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Topic: Neutron transport

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	Neutron transport - Wikipedia, the free encyclopedia
		Beams of free neutrons are obtained by extracting neutrons from neutron sources.
		Neutrons are indeed reflected at a wall, but usually at an extremely low rate, which is a function of the wall's nature and surface quality, the neutron energy, the angle of incidence.
		The corresponding wavelength for the neutron is generally expressed in nanometers (nm) (or sometimes in angstroms, where 1 Å = 0.1 nm).
	en.wikipedia.org /wiki/Neutron_transport (340 words)

	Neutron - Wikipedia, the free encyclopedia
		The neutron and proton are instances of a nucleon.
		Although the neutron has zero net charge, it may interact electromagnetically in two ways: first, the neutron has a magnetic moment of the same order as the proton; second, it is composed of electrically charged quarks.
		One use of neutron emitters is the detection of light nuclei, particularly the hydrogen found in water molecules.
	en.wikipedia.org /wiki/Neutron (1233 words)

	Neutron transport with MCNP (Site not responding. Last check: 2007-10-09)
		Neutron histories in arbitrary geometries are tracked from the instant of creation by the use of evaluated cross sections data libraries for elastic scattering, inelastic scattering and absorption on nuclei present in specified (fixed) materials.
		The resulting collection of neutron tracks represents the neutron flux and can be folded track by track with reaction cross sections and heating functions in order to obtain estimates of transmutation rates and power densities, including statistical error estimations.
		The spallation neutron source is to be approximated with a point source in the centre of the target, emitting neutrons with an evaporation spectrum.
	www.neutron.kth.se /courses/transmutation/NeutronTransport/NeutronTransport.html (1752 words)

	Mohr and Associates, Inc. -- Limitations in Current Fissile Waste Nondestructive Assay Systems for Nuclear Criticality ... (Site not responding. Last check: 2007-10-09)
		Neutron detectors are used on the surface of the waste drum and one attempts to quantify/image the fissile contents of the drum with a pulsed (14 MeV) neutron source.
		Foreground neutrons are separated from background neutrons by subtracting the time-dependent die-away of the reflected neutron pulse.
		These results show that, for neutron transport process 1, the number of fissions produced in the drum can vary by about a factor of 20 depending on the location of the plutonium lump within the drum and the shielding of the waste in the drum.
	www.mohr-engineering.com /Publications/INEEL_ANPmethod_5_2000/ipan_system.htm (848 words)

	Sandia National Laboratories - Neutron generator model
		The neutron generator is Sandia¹s first new weapon component to enter the stockpile that must be qualified as radiation hardened in the absence of underground testing.
		As the part of the nuclear weapon that serves as the ³trigger,² the neutron generator produces a burst of neutrons, fragments of atomic nuclei.
		Because it uses short-lived tritium, the neutron generator is one of the components in a nuclear weapon that must be replaced regularly.
	www.sandia.gov /LabNews/LN01-28-00/neutron_story.html (848 words)

	PSU Neutron Beam Group (NBG)
		Neutron radiography and radioscopy are excellent non-intrusive techniques for visualization and quantification of the two-phase flow within the fuel cell in real time or steady-state.
		The results of velocity measurements indicate that the droplet velocity cannot be assumed to be on the same order of magnitude as the gas flow velocity for the fuel cell configuration and conditions tested, and a homogeneous, no-slip model of the two-phase channel flow is inappropriate for channel level two-phase modeling.
		Neutron radiography and radioscopy yield excellent spatial and temporal resolution for the investigation of water transport phenomenon and the measurement of liquid water inside an operating polymer electrolyte fuel cell.
	www.engr.psu.edu /nbg/nImagingH2ODistFuelCell.htm (1057 words)

	IAEA0929: RICANT, Neutron Transport in X-Y Geometry for Isotropic Scattering
		The transport equation is setup in each of these small regions.
		The outgoing neutron current at one surface is related to the incoming currents at the other surfaces of the region and the neutron sources inside the region.
		Neutron conservation is imposed for the calculation of transmission probabilities of current components and also in finding the average flux inside a region.
	www.nea.fr /abs/html/iaea0929.html (465 words)

	Nat' Academies Press, Status of the Dosimetry for the Radiation Effects Research Foundation (DS86) (2001)
		Later improvements in the delayed-neutron source and transport indicate that the contribution from delayed neutrons was significantly underestimated in DS86 (Egbert 1999).
		The revised delayed-neutron transport, which has not yet been implemented in DS86 (see Chapter 4), would tend to reduce the uncertainty associated with the delayed-neutron contribution to kerma, according to the observed improvement in the agreement between activation calculations and measurements at Nagasaki and for NTS tests of devices similar to the Nagasaki device.
		The preliminary uncertainty assessment in DS86 recognized the apparent discrepancy between the Hiroshima neutron calculations and measurements, and was one of the reasons for the decision to defer a final uncertainty assessment until its cause was resolved.
	www.nap.edu /books/0309075599/html/79.html (3897 words)

	Bulk Shielding Experiments at Apsara for Prototype Fast Breeder Reactor (Site not responding. Last check: 2007-10-09)
		Detailed incident neutron spectrum on the emergent face of CA was measured by unfolding the measured reaction rates of a large number of activation detectors.
		The calculated neutron spectrum on the emergent face of CA is compared with the measured neutron spectrum in Fig.2.
		The measured neutron spectrum incident on the emergent face of CA is close to the expected blanket leakage neutron spectrum of PFBR.
	www.igcar.ernet.in /lis/nl50/igc50-A1.htm (2372 words)

	Nuclear Engineering & Radiological Sciences :: Academics (Site not responding. Last check: 2007-10-09)
		Neutron cross sections, reaction rates, flux and current (9 h), neutron transport equation and simple solutions (8 h), diffusion theory (12 h), numerical solution of the diffusion equation (4 h), multigroup diffusion (3 h), point kinetics (8 h), neutron slowing down theory (8 h).
		Demonstrate a solid understanding of fundamental transport concepts such as angular flux, scalar flux, and current.
		Analytically solve problems in neutron transport and diffusion in both non-multiplying and multiplying media.
	www-ners.engin.umich.edu /academics/courses/ners441.shtml (289 words)

	CCS-4, Computer Graphics Contest Images (Site not responding. Last check: 2007-10-09)
		Colors on the horizontal plane are proportional to the transport solution (the total neutron flux) on the plane, red being the highest values and blue the lowest.
		The neutron transport calculation was performed using the massively parallel 3D even-parity transport code NIKE.
		The photon transport calculation was performed using the massively parallel 3D even-parity transport code NIKE, recently developed at LANL.
	www.ccs.lanl.gov /rad/htdocs/XTM/comgra.shtml (897 words)

	RAND \| Research Memoranda \| Energy-Dependent Neutron Transport Theory Near a Temperature Discontinuity.
		Energy-Dependent Neutron Transport Theory Near a Temperature Discontinuity.
		An exact solution is obtained for the energy-dependent Boltzmann transport equation for thermal neutrons near a temperature discontinuity.
		Numerical calculations of both scalar neutron flux and total neutron density are included for various temperature ratios and neutron-to-moderator mass ratios.
	www.rand.org /pubs/research_memoranda/RM5300 (313 words)

	Tim D. Bohm's Resume
		Fast neutron therapy sources are often produced by bombarding a Be target using protons or deuterons with an energy near 50 MeV and then shaping the resulting neutron beam with various filters and collimators.
		The particle transport codes LAHET and MCNP along with extended evaluated nuclear data were used to model the incident beam and target, as well as the neutron beam and collimators for three neutron therapy facilities.
		Determining the neutron exposure to the reactor vessel in a pressurized water reactor is important in determining the embrittlement of the vessel materials.
	www.medphysics.wisc.edu /~bohm/resume.html (2001 words)

	ANS : Publications : Journals : Nuclear Technology : Volume 148 : Deterministic and Monte Carlo Neutron Transport ...
		Neutron transport calculations have been performed for a fixed source using a spatially lumped fission neutron distribution, which has been derived from measurements.
		The deterministic code used is based on the finite element approximation to the multigroup second-order even-parity neutron transport equation, which is implemented in the EVENT code.
		We have compared neutron spectra at various locations not only to show differences between using multigroup deterministic and continuous energy (point nuclear data) Monte Carlo methods but also to assess neutron-induced activation levels calculated using the spectra obtained from both methods.
	www.ans.org /pubs/journals/nt/va-148-3-223-234 (302 words)

	ne155 \| Nuclear Engineering, UC, Berkeley
		Examples from neutron and photon transport; numerical solutions of neutron/photon diffusion and transport equations.
		Write discretized forms of neutron diffusion and transport equations in one-dimensional geometries, with full understanding of the discretization requirements for spatial, anglular, temporal, and energy variables.
		This course contributes to the NE program objectives by providing education in a fundamental area of numerical simulations of radiation transport which is important for a career in nuclear engineering.
	www.nuc.berkeley.edu /courses/classes/ne155.html (728 words)

	[No title] (Site not responding. Last check: 2007-10-09)
		For both models, in the absence of neutrino transport, the angle-averaged radial and angular convection velocities in the initial Ledoux unstable region below the shock after bounce achieve their peak values in ~20 ms, after which they decrease as the convection in this region dissipates.
		In the presence of neutrino transport, proto-neutron star convection velocities are too small relative to bulk inflow velocities to result in any significant convective transport of entropy and leptons.
		Moreover, when transport is included, the initial postbounce entropy gradient is smoothed out by neutrino diffusion, whereas the initial lepton gradient is maintained by electron capture and neutrino escape near the neutrinosphere.
	wonka.physics.ncsu.edu /Astro/papers/blo98c.html (546 words)

	Imaging passive and active neutron (IPAN) system -- Mohr Engineering (Site not responding. Last check: 2007-10-09)
		The transport and moderation of fission neutrons through the waste matrix to the surface neutron detector array.
		O content to 50% increases self-shielding effects for neutron transport process 1 to a factor of approximately 30.
		O, self-shielding effects for neutron transport process 2 increase to as much as a factor of approximately 65.
	www.mohr-engineering.com /INEEL_ANPmethod_5_2000/ipansystem.htm (816 words)

	COG: A New, High-Resolution Transport Code (Site not responding. Last check: 2007-10-09)
		The Boltzmann equation for the transport of particles such as neutrons is a conservation equation in a six-dimensional phase space.
		COG's neutron transport capabilities allow it to calculate the degree of criticality of an assembly of fissile materials-for example, the fuel rods of a nuclear reactor (Figure 2).
		This type of transport is central to the well-known EGS electron transport code,3 and we plan to adopt this model.
	www-phys.llnl.gov /N_Div/COG/ETR/ETR_9306.html (3756 words)

	Courses at department of Nuclear Engineering
		This course partly describes the basics of neutron reactions and neutron transport properties that are relevant for thermal fission reactors and partly the mathematical methods that are necessary to understand the principle and properties of a nuclear reactor.
		The aim of the course Transport Theory and Random Processes is to give an introduction to transport theory and fluctuations in transport processes, primarily in neutron transport.
		The theory of neutron flux fluctuations in power reactors, as well as their applications in reactor diagnostics is also included.
	www.nephy.chalmers.se /courses/courses.html (309 words)

	SD methods for the neutron transport equation
		We have conducted a study of the parallelization of numerical methods for the neutron transport equation .
		We have completed an implementation in a model case, where the SD method is used to solve the transport equation for each given velocity, i.e.
		We give a numerical example concerning neutron transport in 2D with initial angular flux equal to 1.0 in a disc of radius 0.25 for all angles, and with given source equal to zero.
	www.pdc.kth.se /info/reports/yr-johani/node15.html (255 words)

	A draft writeup for the ANS 19
		Transport theory calculations shall be performed via the "deterministic' discrete ordinates (SN) and/or the statistical Monte Carlo approaches as discussed in Sections 3.A and 3.B, respectively.
		and, is neutron flux, Q and S refer to source distributions, H is the core height, g is energy group index, and r,, and z correspond to radial, azimuthal, and axial positions, respectively.
		Prior to performing transport calculations for a particular facility, the calculational methodology shall be validated by comparing results with measurements made an a benchmark experiment, and demonstrating that it accurately determines appropriate benchmark dosimetry results.
	home.att.net /~rmrubin/ANS1910/transportdot/transportdot.htm (3194 words)

	Reactor Analysisand Engineering: Computer Codes: VIM
		Problem Solved: VIM solves the steady-state neutron or photon transport problem in any detailed three-dimensional geometry using either continuous energy-dependent ENDF nuclear data or multigroup cross sections.[1] Neutron transport is carried out in a criticality mode, or in a fixed source mode (optionally incorporating subcritical multiplication).
		The geometry options are infinite medium, combinatorial geometry,[2] and hexagonal or rectangular lattices of combinatorial geometry unit cells, and rectangular lattices of cells of assembled plates.
		The starting neutrons for each subsequent generation are randomly selected from the potential fission sites in the previous generation.
	www.rae.anl.gov /codes/vim (1139 words)

	[No title]
		It includes eigenvalue calculations of critical and subcritical systems as well as neutron transport calculations in fixed source mode or k-code mode to obtain reaction rates and energy deposition that are necessary for burnup calculations.
		Heating is calculated automatically on a similar way as the reaction rates during the neutron transport simulation by using heating cross section i.
		IV.B Neutron Source Specification Since MCB performs calculation in transport mode (with external source) or in k-code mode (with fission source) with one load of input file the user need to support neutron source specification for both cases.
	nucleartimes.jrc.nl /Doc/Jerzy_Paper.doc (5762 words)

	Abstracts/Summaries (Site not responding. Last check: 2007-10-09)
		Although the diffusion approximation for solving neutron transport problems have reached a high degree of accuracy at a quite modest computational cost, there are situations where it could be inherently unsuitable for calculating the distribution of particles.
		For such problems, the neutron transport (Boltzman) equation should be employed.
		Attempts have been made using asymptotic expansions for a systematic derivation of a nodal method with transport theory precision that is applicable to systems with heterogeneous nodes.
	www.me.gatech.edu /me/theses/neAbstracts.htm (1371 words)

	MC++: A Parallel, Portable, Monte Carlo Neutron Transport Code in C++ (Site not responding. Last check: 2007-10-09)
		MC++ is a multi-group Monte Carlo neutron transport code written in C++ and based on the Parallel Object-Oriented Methods and Applications (POOMA) class library.
		It is also intended to form the basis of the first transport physics framework (TPF), which is a C++ class library containing appropriate abstractions, objects, and methods for the particle transport problem.
		The transport problem is briefly described, as well as the current status and algorithms in MC++ for solving the transport equation.
	csdl2.computer.org /persagen/DLAbsToc.jsp?resourcePath=/dl/proceedings/&toc=comp/proceedings/ss/1997/7934/00/7934toc.xml&DOI=10.1109/SIMSYM.1997.586500 (287 words)

	Brian Triplett - UF Journal of Undergraduate Research Paper
		A transport calculation can necessitate large amounts of inspection time which may not be practical in some situations.
		Knowing the flux of neutrons is critical when determining how neutrons will interact with their surroundings.
		the source of neutrons originated from fission and fission depends on the neutron flux) iterations had to be performed.
	www.clas.ufl.edu /jur/200503/papers/paper_triplett.html (1447 words)

	Gerald C. Pomraning Publications List
		"On the Singular Eigenfunctions for Linear Transport in an Exponential Atmosphere," with E.W. Larsen and V.C. Badham, J.
		"On the Transport of Neutral Hydrogen Atoms in a Hydrogen Plasma," with R.D.M. Garcia and C.E. Siewert, Plasma Physics 24, 903 (1982).
		"Transport and Diffusion in a Statistical Medium," Transport Theory and Statistical Physics 5, 773 (1986).
	pages.prodigy.net /luciap/www/gcppubs.html (3646 words)

	Calculations of Neutron Transport Through a Simulated Waste Matrix (Site not responding. Last check: 2007-10-09)
		The differential die-away technique that is currently tried as a possible method for the assay, requires the incidence of a large fluence of thermal neutrons within the volume element that contains the fissile material.
		To study the neutron moderation and transport through nuclear waste matrices of different densities, the simulation of neutron transport was initiated using the Monte Carlo N- Particle (MCNP) code.
		The neutron source energy was varied between 10 keV and 15 MeV.
	flux.aps.org /meetings/BAPSMAY96/abs/S80009.html (233 words)

	RSICC CODE PACKAGE PSR-530
		Moreover, a binary file is optionally written with the density distribution of the different materials contained in the single mesh (for those meshes cutting at the same time more than one material zone), allowing a local density correction (per mesh) in alternative to a global density correction on the whole domain of the material zone.
		All the co-ordinate values that characterise the geometrical scheme at the basis of the 3D transport code geometrical and material model are read, sorted and all stored if different from the neighbouring ones more than an input tolerance established by the user.
		Fixed neutron sources, if any, are adapted to the mesh refinement at the same time.
	www-rsicc.ornl.gov /codes/psr/psr5/psr-530.html (2423 words)

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