal neutrons was measured using a large detector filled with a liquid organic cadmiumcontaining scintillator. The numbers of prompt neutrons were measured under 4 pi geometry conditions as a function of fragment mass. The excitation energy spent on prompt neutrons was derived on the basis of Weizsacker's semiempirical formula. A sharp asymmetry was noticed in the distribution of excitation energies between heavy and light fragments. The new data do not agree with the Fong statistical fission theory. (tr-auth)

Apalin, V.F.; Dobrynin, Yu.P.; Zakharova, V.P.; Kutikov, I.E.; Mikaelyan, L.A.

1960-01-01T23:59:59.000Z

403

The Versatile Neutron Imaging Instrument at SNS | ORNL Neutron...  

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

The Versatile Neutron Imaging Instrument at SNS VENUS: Neutron imaging to advance energy efficiency VENUS: Neutron imaging to advance energy efficiency. As its name indicates,...

404

Application of three-dimensional transport code to the analysis of the neutron streaming experiment  

Science Conference Proceedings (OSTI)

This paper summarized the calculational results of neutron streaming through a Clinch River Breeder Reactor (CRBR) Prototype coolant pipe chaseway. Particular emphasis is placed on results at bends in the chaseway. Calculations were performed with three three-dimensional codes: the discrete ordinates radiation transport code TORT and Monte Carlo radiation transport code MORSE, which were developed by Oak Ridge National Laboratory (ORNL), and the discrete ordinates code ENSEMBLE, which was developed in Japan. The purpose of the calculations is not only to compare the calculational results with the experimental results, but also to compare the results of TORT and MORSE with those of ENSEMBLE. In the TORT calculations, two types of difference methods, weighted-difference method was applied in ENSEMBLE calculation. Both TORT and ENSEMBLE produced nearly the same calculational results, but differed in the number of iterations required for converging each neutron group. Also, the two types of difference methods in the TORT calculations showed no appreciable variance in the number of iterations required. However, a noticeable disparity in the computer times and some variation in the calculational results did occur. The comparisons of the calculational results with the experimental results, showed for the epithermal neutron flux generally good agreement in the first and second legs and at the first bend where the two-dimensional modeling might be difficult. Results were fair to poor along the centerline of the first leg near the opening to the second leg because of discrete ordinates ray effects. Additionally, the agreement was good throughout the first and second legs for the thermal neutron region. Calculations with MORSE were made. These calculational results and comparisons are described also. 8 refs., 4 figs.

Chatani, K.; Slater, C.O.

1990-01-01T23:59:59.000Z

405

Imaging with Scattered Neutrons  

E-Print Network (OSTI)

We describe a novel experimental technique for neutron imaging with scattered neutrons. These scattered neutrons are of interest for condensed matter physics, because they permit to reveal the local distribution of incoherent and coherent scattering within a sample. In contrast to standard attenuation based imaging, scattered neutron imaging distinguishes between the scattering cross section and the total attenuation cross section including absorption. First successful low-noise millimeter-resolution images by scattered neutron radiography and tomography are presented.

H. Ballhausen; H. Abele; R. Gaehler; M. Trapp; A. Van Overberghe

2006-10-30T23:59:59.000Z

406

Recent Developments in Neutron Detection and Multiplicity Counting with Liquid Scintillator  

Science Conference Proceedings (OSTI)

For many years at LLNL we have been developing time-correlated neutron detection techniques and algorithms for many applications including Arms Control, Threat Detection and Nuclear Material Assaying. Many of our techniques have been developed specifically for relatively low efficiency (a few %) inherent in the man-portable systems. Historically we used thermal neutron detectors (mainly {sup 3}He) taking advantage of the high thermal neutron interaction cross-sections but more recently we have been investigating fast neutron detection with liquid scintillators and inorganic crystals. We have discovered considerable detection advantages with fast neutron detection as the inherent nano-second production time-scales of fission and neutron induced fission are preserved instead of being lost in neutron thermalization required for thermal neutron detectors. We are now applying fast neutron technology (new fast and portable digital electronics as well as new faster and less hazardous scintillator formulations) to the safeguards regime and faster detector response times and neutron momentum sensitivity show promise in measuring, differentiating and assaying samples that have very high count rates as well as mixed fission sources (e.g. Cm and Pu). We report on measured results with our existing liquid scintillator array and progress on design of nuclear material assaying system that incorporates fast neutron detection.

Nakae, L F; Kerr, P L; Newby, R J; Prasad, M K; Rowland, M S; Snyderman, N J; Verbeke, J M; Wurtz, R E

2010-01-07T23:59:59.000Z

407

METHOD AND APPARATUS FOR CONTROLLING NEUTRON DENSITY  

DOE Patents (OSTI)

A neutronic reactor comprising a moderator containing uniformly sized and spaced channels and uniformly dimensioned fuel elements is patented. The fuel elements have a fissionable core and an aluminum jacket. The cores and the jackets of the fuel elements in the central channels of the reactor are respectively thinner and thicker than the cores and jackets of the fuel elements in the remainder of the reactor, producing a flattened flux.

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

1961-06-27T23:59:59.000Z

408

News & Events | ORNL Neutron Sciences  

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

8 News 8 News Neutron Science In the News - 2008 November October September August July June May March February January Because some media sources archive past articles and require a subscription for access, some of the links below might not be active. If a citation listed here is no longer available, please contact the newspaper or your library directly. October Research Visits Just Budding at Spallation Neutron Source Knoxville News Sentinel 10/29 When the Spallation Neutron Source was in the proposal stage and under construction, its supporters said the $1.4 billion research complex would eventually attract about 2,000 scientists a year to Oak Ridge to perform experiments and otherwise do their thing. That number, as I recall, was lumped together with researchers at the recently upgraded High Flux Isotope

409

Biotechnology & Energy Highlights | Neutron Science | ORNL  

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

Biotechnology & Energy Biotechnology & Energy SHARE Biotechnology and Energy Highlights 1-10 of 10 Results Neutron Imaging Reveals Lithium Distribution in Lithium-Air Electrodes January 01, 2013 - Using neutron-computed tomography, researchers at the CG-1D neutron imaging instrument at Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR) have successfully mapped the three-dimensional spatial distribution of lithium products in electrochemically discharged lithium-air cathodes. Theory meets experiment: structure-property relationships in an electrode material for solid-oxide fuel cells December 01, 2012 - Fuel cell technology is one potentially very efficient and environmentally friendly way to convert the chemical energy of fuels into electricity. Solid-oxide fuel cells (SOFCs) can convert a

410

Time reversal invariance in polarized neutron decay  

SciTech Connect

An experiment to measure the time reversal invariance violating (T-violating) triple correlation (D) in the decay of free polarized neutrons has been developed. The detector design incorporates a detector geometry that provides a significant improvement in the sensitivity over that used in the most sensitive of previous experiments. A prototype detector was tested in measurements with a cold neutron beam. Data resulting from the tests are presented. A detailed calculation of systematic effects has been performed and new diagnostic techniques that allow these effects to be measured have been developed. As the result of this work, a new experiment is under way that will improve the sensitivity to D to 3 {times} 10{sup {minus}4} or better. With higher neutron flux a statistical sensitivity of the order 3 {times} 10{sup {minus}5} is ultimately expected. The decay of free polarized neutrons (n {yields} p + e + {bar v}{sub e}) is used to search for T-violation by measuring the triple correlation of the neutron spin polarization, and the electron and proton momenta ({sigma}{sub n} {center_dot} p{sub p} {times} p{sub e}). This correlation changes sign under reversal of the motion. Since final state effects in neutron decay are small, a nonzero coefficient, D, of this correlation indicates the violation of time reversal invariance. D is measured by comparing the numbers of coincidences in electron and proton detectors arranged symmetrically about a longitudinally polarized neutron beam. Particular care must be taken to eliminate residual asymmetries in the detectors or beam as these can lead to significant false effects. The Standard Model predicts negligible T-violating effects in neutron decay. Extensions to the Standard Model include new interactions some of which include CP-violating components. Some of these make first order contributions to D.

Wasserman, E.G.

1994-03-01T23:59:59.000Z

411

Neutron Science TeraGrid Gateway  

Science Conference Proceedings (OSTI)

The unique contributions of the Neutron Science TeraGrid Gateway (NSTG) are the connection of national user facility instrument data sources to the integrated cyberinfrastructure of the National Science FoundationTeraGrid and the development of a neutron science gateway that allows neutron scientists to use TeraGrid resources to analyze their data, including comparison of experiment with simulation. The NSTG is working in close collaboration with the Spallation Neutron Source (SNS) at Oak Ridge as their principal facility partner. The SNS is a next-generation neutron source. It has completed construction at a cost of $1.4 billion and is ramping up operations. The SNS will provide an order of magnitude greater flux than any previous facility in the world and will be available to all of the nation's scientists, independent of funding source, on a peer-reviewed merit basis. With this new capability, the neutron science community is facing orders of magnitude larger data sets and is at a critical point for data analysis and simulation. There is a recognized need for new ways to manage and analyze data to optimize both beam time and scientific output. The TeraGrid is providing new capabilities in the gateway for simulations using McStas and a fitting service on distributed TeraGrid resources to improved turnaround. NSTG staff are also exploring replicating experimental data in archival storage. As part of the SNS partnership, the NSTG provides access to gateway support, cyberinfrastructure outreach, community development, and user support for the neutron science community. This community includes not only SNS staff and users but extends to all the major worldwide neutron scattering centers.

Lynch, Vickie E [ORNL; Chen, Meili [ORNL; Cobb, John W [ORNL; Kohl, James Arthur [ORNL; Miller, Stephen D [ORNL; Speirs, David A [ORNL; Vazhkudai, Sudharshan S [ORNL

2010-01-01T23:59:59.000Z

412

Method and apparatus for determining vertical heat flux of geothermal field  

DOE Patents (OSTI)

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.

Poppendiek, Heinz F. (LaJolla, CA)

1982-01-01T23:59:59.000Z

413

NEUTRONIC REACTOR  

DOE Patents (OSTI)

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

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

1958-09-01T23:59:59.000Z

414

High Flux Beam Reactor | Environmental Restoration Projects | BNL  

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

Why is the High Flux Beam Reactor Being Decommissioned? Why is the High Flux Beam Reactor Being Decommissioned? HFBR The High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL) is being decommissioned because the Department of Energy (DOE) decided in 1999 that it would be permanently closed. The reactor was shut down in 1997 after tritium from a leak in the spent-fuel pool was found in the groundwater. The HFBR, which had operated from 1965 to 1996, was used solely for scientific research, providing neutrons for materials science, chemistry, biology, and physics experiments. The reactor was shut down for routine maintenance in November of 1996. In January 1997, tritium, a radioactive form of hydrogen and a by-product of reactor operations, was found in groundwater monitoring wells immediately south of the HFBR. The tritium

415

Neutron streak camera  

DOE Patents (OSTI)

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.

Wang, C.L.

1981-05-14T23:59:59.000Z

416

Organic metal neutron detector  

DOE Patents (OSTI)

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.

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

1984-11-21T23:59:59.000Z

417

Layered semiconductor neutron detectors  

SciTech Connect

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.

Mao, Samuel S; Perry, Dale L

2013-12-10T23:59:59.000Z

418

Porcelain enamel neutron absorbing material  

DOE Patents (OSTI)

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.

Iverson, D.C.

1987-11-20T23:59:59.000Z

419

MATERIALS FOR SPALLATION NEUTRON SOURCES: IV: Neutronics  

Science Conference Proceedings (OSTI)

The Department of Energy has initiated a pre-conceptual design study for the National Spallation Neutron Source (NSNS) and given preliminary approval for the ...

420

Neutronics-processing interface analyses for the Accelerator Transmutation of Waste (ATW) aqueous-based blanket system  

Science Conference Proceedings (OSTI)

Neutronics-processing interface parameters have large impacts on the neutron economy and transmutation performance of an aqueous-based Accelerator Transmutation of Waste (ATW) system. A detailed assessment of the interdependence of these blanket neutronic and chemical processing parameters has been performed. Neutronic performance analyses require that neutron transport calculations for the ATW blanket systems be fully coupled with the blanket processing and include all neutron absorptions in candidate waste nuclides as well as in fission and transmutation products. The effects of processing rates, flux levels, flux spectra, and external-to-blanket inventories on blanket neutronic performance were determined. In addition, the inventories and isotopics in the various subsystems were also calculated for various actinide and long-lived fission product transmutation strategies.

Davidson, J.W.; Battat, M.E.

1993-07-01T23:59:59.000Z

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421

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

Science Conference Proceedings (OSTI)

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.

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

2011-10-01T23:59:59.000Z

422

Design considerations for neutron activation and neutron source strength monitors for ITER  

SciTech Connect

The International Thermonuclear Experimental Reactor will require highly accurate measurements of fusion power production in time, space, and energy. Spectrometers in the neutron camera could do it all, but experience has taught us that multiple methods with redundancy and complementary uncertainties are needed. Previously, conceptual designs have been presented for time-integrated neutron activation and time-dependent neutron source strength monitors, both of which will be important parts of the integrated suite of neutron diagnostics for this purpose. The primary goals of the neutron activation system are: to maintain a robust relative measure of fusion energy production with stability and wide dynamic range; to enable an accurate absolute calibration of fusion power using neutronic techniques as successfully demonstrated on JET and TFTR; and to provide a flexible system for materials testing. The greatest difficulty is that the irradiation locations need to be close to plasma with a wide field of view. The routing of the pneumatic system is difficult because of minimum radius of curvature requirements and because of the careful need for containment of the tritium and activated air. The neutron source strength system needs to provide real-time source strength vs. time with {approximately}1 ms resolution and wide dynamic range in a robust and reliable manner with the capability to be absolutely calibrated by in-situ neutron sources as done on TFTR, JT-60U, and JET. In this paper a more detailed look at the expected neutron flux field around ITER is folded into a more complete design of the fission chamber system.

Barnes, C.W. [Los Alamos National Lab., NM (United States); Jassby, D.L.; LeMunyan, G.; Roquemore, A.L. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Walker, C. [ITER Joint Central Team, Garching (Germany)

1997-12-31T23:59:59.000Z

423

High flux reactor  

DOE Patents (OSTI)

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

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

1988-01-01T23:59:59.000Z

424

Magnetic and Electric Dipole Constraints on Extra Dimensions and Magnetic Fluxes  

E-Print Network (OSTI)

The propagation of charged particles and gauge fields in a compact extra dimension contributes to $g-2$ of the charged particles. In addition, a magnetic flux threading this extra dimension generates an electric dipole moment for these particles. We present constraints on the compactification size and on the possible magnetic flux imposed by the comparison of data and theory of the magnetic moment of the muon and from limits on the electric dipole moments of the muon, neutron and electron.

Aaron J. Roy; Myron Bander

2008-05-10T23:59:59.000Z

425

Time-parareal parallel in time integrator solver for time-dependent neutron diffusion equation  

E-Print Network (OSTI)

-dependent spatial flux distribution in nuclear reactor is required for nuclear safety and design. The motivation present a time-parallel algorithm that simulate the kinetic of neutron2 in a nuclear reactor. We consider engine behavior and in particular its energy4 production.5 The flux distribution on the nuclear reactor

426

Nodal weighting factor method for ex-core fast neutron fluence evaluation  

SciTech Connect

The nodal weighting factor method is developed for evaluating ex-core fast neutron flux in a nuclear reactor by utilizing adjoint neutron flux, a fictitious unit detector cross section for neutron energy above 1 or 0.1 MeV, the unit fission source, and relative assembly nodal powers. The method determines each nodal weighting factor for ex-core neutron fast flux evaluation by solving the steady-state adjoint neutron transport equation with a fictitious unit detector cross section for neutron energy above 1 or 0.1 MeV as the adjoint source, by integrating the unit fission source with a typical fission spectrum to the solved adjoint flux over all energies, all angles and given nodal volume, and by dividing it with the sum of all nodal weighting factors, which is a normalization factor. Then, the fast neutron flux can be obtained by summing the various relative nodal powers times the corresponding nodal weighting factors of the adjacent significantly contributed peripheral assembly nodes and times a proper fast neutron attenuation coefficient over an operating period. A generic set of nodal weighting factors can be used to evaluate neutron fluence at the same location for similar core design and fuel cycles, but the set of nodal weighting factors needs to be re-calibrated for a transition-fuel-cycle. This newly developed nodal weighting factor method should be a useful and simplified tool for evaluating fast neutron fluence at selected locations of interest in ex-core components of contemporary nuclear power reactors. (authors)

Chiang, R. T. [AREVA NP Inc., 6399 San Ignacio Ave., San Jose, CA 95119 (United States)

2012-07-01T23:59:59.000Z

427

Electroslag remelting with used fluxes  

Science Conference Proceedings (OSTI)

The Ukranian Scientific-Research Institute of Specialty Steel collaborated with plants engaged in the production of quality metals to introduce a low-waste electroslag remelting (ESR) technology employing used fluxes. It was established that the fluoride (type ANF-1) and fluoride-oxide (type ANF-6) fluxes which are widely used in ESR still have a high content of calcium fluoride and alumina and a low impurity content after 8-10 h of ESR. In the ESR of steels with used fluxes, the content of monitored components in the final slags changes negligibly, while the content of most impurities decreases. The used flux is also characterized by a low concentration of phosphorus and sulfur. It was found that flux can be used 3-5 times when it makes up 50% of the flux mixture in the charge. The savings realized from the use of spent flux in ESR amounts to 4-9 rubles/ton steel.