Energy Technology Data Exchange (ETDEWEB)
Cullen, D; Latkowski, J; Sanz, J
1999-06-18
Recent modifications to the TART Monte Carlo neutron and photon transport code enable calculation of 566-group neutron spectra. This expanded group structure represents a significant improvement over the 50- and 175-group structures that have been previously available. To support use of this new capability, neutron activation cross section libraries have been created in the 175- and 566-group structures starting from the FENDL/A-2.0 pointwise data. Neutron spectra have been calculated for the first walls of the HYLIFE-II and SOMBRERO inertial fusion energy power plant designs and have been used in subsequent neutron activation calculations. The results obtained using the two different group structures are compared to each other as well as to those obtained using a 175-group version of the EAF3.1 activation cross section library.
Latkowski, J F; Sanz, J
2000-01-01
Recent modifications to the TART Monte Carlo neutron and photon transport code allow enable calculation of 566-group neutron spectra. This expanded group structure represents a significant improvement over the 50- and 175-group structures that have been previously available. To support use of this new capability, neutron activation cross-section libraries have been created in the 175- and 566-group structures starting from the FENDL/A-2.0 pointwise data. Neutron spectra have been calculated for the first walls of the HYLIFE-II and Sombrero inertial fusion energy power plant designs and have been used in subsequent neutron activation calculations. The results obtained using the two different group structures are compared with each other as well as to those obtained using a 175-group version of the EAF3.1 activation cross-section library.
Neutron transport simulation (selected topics)
Energy Technology Data Exchange (ETDEWEB)
Vaz, P. [Instituto Tecnologico e Nuclear, Estrada Nacional 10, P-2686-953 Sacavem (Portugal)], E-mail: pedrovaz@itn.pt
2009-10-15
Neutron transport simulation is usually performed for criticality, power distribution, activation, scattering, dosimetry and shielding problems, among others. During the last fifteen years, innovative technological applications have been proposed (Accelerator Driven Systems, Energy Amplifiers, Spallation Neutron Sources, etc.), involving the utilization of intermediate energies (hundreds of MeV) and high-intensity (tens of mA) proton accelerators impinging in targets of high Z elements. Additionally, the use of protons, neutrons and light ions for medical applications (hadrontherapy) impose requirements on neutron dosimetry-related quantities (such as kerma factors) for biologically relevant materials, in the energy range starting at several tens of MeV. Shielding and activation related problems associated to the operation of high-energy proton accelerators, emerging space-related applications and aircrew dosimetry-related topics are also fields of intense activity requiring as accurate as possible medium- and high-energy neutron (and other hadrons) transport simulation. These applications impose specific requirements on cross-section data for structural materials, targets, actinides and biologically relevant materials. Emerging nuclear energy systems and next generation nuclear reactors also impose requirements on accurate neutron transport calculations and on cross-section data needs for structural materials, coolants and nuclear fuel materials, aiming at improved safety and detailed thermal-hydraulics and radiation damage studies. In this review paper, the state-of-the-art in the computational tools and methodologies available to perform neutron transport simulation is presented. Proton- and neutron-induced cross-section data needs and requirements are discussed. Hot topics are pinpointed, prospective views are provided and future trends identified.
Neutron transport simulation (selected topics)
Vaz, P.
2009-10-01
Neutron transport simulation is usually performed for criticality, power distribution, activation, scattering, dosimetry and shielding problems, among others. During the last fifteen years, innovative technological applications have been proposed (Accelerator Driven Systems, Energy Amplifiers, Spallation Neutron Sources, etc.), involving the utilization of intermediate energies (hundreds of MeV) and high-intensity (tens of mA) proton accelerators impinging in targets of high Z elements. Additionally, the use of protons, neutrons and light ions for medical applications (hadrontherapy) impose requirements on neutron dosimetry-related quantities (such as kerma factors) for biologically relevant materials, in the energy range starting at several tens of MeV. Shielding and activation related problems associated to the operation of high-energy proton accelerators, emerging space-related applications and aircrew dosimetry-related topics are also fields of intense activity requiring as accurate as possible medium- and high-energy neutron (and other hadrons) transport simulation. These applications impose specific requirements on cross-section data for structural materials, targets, actinides and biologically relevant materials. Emerging nuclear energy systems and next generation nuclear reactors also impose requirements on accurate neutron transport calculations and on cross-section data needs for structural materials, coolants and nuclear fuel materials, aiming at improved safety and detailed thermal-hydraulics and radiation damage studies. In this review paper, the state-of-the-art in the computational tools and methodologies available to perform neutron transport simulation is presented. Proton- and neutron-induced cross-section data needs and requirements are discussed. Hot topics are pinpointed, prospective views are provided and future trends identified.
Influence of the neutron transport tube on neutron resonance densitometry
Directory of Open Access Journals (Sweden)
Kitatani Fumito
2017-01-01
Full Text Available Neutron Resonance Densitometry (NRD is a non-destructive assay technique of nuclear materials in particle-like debris that contains various materials. An aim of NRD is to quantify nuclear materials in a melting fuel of Fukusima Daiichi plant, spent nuclear fuel and annihilation disposal fuel etc. NRD consists of two techniques of Neutron Resonance Transmission Analysis (NRTA and Neutron Resonance Capture Analysis (NRCA or Prompt Gamma-ray Analysis (PGA. A density of nuclear material isotopes is decided with NRTA. The materials absorbing a neutron in a wide energy range such as boron in a sample are identified by NRCA/PGA. The information of NRCA/PGA is used in NRTA analysis to quantify nuclear material isotopes. A neutron time of flight (TOF method is used in NRD measurements. A facility, consisting of a neutron source, a neutron flight path, and a detector is required. A short flight path and a strong neutron source are needed to downsize such a facility and put NRD into practical use. A neutron transport tube covers a flight path to prevent noises. In order to investigate the effect of neutron transport tube and pulse width of a neutron source, we carried out NRTA experiments with a 2-m short neutron transport tube constructed at Kyoto University Research Reactor Institute - Linear Accelerator (KURRI-LINAC, and impacts of shield of neutron transport tube and influence of pulse width of a neutron source were examined. A shield of the neutron transport tube reduced a background and had a good influence on the measurement. The resonance dips of 183W at 27 eV was successfully observed with a pulse width of a neutron source less than 2 μs.
Neutron Transport Simulations for NIST Neutron Lifetime Experiment
Li, Fangchen; BL2 Collaboration Collaboration
2016-09-01
Neutrons in stable nuclei can exist forever; a free neutron lasts for about 15 minutes on average before it beta decays to a proton, an electron, and an antineutrino. Precision measurements of the neutron lifetime test the validity of weak interaction theory and provide input into the theory of the evolution of light elements in the early universe. There are two predominant ways of measuring the neutron lifetime: the bottle method and the beam method. The bottle method measures decays of ultracold neutrons that are stored in a bottle. The beam method measures decay protons in a beam of cold neutrons of known flux. An improved beam experiment is being prepared at the National Institute of Science and Technology (Gaithersburg, MD) with the goal of reducing statistical and systematic uncertainties to the level of 1 s. The purpose of my studies was to develop computer simulations of neutron transport to determine the beam collimation and study the neutron distribution's effect on systematic effects for the experiment, such as the solid angle of the neutron flux monitor. The motivation for the experiment and the results of this work will be presented. This work was supported, in part, by a Grant to Gettysburg College from the Howard Hughes Medical Institute through the Precollege and Undergraduate Science Education Program.
Exact solution of the neutron transport equation in spherical geometry
Energy Technology Data Exchange (ETDEWEB)
Anli, Fikret; Akkurt, Abdullah; Yildirim, Hueseyin; Ates, Kemal [Kahramanmaras Suetcue Imam Univ. (Turkey). Faculty of Sciences and Letters
2017-03-15
Solution of the neutron transport equation in one dimensional slab geometry construct a basis for the solution of neutron transport equation in a curvilinear geometry. Therefore, in this work, we attempt to derive an exact analytical benchmark solution for both neutron transport equations in slab and spherical medium by using P{sub N} approximation which is widely used in neutron transport theory.
Parallel Deterministic Neutron Transport with AMR
Energy Technology Data Exchange (ETDEWEB)
Clouse, C
2005-03-25
AMTRAN, a one, two and three dimensional Sn neutron transport code with adaptive mesh refinement (AMR) has been parallelized with MPI over spatial domains and energy groups and with threads over angles. Block refined AMR is used with linear finite element representations for the fluxes, which are node centered. AMR requirements are determined by minimum mean free path calculations throughout the problem and can provide an order of magnitude or more reduction in zoning requirements for the same level of accuracy, compared to a uniformly zoned problem.
Maldonado-Velázquez, M.; Barrón-Palos, L.; Crawford, C.; Snow, W. M.
2017-05-01
The neutron spin is a critical degree of freedom for many precision measurements using low-energy neutrons. Fundamental symmetries and interactions can be studied using polarized neutrons. Parity-violation (PV) in the hadronic weak interaction and the search for exotic forces that depend on the relative spin and velocity, are two questions of fundamental physics that can be studied via the neutron spin rotations that arise from the interaction of polarized cold neutrons and unpolarized matter. The Neutron Spin Rotation (NSR) collaboration developed a neutron polarimeter, capable of determining neutron spin rotations of the order of 10-7 rad per meter of traversed material. This paper describes two key components of the NSR apparatus, responsible for the transport and manipulation of the spin of the neutrons before and after the target region, which is surrounded by magnetic shielding and where residual magnetic fields need to be below 100 μG. These magnetic field devices, called input and output coils, provide the magnetic field for adiabatic transport of the neutron spin in the regions outside the magnetic shielding while producing a sharp nonadiabatic transition of the neutron spin when entering/exiting the low-magnetic-field region. In addition, the coils are self contained, forcing the return magnetic flux into a compact region of space to minimize fringe fields outside. The design of the input and output coils is based on the magnetic scalar potential method.
Design of a transportable high efficiency fast neutron spectrometer
Energy Technology Data Exchange (ETDEWEB)
Roecker, C., E-mail: calebroecker@berkeley.edu [Department of Nuclear Engineering, University of California at Berkeley, CA 94720 (United States); Bernstein, A.; Bowden, N.S. [Nuclear and Chemical Sciences Division, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Cabrera-Palmer, B. [Radiation and Nuclear Detection Systems, Sandia National Laboratories, Livermore, CA 94550 (United States); Dazeley, S. [Nuclear and Chemical Sciences Division, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Gerling, M.; Marleau, P.; Sweany, M.D. [Radiation and Nuclear Detection Systems, Sandia National Laboratories, Livermore, CA 94550 (United States); Vetter, K. [Department of Nuclear Engineering, University of California at Berkeley, CA 94720 (United States); Nuclear Science Division, Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States)
2016-08-01
A transportable fast neutron detection system has been designed and constructed for measuring neutron energy spectra and flux ranging from tens to hundreds of MeV. The transportability of the spectrometer reduces the detector-related systematic bias between different neutron spectra and flux measurements, which allows for the comparison of measurements above or below ground. The spectrometer will measure neutron fluxes that are of prohibitively low intensity compared to the site-specific background rates targeted by other transportable fast neutron detection systems. To measure low intensity high-energy neutron fluxes, a conventional capture-gating technique is used for measuring neutron energies above 20 MeV and a novel multiplicity technique is used for measuring neutron energies above 100 MeV. The spectrometer is composed of two Gd containing plastic scintillator detectors arranged around a lead spallation target. To calibrate and characterize the position dependent response of the spectrometer, a Monte Carlo model was developed and used in conjunction with experimental data from gamma ray sources. Multiplicity event identification algorithms were developed and used with a Cf-252 neutron multiplicity source to validate the Monte Carlo model Gd concentration and secondary neutron capture efficiency. The validated Monte Carlo model was used to predict an effective area for the multiplicity and capture gating analyses. For incident neutron energies between 100 MeV and 1000 MeV with an isotropic angular distribution, the multiplicity analysis predicted an effective area of 500 cm{sup 2} rising to 5000 cm{sup 2}. For neutron energies above 20 MeV, the capture-gating analysis predicted an effective area between 1800 cm{sup 2} and 2500 cm{sup 2}. The multiplicity mode was found to be sensitive to the incident neutron angular distribution.
UPWIND DISCONTINUOUS GALERKIN METHODS FOR TWO DIMENSIONAL NEUTRON TRANSPORT EQUATIONS
Institute of Scientific and Technical Information of China (English)
袁光伟; 沈智军; 闫伟
2003-01-01
In this paper the upwind discontinuous Galerkin methods with triangle meshes for two dimensional neutron transport equations will be studied.The stability for both of the semi-discrete and full-discrete method will be proved.
Scattered Neutron Tomography Based on A Neutron Transport Inverse Problem
Energy Technology Data Exchange (ETDEWEB)
William Charlton
2007-07-01
Neutron radiography and computed tomography are commonly used techniques to non-destructively examine materials. Tomography refers to the cross-sectional imaging of an object from either transmission or reflection data collected by illuminating the object from many different directions.
Neutron transport with anisotropic scattering: theory and applications
Van den Eynde, Gert
2005-01-01
This thesis is a blend of neutron transport theory and numerical analysis. We start with the study of the problem of the Mika/Case eigenexpansion used in the solution process of the homogeneous one-speed Boltzmann neutron transport equation with anisotropic scattering for plane symmetry. The anisotropic scattering is expressed as a finite Legendre series in which the coefficients are the ``scattering coefficients'. This eigenexpansion consists of a discrete spectrum of eigenvalues with its co...
Solar Neutron Transport in the Earth's Atmosphere
Valdes-Galicia, J. F.; Dorman, L. I.; Dorman, I. V.
1998-11-01
We present results of a numerical simulation and analytical solution of small scale neutron multi-scattering and attenuation in the earth atmosphere. A range of initial zenith angles and different atmpspheric depths are considered. We show that the angular distribution of neutrons remains symetrycal only for vertical arrival. For inclined arrival the distribution becomes asymetrical; the asymmetry grows with increasing zenith angle. This effect is caused by the stronger attenuation of neutrons scattered to zenith angles larger than the arrival angle. Our analytical solution shows reasonable coincidence with the numerical simulation results. These solutions are able to reproduce the normalised observed counting rates of neutron monitors for the event of 24 may 1990, the largest Solar Neutron event observed on Earth.
Transport coefficients in superfluid neutron stars
Energy Technology Data Exchange (ETDEWEB)
Tolos, Laura [Instituto de Ciencias del Espacio (IEEC/CSIC) Campus Universitat Autònoma de Barcelona, Facultat de Ciències, Torre C5, E-08193 Bellaterra (Barcelona) (Spain); Frankfurt Institute for Advances Studies. Johann Wolfgang Goethe University, Ruth-Moufang-Str. 1, 60438 Frankfurt am Main (Germany); Manuel, Cristina [Instituto de Ciencias del Espacio (IEEC/CSIC) Campus Universitat Autònoma de Barcelona, Facultat de Ciències, Torre C5, E-08193 Bellaterra (Barcelona) (Spain); Sarkar, Sreemoyee [Tata Institute of Fundamental Research, Homi Bhaba Road, Mumbai-400005 (India); Tarrus, Jaume [Physik Department, Technische Universität München, D-85748 Garching (Germany)
2016-01-22
We study the shear and bulk viscosity coefficients as well as the thermal conductivity as arising from the collisions among phonons in superfluid neutron stars. We use effective field theory techniques to extract the allowed phonon collisional processes, written as a function of the equation of state and the gap of the system. The shear viscosity due to phonon scattering is compared to calculations of that coming from electron collisions. We also comment on the possible consequences for r-mode damping in superfluid neutron stars. Moreover, we find that phonon collisions give the leading contribution to the bulk viscosities in the core of the neutron stars. We finally obtain a temperature-independent thermal conductivity from phonon collisions and compare it with the electron-muon thermal conductivity in superfluid neutron stars.
Transport coefficients in superfluid neutron stars
Tolos, Laura; Sarkar, Sreemoyee; Tarrus, Jaume
2014-01-01
We study the shear and bulk viscosity coefficients as well as the thermal conductivity as arising from the collisions among phonons in superfluid neutron stars. We use effective field theory techniques to extract the allowed phonon collisional processes, written as a function of the equation of state and the gap of the system. The shear viscosity due to phonon scattering is compared to calculations of that coming from electron collisions. We also comment on the possible consequences for r-mode damping in superfluid neutron stars. Moreover, we find that phonon collisions give the leading contribution to the bulk viscosities in the core of the neutron stars. We finally obtain a temperature-independent thermal conductivity from phonon collisions and compare it with the electron-muon thermal conductivity in superfluid neutron stars.
Neutron transport study of a beam port based dynamic neutron radiography facility
Khaial, Anas M.
Neutron radiography has the ability to differentiate between gas and liquid in two-phase flow due both to the density difference and the high neutron scattering probability of hydrogen. Previous studies have used dynamic neutron radiography -- in both real-time and high-speed -- for air-water, steam-water and gas-liquid metal two-phase flow measurements. Radiography with thermal neutrons is straightforward and efficient as thermal neutrons are easier to detect with relatively higher efficiency and can be easily extracted from nuclear reactor beam ports. The quality of images obtained using neutron radiography and the imaging speed depend on the neutron beam intensity at the imaging plane. A high quality neutron beam, with thermal neutron intensity greater than 3.0x 10 6 n/cm2-s and a collimation ratio greater than 100 at the imaging plane, is required for effective dynamic neutron radiography up to 2000 frames per second. The primary objectives of this work are: (1) to optimize a neutron radiography facility for dynamic neutron radiography applications and (2) to investigate a new technique for three-dimensional neutron radiography using information obtained from neutron scattering. In this work, neutron transport analysis and experimental validation of a dynamic neutron radiography facility is studied with consideration of real-time and high-speed neutron radiography requirements. A beam port based dynamic neutron radiography facility, for a target thermal neutron flux of 1.0x107 n/cm2-s, has been analyzed, constructed and experimentally verified at the McMaster Nuclear Reactor. The neutron source strength at the beam tube entrance is evaluated experimentally by measuring the thermal and fast neutron fluxes using copper activation flux-mapping technique. The development of different facility components, such as beam tube liner, gamma ray filter, beam shutter and biological shield, is achieved analytically using neutron attenuation and divergence theories. Monte
On generating neutron transport tables with the NJOY system
Energy Technology Data Exchange (ETDEWEB)
Caldeira, Alexandre D.; Claro, Luiz H., E-mail: alexdc@ieav.cta.br, E-mail: luizhenu@ieav.cta.br [Instituto de Estudos Avancados (IEAv), Sao Jose dos Campos, SP (Brazil)
2013-07-01
Incorrect values for the product of the average number of neutrons released per fission and the fission microscopic cross-section were detected in several energy groups of a neutron transport table generated with the most updated version of the NJOY system. It was verified that the problem persists when older versions of this system are utilized. Although this problem exists for, at least, ten years, it is still an open question. (author)
FLUKA simulations of neutron transport in the Dresden Felsenkeller
Energy Technology Data Exchange (ETDEWEB)
Grieger, Marcel [Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Dresden (Germany); Technische Universitaet Dresden (Germany); Bemmerer, Daniel; Mueller, Stefan E.; Szuecs, Tamas [Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Dresden (Germany); Zuber, Kai [Technische Universitaet Dresden (Germany)
2015-07-01
A new underground ion accelerator with 5 MV acceleration potential is currently being prepared for installation in the Dresden Felsenkeller. The Felsenkeller site consists of altogether nine mutually connected tunnels. It is shielded from cosmic radiation by a 45 m thick rock overburden, enabling uniquely sensitive experiments. In order to exclude any possible effect by the new accelerator in tunnel VIII on the existing low-background γ-counting facility in tunnel IV, Monte Carlo simulations of neutron transport are being performed. A realistic neutron source field is developed, and the resulting additional neutron flux at the γ-counting facility is modeled by FLUKA simulations.
Considerations in the design of an improved transportable neutron spectrometer
Williams, A M; Brushwood, J M; Beeley, P A
2002-01-01
The Transportable Neutron Spectrometer (TNS) has been used by the Ministry of Defence for over 15 years to characterise neutron fields in workplace environments and provide local correction factors for both area and personal dosimeters. In light of advances in neutron spectrometry, a programme to evaluate and improve TNS has been initiated. This paper describes TNS, presents its operation in known radioisotope fields and in a reactor environment. Deficiencies in the operation of the instrument are highlighted, together with proposals for updating the response functions and spectrum unfolding methodologies.
Calculated characteristics of subcritical assembly with anisotropic transport of neutrons
Energy Technology Data Exchange (ETDEWEB)
Gorin, N.V.; Lipilina, E.N.; Lyutov, V.D.; Saukov, A.I. [Zababakhin Russian Federal Nuclear Center - All-Russian Scientific Researching Institute of Technical Physics (Russian Federation)
2003-07-01
There was considered possibility of creating enough sub-critical system that multiply neutron fluence from a primary source by many orders. For assemblies with high neutron tie between parts, it is impossible. That is why there was developed a construction consisting of many units (cascades) having weak feedback with preceding cascades. The feedback attenuation was obtained placing layers of slow neutron absorber and moderators between the cascades of fission material. Anisotropy of fast neutron transport through the layers was used. The system consisted of many identical cascades aligning one by another. Each cascade consists of layers of moderator, fissile material and absorber of slow neutrons. The calculations were carried out using the code MCNP.4a with nuclear data library ENDF/B5. In this construction neutrons spread predominantly in one direction multiplying in each next fissile layer, and they attenuate considerably in the opposite direction. In a calculated construction, multiplication factor of one cascade is about 1.5 and multiplication factor of whole construction composed of n cascades is 1.5{sup n}. Calculated keff value is 0.9 for one cascade and does not exceed 0.98 for a system containing any number of cascades. Therefore the assembly is always sub-critical and therefore it is safe in respect of criticality. There was considered using such a sub-critical assembly to create a powerful neutron fluence for neutron boron-capturing therapy. The system merits and demerits were discussed. (authors)
Neutron Transport Associated with the Galactic Cosmic Ray Cascade
Singleterry, Robert Clay, Jr.
Transport of low energy neutrons associated with the galactic cosmic ray cascade is analyzed in this dissertation. A benchmark quality analytical algorithm is demonstrated for use with B scRYNTRN, a computer program written by the High Energy Physics Division of N scASA Langley Research Center, which is used to design and analyze shielding against the radiation created by the cascade. B scRYNTRN uses numerical methods to solve the integral transport equations for baryons with the straight-ahead approximation, and numerical and empirical methods to generate the interaction probabilities. The straight-ahead approximation is adequate for charged particles, but not for neutrons. As N scASA Langley improves B scRYNTRN to include low energy neutrons, a benchmark quality solution is needed for comparison. The neutron transport algorithm demonstrated in this dissertation uses the closed-form Green's function solution to the galactic cosmic ray cascade transport equations to generate a source of neutrons. A basis function expansion for finite heterogeneous and semi-infinite homogeneous slabs with multiple energy groups and isotropic scattering is used to generate neutron fluxes resulting from the cascade. This method, called the F_{rm N} method, is used to solve the neutral particle linear Boltzmann transport equation. As a demonstration of the algorithm coded in the programs M scGSLAB and M scGSEMI, neutron and ion fluxes are shown for a beam of fluorine ions at 1000 MeV per nucleon incident on semi-infinite and finite aluminum slabs. Also, to demonstrate that the shielding effectiveness against the radiation from the galactic cosmic ray cascade is not directly proportional to shield thickness, a graph of transmitted total neutron scalar flux versus slab thickness is shown. A simple model based on the nuclear liquid drop assumption is used to generate cross sections for the galactic cosmic ray cascade. The E scNDF/B V database is used to generate the total and scattering
TRIPOLI-3: a neutron/photon Monte Carlo transport code
Energy Technology Data Exchange (ETDEWEB)
Nimal, J.C.; Vergnaud, T. [Commissariat a l' Energie Atomique, Gif-sur-Yvette (France). Service d' Etudes de Reacteurs et de Mathematiques Appliquees
2001-07-01
The present version of TRIPOLI-3 solves the transport equation for coupled neutron and gamma ray problems in three dimensional geometries by using the Monte Carlo method. This code is devoted both to shielding and criticality problems. The most important feature for particle transport equation solving is the fine treatment of the physical phenomena and sophisticated biasing technics useful for deep penetrations. The code is used either for shielding design studies or for reference and benchmark to validate cross sections. Neutronic studies are essentially cell or small core calculations and criticality problems. TRIPOLI-3 has been used as reference method, for example, for resonance self shielding qualification. (orig.)
Implementation of the quasi-static method for neutron transport
Energy Technology Data Exchange (ETDEWEB)
Alcaro, Fabio; Dulla, Sandra; Ravetto, Piero, E-mail: fabio.alcaro@polito.it, E-mail: sandra.dulla@polito.it, E-mail: piero.ravetto@polito.it [Dipartimento di Energetica, Politecnico di Torino (Italy); Le Tellier, Romain; Suteau, Christophe, E-mail: romain.le-tellier@cea.fr, E-mail: christophe.suteau@cea.fr [CEA, DEN, DER/SPRC/LEPh, Cadarache, Saint Paul-lez-Durance (France)
2011-07-01
The study of the dynamic behavior of next generation nuclear reactors is a fundamental aspect for safety and reliability assessments. Despite the growing performances of modern computers, the full solution of the neutron Boltzmann equation in the time domain is still an impracticable task, thus several approximate dynamic models have been proposed for the simulation of nuclear reactor transients; the quasi-static method represents the standard tool currently adopted for the space-time solution of neutron transport problems. All the practical applications of this method that have been proposed contain a major limit, consisting in the use of isotropic quantities, such as scalar fluxes and isotropic external neutron sources, being the only data structures available in most deterministic transport codes. The loss of the angular information produces both inaccuracies in the solution of the kinetic model and the inconsistency of the quasi-static method itself. The present paper is devoted to the implementation of a consistent quasi-static method. The computational platform developed by CEA in Cadarache has been used for the creation of a kinetic package to be coupled with the existing SNATCH solver, a discrete-ordinate multi-dimensional neutron transport solver, employed for the solution of the steady-state Boltzmann equation. The work aims at highlighting the effects of the angular treatment of the neutron flux on the transient analysis, comparing the results with those produced by the previous implementations of the quasi-static method. (author)
STABILITY OF P2 METHODS FOR NEUTRON TRANSPORT EQUATIONS
Institute of Scientific and Technical Information of China (English)
袁光伟; 沈智军; 沈隆钧; 周毓麟
2002-01-01
In this paper the P2 approximation to the one-group planar neutron transport theory is discussed. The stability of the solutions for P2 equations with general boundary conditions, including the Marshak boundary condition, is proved. Moreover,the stability of the up-wind difference scheme for the P2 equation is demonstrated.
Graphical User Interface for Simplified Neutron Transport Calculations
Energy Technology Data Exchange (ETDEWEB)
Schwarz, Randolph; Carter, Leland L
2011-07-18
A number of codes perform simple photon physics calculations. The nuclear industry is lacking in similar tools to perform simplified neutron physics shielding calculations. With the increased importance of performing neutron calculations for homeland security applications and defense nuclear nonproliferation tasks, having an efficient method for performing simple neutron transport calculations becomes increasingly important. Codes such as Monte Carlo N-particle (MCNP) can perform the transport calculations; however, the technical details in setting up, running, and interpreting the required simulations are quite complex and typically go beyond the abilities of most users who need a simple answer to a neutron transport calculation. The work documented in this report resulted in the development of the NucWiz program, which can create an MCNP input file for a set of simple geometries, source, and detector configurations. The user selects source, shield, and tally configurations from a set of pre-defined lists, and the software creates a complete MCNP input file that can be optionally run and the results viewed inside NucWiz.
Neutron shielding evaluation for a small fuel transport case
Coeck, M; Vanhavere, F
2002-01-01
We investigated the effectiveness of a small neutron shield configuration for the transportation of fresh MOX fuel rods in an experimental facility, this in order to reduce the dose received by the personnel. Monte Carlo simulations using the Tripoli and MCNP4B code were applied. Different configurations were studied, starting from the bare fuel rod positioned on an iron plate up to a fuel rod covered by a box-shaped shield made of different materials such as polyethylene, polyethylene with boron and polyethylene with a cadmium layer. We compared the neutron spectra for the different cases and calculated the corresponding ambient equivalent dose rate H*(10).
Development of Library Processing System for Neutron Transport Calculation
Energy Technology Data Exchange (ETDEWEB)
Song, J. S.; Park, S. Y.; Kim, H. Y. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)] (and others)
2008-12-15
A system for library generation was developed for the lattice neutron transport program for pressurized water reactor core analysis. The system extracts multi energy group nuclear data for requested nuclides from ENDF/B whose data are based on continuous energy, generates hydrogen equivalent factor and resonance integral table as functions of temperature and background cross section for resonance nuclides, generates subgroup data for the lattice program to treat resonance exactly as possible, and generates multi-group neutron library file including nuclide depletion data for use of the lattice program.
Recent improvements to a transportable neutron spectrometer (TNS)
Weaver, J A; Peyton, A J; Roskell, J
2002-01-01
This paper reviews the design, operation and future development of a transportable neutron spectrometer (TNS). Analogue signal processing techniques are used to condition the signals from an array of radiation sensors, comprising five gas-filled sensors and a hydrogenous oil-filled scintillator. This high reliance on analogue signal processing techniques is because of the nano-second rise time of the pulses produced from the sensor array. The analogue circuitry requires a high degree of expertise from the operator and frequent instrument calibration. An overview of the present instrument will be given together with a description of how the raw data from the individual sensor channels are combined to give a continuous neutron energy spectrum. Digital processing techniques are now being applied to the TNS to handle some of the more complex analogue functions, particularly neutron/gamma-ray pulse-shape discrimination for the photo-scintillator column. Potential advantages of this approach are on qualities such a...
Electron transport through nuclear pasta in magnetized neutron stars
Yakovlev, D G
2015-01-01
We present a simple model for electron transport in a possible layer of exotic nuclear clusters (in the so called nuclear pasta layer) between the crust and liquid core of a strongly magnetized neutron star. The electron transport there can be strongly anisotropic and gyrotropic. The anisotropy is produced by different electron effective collision frequencies along and across local symmetry axis in domains of exotic ordered nuclear clusters and by complicated effects of the magnetic field. We also calculate averaged kinetic coefficients in case local domains are freely oriented. Possible applications of the obtained results and open problems are outlined.
A transportable neutron radiography system based on a SbBe neutron source
Fantidis, J. G.; Nicolaou, G. E.; Tsagas, N. F.
2009-07-01
A transportable neutron radiography system, incorporating a SbBe neutron source, has been simulated using the MCNPX code. Design provisions have allowed two radiography systems to be utilised using the same SbBe neutron source. In this respect, neutron radiographies can be carried out using the photoneutrons produced when the 124Sb is surrounded by the Be target. Alternatively, γ-radiography can be utilised with the photons from the 124Sb with the target removed. Appropriate collimators were simulated for each of the radiography modes. Apart from Be, the materials considered were compatible with the European Union Directive on 'Restriction of Hazardous Substances' (RoHS) 2002/95/EC, hence excluding the use of cadmium and lead. Bismuth was chosen as the material for γ-radiation shielding and the proposed system allowed a maximum activity of the 124Sb up to 1.85×1013 Bq. The system simulated allows different object sizes to be studied with a wide range of radiography parameters.
Beam-transport optimization for cold-neutron spectrometer
Directory of Open Access Journals (Sweden)
Nakajima Kenji
2015-01-01
Full Text Available We report the design of the beam-transport system (especially the vertical geometry for a cold-neutron disk-chopper spectrometer AMATERAS at J-PARC. Based on the elliptical shape, which is one of the most effective geometries for a ballistic mirror, the design was optimized to obtain, at the sample position, a neutron beam with high flux without serious degrading in divergence and spacial homogeneity within the boundary conditions required from actual spectrometer construction. The optimum focal point was examined. An ideal elliptical shape was modified to reduce its height without serious loss of transmission. The final result was adapted to the construction requirements of AMATERAS. Although the ideas studied in this paper are considered for the AMATERAS case, they can be useful also to other spectrometers in similar situations.
Benchmarking of neutron production of heavy-ion transport codes
Energy Technology Data Exchange (ETDEWEB)
Remec, I. [Oak Ridge National Laboratory, Oak Ridge, TN 37831-6172 (United States); Ronningen, R. M. [Michigan State Univ., National Superconductiong Cyclotron Laboratory, East Lansing, MI 48824-1321 (United States); Heilbronn, L. [Univ. of Tennessee, 1004 Estabrook Rd., Knoxville, TN 37996-2300 (United States)
2011-07-01
Document available in abstract form only, full text of document follows: Accurate prediction of radiation fields generated by heavy ion interactions is important in medical applications, space missions, and in design and operation of rare isotope research facilities. In recent years, several well-established computer codes in widespread use for particle and radiation transport calculations have been equipped with the capability to simulate heavy ion transport and interactions. To assess and validate these capabilities, we performed simulations of a series of benchmark-quality heavy ion experiments with the computer codes FLUKA, MARS15, MCNPX, and PHITS. We focus on the comparisons of secondary neutron production. Results are encouraging; however, further improvements in models and codes and additional benchmarking are required. (authors)
Establishment and Verification of MCNP Neutron Transport Model About Tianwan Nuclear Power Plant
Institute of Scientific and Technical Information of China (English)
ZHOU; Qi
2012-01-01
<正>In order to calculating the neutron flux in the surveillance box and reactor pressure vessel of the Tianwan NPP, we need to build up the neutron transport model by using the Monte Carlo code MCNP. The core of the NPP is very complicated for modeling so we put forward some assumptions that can simplify the neutron transport model. A lot of calculation works have been done to prove that the assumptions are right and suitable.
Wang, Yahui; Yan, Liming; Ma, Yu
2017-06-01
Applications of the transient Boltzmann transport equation (BTE) have undergone much investigation, such as radiative heat transfer and neutron transport. This paper provides a lattice Boltzmann model to efficiently resolve the multidimensional transient BTE. For a higher angular resolution, enough transport directions are considered while the transient BTE in each direction is treated as a conservation law equation and solved independently. Both macroscopic equations recovered from a Chapman-Enskog expansion and simulated results of typical benchmark problems show not only the second-order accuracy but also the flexibility and applicability of the proposed lattice Boltzmann model. This approach may contribute a powerful technique for the parallel simulation of large-scale engineering and some alternative perspectives for solving the nonlinear transport problem further.
Cosmic-ray neutron transport at a forest field site
DEFF Research Database (Denmark)
Andreasen, Mie; Jensen, Karsten Høgh; Desilets, Darin
2017-01-01
parameters describing the subsurface to match measured height profiles and time series of thermal and epithermal neutron intensities at a field site in Denmark. Overall, modeled thermal and epithermal neutron intensities are in satisfactory agreement with measurements; however, the choice of forest canopy...... conceptualization is found to be significant. Modeling results show that the effect of canopy interception, soil chemistry and dry bulk density of litter and mineral soil on neutron intensity is small. On the other hand, the neutron intensity decreases significantly with added litter-layer thickness, especially...... for epithermal neutron energies. Forest biomass also has a significant influence on the neutron intensity height profiles at the examined field site, altering both the shape of the profiles and the ground-level thermal-to-epithermal neutron ratio. This ratio increases with increasing amounts of biomass...
Interfacing MCNPX and McStas for simulation of neutron transport
Klinkby, Esben Bryndt; Lauritzen, Bent; Nonbøl, Erik; Willendrup, Peter Kjær; Filges, Uwe; Wohlmuther, Michael; Gallmeier, Franz X.
2013-01-01
Simulations of target-moderator-reflector system at spallation sources are conventionally carried out using Monte Carlo codes such as MCNPX[1] or FLUKA[2, 3] whereas simulations of neutron transport from the moderator and the instrument response are performed by neutron ray tracing codes such as McStas[4, 5, 6, 7]. The coupling between the two simulation suites typically consists of providing analytical fits of MCNPX neutron spectra to McStas. This method is generally successful but has limit...
Interfacing MCNPX and McStas for simulation of neutron transport
DEFF Research Database (Denmark)
Klinkby, Esben Bryndt; Lauritzen, Bent; Nonbøl, Erik
2013-01-01
Simulations of target-moderator-reflector system at spallation sources are conventionally carried out using Monte Carlo codes such as MCNPX[1] or FLUKA[2, 3] whereas simulations of neutron transport from the moderator and the instrument response are performed by neutron ray tracing codes such as Mc...
Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system
Energy Technology Data Exchange (ETDEWEB)
Iga, Kiminori [Kyushu Univ., Fukuoka (Japan); Takada, Hiroshi; Nagao, Tadashi
1998-01-01
In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B{sub 4}C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)
Neutron and photon transport calculations in fusion system. 2
Energy Technology Data Exchange (ETDEWEB)
Sato, Satoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment
1998-03-01
On the application of MCNP to the neutron and {gamma}-ray transport calculations for fusion reactor system, the wide range design calculation has been carried out in the engineering design activities for the international thermonuclear fusion experimental reactor (ITER) being developed jointly by Japan, USA, EU and Russia. As the objects of shielding calculation for fusion reactors, there are the assessment of dose equivalent rate for living body shielding and the assessment of the nuclear response for the soundness of in-core structures. In the case that the detailed analysis of complicated three-dimensional shapes is required, the assessment using MCNP has been carried out. Also when the nuclear response of peripheral equipment due to the gap streaming between blanket modules is evaluated with good accuracy, the calculation with MCNP has been carried out. The analyses of the shieldings for blanket modules and NBI port are explained, and the examples of the results of analyses are shown. In the blanket modules, there are penetrating holes and continuous gap. In the case of the NBI port, shielding plug cannot be installed. These facts necessitate the MCNP analysis with high accuracy. (K.I.)
PHISICS multi-group transport neutronic capabilities for RELAP5
Energy Technology Data Exchange (ETDEWEB)
Epiney, A.; Rabiti, C.; Alfonsi, A.; Wang, Y.; Cogliati, J.; Strydom, G. [Idaho National Laboratory (INL), 2525 N. Fremont Ave., Idaho Falls, ID 83402 (United States)
2012-07-01
PHISICS is a neutronic code system currently under development at INL. Its goal is to provide state of the art simulation capability to reactor designers. This paper reports on the effort of coupling this package to the thermal hydraulic system code RELAP5. This will enable full prismatic core and system modeling and the possibility to model coupled (thermal-hydraulics and neutronics) problems with more options for 3D neutron kinetics, compared to the existing diffusion theory neutron kinetics module in RELAP5 (NESTLE). The paper describes the capabilities of the coupling and illustrates them with a set of sample problems. (authors)
Transport of ultracold neutrons through a mirror system with surface roughness as a velocity filter
Chizhova, L A; Jenke, T; Cronenberg, G; Geltenbort, P; Abele, H; Burgdörfer, J
2012-01-01
We perform classical Monte Carlo simulations of ultracold neutron transport through an absorbing-reflecting mirror system in the Earth's gravitational field. We show that the underlying mixed phase space of regular skipping motion and random motion due to disorder scattering can be exploited to realize a velocity filter for ultracold neutrons. The range of velocities selected is controlled by geometric parameters of the wave guide. Possible applications include investigations of transport and scattering dynamics in confined systems.
VVER-440 Ex-Core Neutron Transport Calculations by MCNP-5 Code and Comparison with Experiment
Energy Technology Data Exchange (ETDEWEB)
Borodkin, Pavel; Khrennikov, Nikolay [Scientific and Engineering Centre for Nuclear and Radiation Safety (SEC NRS) Malaya Krasnoselskaya ul., 2/8, bld. 5, 107140 Moscow (Russian Federation)
2008-07-01
Ex-core neutron transport calculations are needed to evaluate radiation loading parameters (neutron fluence, fluence rate and spectra) on the in-vessel equipment, reactor pressure vessel (RPV) and support constructions of VVER type reactors. Due to these parameters are used for reactor equipment life-time assessment, neutron transport calculations should be carried out by precise and reliable calculation methods. In case of RPVs, especially, of first generation VVER-440s, the neutron fluence plays a key role in the prediction of RPV lifetime. Main part of VVER ex-core neutron transport calculations are performed by deterministic and Monte-Carlo methods. This paper deals with precise calculations of the Russian first generation VVER-440 by MCNP-5 code. The purpose of this work was an application of this code for expert calculations, verification of results by comparison with deterministic calculations and validation by neutron activation measured data. Deterministic discrete ordinates DORT code, widely used for RPV neutron dosimetry and many times tested by experiments, was used for comparison analyses. Ex-vessel neutron activation measurements at the VVER-440 NPP have provided space (in azimuth and height directions) and neutron energy (different activation reactions) distributions data for experimental (E) validation of calculated results. Calculational intercomparison (DORT vs. MCNP-5) and comparison with measured values (MCNP-5 and DORT vs. E) have shown agreement within 10-15% for different space points and reaction rates. The paper submits a discussion of results and makes conclusions about practice use of MCNP-5 code for ex-core neutron transport calculations in expert analysis. (authors)
Neutron interaction and their transport with bulk materials
Energy Technology Data Exchange (ETDEWEB)
Rani, Esther Kalpana, E-mail: esther.kalpanarani@gmail.com [Department of Physics JNT University, Nachupally, Karimnagar, Telangana, 500055 (India); Radhika, K., E-mail: radhikanit@gmail.com [Department of Humanities and Applied Sciences, Talla Padmavathi College of Engineering, Warangal, Telangana, 506004 (India)
2015-05-15
In the current paper an attempt was made to study and provide fundamental information about neutron interactions that are important to nuclear material measurements. The application of this study is explained about macroscopic interactions with bulk compound materials through a program in DEV C++ language which is done by enabling interaction of neutrons in nature. The output of the entire process depends upon the random number (i.e., incident neutron number), thickness of the material and mean free path as input parameters. Further the current study emphasizes on the usage of materials in shielding.
Comparison of 2D and 3D Neutron Transport Analyses on Yonggwang Unit 3 Reactor
Energy Technology Data Exchange (ETDEWEB)
Maeng, Aoung Jae; Kim, Byoung Chul; Lim, Mi Joung; Kim, Kyung Sik; Jeon, Young Kyou [Korea Reactor Integrity Surveillance Technology, Daejeon (Korea, Republic of); Yoo, Choon Sung [Korea Atomic Energy Research Institutes, Daejeon (Korea, Republic of)
2012-10-15
10 CFR Part 50 Appendix H requires periodical surveillance program in the reactor vessel (RV) belt line region of light water nuclear power plant to check vessel integrity resulting from the exposure to neutron irradiation and thermal environment. Exact exposure analysis of the neutron fluence based on right modeling and simulations is the most important in the evaluation. Traditional 2 dimensional (D) and 1D synthesis methodologies have been widely applied to evaluate the fast neutron (E > 1.0 MeV) fluence exposure to RV. However, 2D and 1D methodologies have not provided accurate fast neutron fluence evaluation at elevations far above or below the active core region. RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3D Geometries) program for 3D geometries calculation was therefore developed both by Westinghouse Electronic Company, USA and Korea Reactor Integrity Surveillance Technology (KRIST) for the analysis of In-Vessel Surveillance Test and Ex-Vessel Neutron Dosimetry (EVND). Especially EVND which is installed at active core height between biological shielding material and concrete also evaluates axial neutron fluence by placing three dosimetries each at Top, Middle and Bottom part of the angle representing maximum neutron fluence. The EVND programs have been applied to the Korea Nuclear Plants. The objective of this study is therefore to compare the 3D and the 2D Neutron Transport Calculations and Analyses on the Yonggwang unit 3 Reactor as an example.
MCNPX Monte Carlo simulations of particle transport in SiC semiconductor detectors of fast neutrons
Sedlačková, K.; Zat'ko, B.; Šagátová, A.; Pavlovič, M.; Nečas, V.; Stacho, M.
2014-05-01
The aim of this paper was to investigate particle transport properties of a fast neutron detector based on silicon carbide. MCNPX (Monte Carlo N-Particle eXtended) code was used in our study because it allows seamless particle transport, thus not only interacting neutrons can be inspected but also secondary particles can be banked for subsequent transport. Modelling of the fast-neutron response of a SiC detector was carried out for fast neutrons produced by 239Pu-Be source with the mean energy of about 4.3 MeV. Using the MCNPX code, the following quantities have been calculated: secondary particle flux densities, reaction rates of elastic/inelastic scattering and other nuclear reactions, distribution of residual ions, deposited energy and energy distribution of pulses. The values of reaction rates calculated for different types of reactions and resulting energy deposition values showed that the incident neutrons transfer part of the carried energy predominantly via elastic scattering on silicon and carbon atoms. Other fast-neutron induced reactions include inelastic scattering and nuclear reactions followed by production of α-particles and protons. Silicon and carbon recoil atoms, α-particles and protons are charged particles which contribute to the detector response. It was demonstrated that although the bare SiC material can register fast neutrons directly, its detection efficiency can be enlarged if it is covered by an appropriate conversion layer. Comparison of the simulation results with experimental data was successfully accomplished.
Spin diffusive modes and thermal transport in neutron star crusts
Sedrakian, Armen
2015-01-01
In this contribution we first review a method for obtaining the collective modes of pair-correlated neutron matter as found in a neutron star inner crust. We discuss two classes of modes corresponding to density and spin perturbations with energy spectra $\\omega = \\omega_0 + \\alpha q^2$, where $\\omega_0 = 2\\Delta$ is the threshold frequency and $\\Delta$ is the gap in the neutron fluid spectrum. For characteristic values of Landau parameters in neutron star crusts the exitonic density modes have $\\alpha 0$ and they exist above $\\omega_0$ which implies that these modes are damped. As an application of these findings we compute the thermal conductivity due to spin diffusive modes and show that it scales as $T^{1/2} \\exp(-2\\omega_0/T)$ in the case where their two-by-two scattering cross-section is weakly dependent on temperature.
Spherical harmonics method for neutron transport equation based on unstructured-meshes
Institute of Scientific and Technical Information of China (English)
CAO Liang-Zhi; WU Hong-Chun
2004-01-01
Based on a new second-order neutron transport equation, self-adjoint angular flux (SAAF) equation, the spherical harmonics (PN) method for neutron transport equation on unstructured-meshes is derived. The spherical harmonics function is used to expand the angular flux. A set of differential equations about the spatial variable, which are coupled with each other, can be obtained. They are solved iteratively by using the finite element method on unstructured-meshes. A two-dimension transport calculation program is coded according to the model. The numerical results of some benchmark problems demonstrate that this method can give high precision results and avoid the ray effect very well.
Institute of Scientific and Technical Information of China (English)
2008-01-01
A discrete ordinates method for a threedimensional first-order neutron transport equation based on unstructured-meshes that avoids the singularity of the second-order neutron transport equation in void regions was derived.The finite element variation equation was obtained using the least-squares method.A three-dimensional transport calculation code was developed.Both the triangular-z and the tetrahedron elements were included.The numerical results of some benchmark problems demonstrated that this method can solve neutron transport problems in unstructuredmeshes very well.For most problems,the error of the eigenvalue and the angular flux is less than 0.3% and 3.0% respectively.
How to polarise all neutrons in one beam: a high performance polariser and neutron transport system
Rodriguez, D. Martin; Bentley, P. M.; Pappas, C.
2016-09-01
Polarised neutron beams are used in disciplines as diverse as magnetism,soft matter or biology. However, most of these applications often suffer from low flux also because the existing neutron polarising methods imply the filtering of one of the spin states, with a transmission of 50% at maximum. With the purpose of using all neutrons that are usually discarded, we propose a system that splits them according to their polarisation, flips them to match the spin direction, and then focuses them at the sample. Monte Carlo (MC) simulations show that this is achievable over a wide wavelength range and with an outstanding performance at the price of a more divergent neutron beam at the sample position.
Neutron cross-section probability tables in TRIPOLI-3 Monte Carlo transport code
Energy Technology Data Exchange (ETDEWEB)
Zheng, S.H.; Vergnaud, T.; Nimal, J.C. [Commissariat a l`Energie Atomique, Gif-sur-Yvette (France). Lab. d`Etudes de Protection et de Probabilite
1998-03-01
Neutron transport calculations need an accurate treatment of cross sections. Two methods (multi-group and pointwise) are usually used. A third one, the probability table (PT) method, has been developed to produce a set of cross-section libraries, well adapted to describe the neutron interaction in the unresolved resonance energy range. Its advantage is to present properly the neutron cross-section fluctuation within a given energy group, allowing correct calculation of the self-shielding effect. Also, this PT cross-section representation is suitable for simulation of neutron propagation by the Monte Carlo method. The implementation of PTs in the TRIPOLI-3 three-dimensional general Monte Carlo transport code, developed at Commissariat a l`Energie Atomique, and several validation calculations are presented. The PT method is proved to be valid not only in the unresolved resonance range but also in all the other energy ranges.
Nobuhara, Fumiyoshi; Kuroyanagi, Makoto; Masumoto, Kazuyoshi; Nakamura, Hajime; Toyoda, Akihiro; Takahashi, Katsuhiko
2017-09-01
In order to evaluate the state of activation in a cyclotron facility used for the radioisotope production of PET diagnostics, we measured the neutron flux by using gold foils and TLDs. Then, the spatial distribution of neutrons and induced activity inside the cyclotron vault were simulated with the Monte Calro calculation code for neutron transport and DCHAIN-SP for activation calculation. The calculated results are in good agreement with measured values within factor 3. Therefore, the adaption of the advanced evaluation procedure for activation level is proved to be important for the planning of decommissioning of these facilities.
Neutron and Photon Transport in Sea-Going Cargo Containers
Energy Technology Data Exchange (ETDEWEB)
Pruet, J; Descalle, M; Hall, J; Pohl, B; Prussin, S G
2005-02-09
Factors affecting sensing of small quantities of fissionable material in large sea-going cargo containers by neutron interrogation and detection of {beta}-delayed photons are explored. The propagation of variable-energy neutrons in cargos, subsequent fission of hidden nuclear material and production of the {beta}-delayed photons, and the propagation of these photons to an external detector are considered explicitly. Detailed results of Monte Carlo simulations of these stages in representative cargos are presented. Analytical models are developed both as a basis for a quantitative understanding of the interrogation process and as a tool to allow ready extrapolation of the results to cases not specifically considered here.
Development of deterministic transport methods for low energy neutrons for shielding in space
Ganapol, Barry
1993-09-01
Transport of low energy neutrons associated with the galactic cosmic ray cascade is analyzed in this dissertation. A benchmark quality analytical algorithm is demonstrated for use with BRYNTRN, a computer program written by the High Energy Physics Division of NASA Langley Research Center, which is used to design and analyze shielding against the radiation created by the cascade. BRYNTRN uses numerical methods to solve the integral transport equations for baryons with the straight-ahead approximation, and numerical and empirical methods to generate the interaction probabilities. The straight-ahead approximation is adequate for charged particles, but not for neutrons. As NASA Langley improves BRYNTRN to include low energy neutrons, a benchmark quality solution is needed for comparison. The neutron transport algorithm demonstrated in this dissertation uses the closed-form Green's function solution to the galactic cosmic ray cascade transport equations to generate a source of neutrons. A basis function expansion for finite heterogeneous and semi-infinite homogeneous slabs with multiple energy groups and isotropic scattering is used to generate neutron fluxes resulting from the cascade. This method, called the FN method, is used to solve the neutral particle linear Boltzmann transport equation. As a demonstration of the algorithm coded in the programs MGSLAB and MGSEMI, neutron and ion fluxes are shown for a beam of fluorine ions at 1000 MeV per nucleon incident on semi-infinite and finite aluminum slabs. Also, to demonstrate that the shielding effectiveness against the radiation from the galactic cosmic ray cascade is not directly proportional to shield thickness, a graph of transmitted total neutron scalar flux versus slab thickness is shown. A simple model based on the nuclear liquid drop assumption is used to generate cross sections for the galactic cosmic ray cascade. The ENDF/B V database is used to generate the total and scattering cross sections for neutrons in
Improved Algorithms and Coupled Neutron-Photon Transport for Auto-Importance Sampling Method
Wang, Xin; Qiu, Rui; Li, Chun-Yan; Liang, Man-Chun; Zhang, Hui; Li, Jun-Li
2016-01-01
Auto-Importance Sampling (AIS) method is a Monte Carlo variance reduction technique proposed by Tsinghua University for deep penetration problem, which can improve computational efficiency significantly without pre-calculations for importance distribution. However AIS method is only validated with several basic deep penetration problems of simple geometries and cannot be used for coupled neutron-photon transport. This paper firstly presented the latest algorithm improvements for AIS method including particle transport, fictitious particles creation and adjustment, fictitious surface geometry, random number allocation and calculation of estimated relative error, which made AIS method applicable to complicated deep penetration problem. Then, a coupled Neutron-Photon Auto-Importance Sampling (NP-AIS) method was proposed to apply AIS method with the improved algorithms in coupled neutron-photon Monte Carlo transport. Finally, the NUREG/CR-6115 PWR benchmark model was calculated with the method of geometry splitti...
Least-squares finite element discretizations of neutron transport equations in 3 dimensions
Energy Technology Data Exchange (ETDEWEB)
Manteuffel, T.A [Univ. of Colorado, Boulder, CO (United States); Ressel, K.J. [Interdisciplinary Project Center for Supercomputing, Zurich (Switzerland); Starkes, G. [Universtaet Karlsruhe (Germany)
1996-12-31
The least-squares finite element framework to the neutron transport equation introduced in is based on the minimization of a least-squares functional applied to the properly scaled neutron transport equation. Here we report on some practical aspects of this approach for neutron transport calculations in three space dimensions. The systems of partial differential equations resulting from a P{sub 1} and P{sub 2} approximation of the angular dependence are derived. In the diffusive limit, the system is essentially a Poisson equation for zeroth moment and has a divergence structure for the set of moments of order 1. One of the key features of the least-squares approach is that it produces a posteriori error bounds. We report on the numerical results obtained for the minimum of the least-squares functional augmented by an additional boundary term using trilinear finite elements on a uniform tesselation into cubes.
Neutron and photon transport uncertainties of deep penetration in graphite
Energy Technology Data Exchange (ETDEWEB)
Jatuff, Fabian E. [INVAP S.E., San Carlos de Bariloche (Argentina)
1996-07-01
We describe the results obtained for the characterization of neutron and photon fields of the graphite thermal column of a 10-MW MTR-type research reactor, together with the mandatory Verification and Validation (V and V) procedure. The graphite thermal column exhibits a relatively small cross sectional are (50 cm X 50 cm) and a large depth ({approx}140cm), representing a difficult deep-penetration problem. The calculation line relied on estimations made the Monte Carlo MCNP-4.2 code. The Validation and Verification (V and V) procedure required: mandatory norms and an auditable path; ANISN/VITAMIN-C (deterministic) calculations for comparison with MCNP; identification of all approximations used, together with the method of justification; explicit statement of parameters for comparison, and statement of the areas of applicability; parametric studies concerning impurities and transverse leakage effects; statement of biases and uncertainties. The results showed a restricted range of applicability ({approx}60 cm) for fast and epithermal neutron fluxes, therefore needing a careful extrapolation to deeper locations. The transverse leakage of neutrons has a greater effect on the diffusion of thermal neutrons when compared to impurities of up to 5 ppm of natural boron. In addition, it is discussed the nature of the biases and uncertainty bands calculated. (author)
The Effect of Anisotropic Scatter on Atmospheric Neutron Transport
2015-03-26
slab geometry, two studies were conducted exploring the relative effect of anisotropic scatter as compared to isotropic scatter in the center of mass... anisotropic scatter. In order to address this question, first anisotropic scatter was implemented, then verified, and finally, the measurement of the... measured value. The relative error between neutron counts in isotropic and anisotropic time- integrated energy bins, isotropic anisotropicrel
Quantifying moisture transport in cementitious materials using neutron radiography
Lucero, Catherine L.
A portion of the concrete pavements in the US have recently been observed to have premature joint deterioration. This damage is caused in part by the ingress of fluids, like water, salt water, or deicing salts. The ingress of these fluids can damage concrete when they freeze and expand or can react with the cementitious matrix causing damage. To determine the quality of concrete for assessing potential service life it is often necessary to measure the rate of fluid ingress, or sorptivity. Neutron imaging is a powerful method for quantifying fluid penetration since it can describe where water has penetrated, how quickly it has penetrated and the volume of water in the concrete or mortar. Neutrons are sensitive to light atoms such as hydrogen and thus clearly detect water at high spatial and temporal resolution. It can be used to detect small changes in moisture content and is ideal for monitoring wetting and drying in mortar exposed to various fluids. This study aimed at developing a method to accurately estimate moisture content in mortar. The common practice is to image the material dry as a reference before exposing to fluid and normalizing subsequent images to the reference. The volume of water can then be computed using the Beer-Lambert law. This method can be limiting because it requires exact image alignment between the reference image and all subsequent images. A model of neutron attenuation in a multi-phase cementitious composite was developed to be used in cases where a reference image is not available. The attenuation coefficients for water, un-hydrated cement, and sand were directly calculated from the neutron images. The attenuation coefficient for the hydration products was then back-calculated. The model can estimate the degree of saturation in a mortar with known mixture proportions without using a reference image for calculation. Absorption in mortars exposed to various fluids (i.e., deionized water and calcium chloride solutions) were investigated
Guideline of Monte Carlo calculation. Neutron/gamma ray transport simulation by Monte Carlo method
2002-01-01
This report condenses basic theories and advanced applications of neutron/gamma ray transport calculations in many fields of nuclear energy research. Chapters 1 through 5 treat historical progress of Monte Carlo methods, general issues of variance reduction technique, cross section libraries used in continuous energy Monte Carlo codes. In chapter 6, the following issues are discussed: fusion benchmark experiments, design of ITER, experiment analyses of fast critical assembly, core analyses of JMTR, simulation of pulsed neutron experiment, core analyses of HTTR, duct streaming calculations, bulk shielding calculations, neutron/gamma ray transport calculations of the Hiroshima atomic bomb. Chapters 8 and 9 treat function enhancements of MCNP and MVP codes, and a parallel processing of Monte Carlo calculation, respectively. An important references are attached at the end of this report.
Energy Technology Data Exchange (ETDEWEB)
Brantley, P S
2006-09-27
We describe an asymptotic analysis of the coupled nonlinear system of equations describing time-dependent three-dimensional monoenergetic neutron transport and isotopic depletion and radioactive decay. The classic asymptotic diffusion scaling of Larsen and Keller [1], along with a consistent small scaling of the terms describing the radioactive decay of isotopes, is applied to this coupled nonlinear system of equations in a medium of specified initial isotopic composition. The analysis demonstrates that to leading order the neutron transport equation limits to the standard time-dependent neutron diffusion equation with macroscopic cross sections whose number densities are determined by the standard system of ordinary differential equations, the so-called Bateman equations, describing the temporal evolution of the nuclide number densities.
Talamo, Alberto
2013-05-01
This study presents three numerical algorithms to solve the time dependent neutron transport equation by the method of the characteristics. The algorithms have been developed taking into account delayed neutrons and they have been implemented into the novel MCART code, which solves the neutron transport equation for two-dimensional geometry and an arbitrary number of energy groups. The MCART code uses regular mesh for the representation of the spatial domain, it models up-scattering, and takes advantage of OPENMP and OPENGL algorithms for parallel computing and plotting, respectively. The code has been benchmarked with the multiplication factor results of a Boiling Water Reactor, with the analytical results for a prompt jump transient in an infinite medium, and with PARTISN and TDTORT results for cross section and source transients. The numerical simulations have shown that only two numerical algorithms are stable for small time steps.
Energy Technology Data Exchange (ETDEWEB)
Dumazert, Jonathan; Coulon, Romain; Carrel, Frédérick; Corre, Gwenolé; Normand, Stéphane [CEA, LIST, Laboratoire Capteurs Architectures Electroniques, 91191 Gif-sur-Yvette (France); Méchin, Laurence [CNRS, UCBN, Groupe de Recherche en Informatique, Image, Automatique et Instrumentation de Caen, 14050 Caen (France); Hamel, Matthieu [CEA, LIST, Laboratoire Capteurs Architectures Electroniques, 91191 Gif-sur-Yvette (France)
2016-08-21
Neutron detection forms a critical branch of nuclear-related issues, currently driven by the search for competitive alternative technologies to neutron counters based on the helium-3 isotope. The deployment of plastic scintillators shows a high potential for efficient detectors, safer and more reliable than liquids, more easily scalable and cost-effective than inorganic. In the meantime, natural gadolinium, through its 155 and mostly 157 isotopes, presents an exceptionally high interaction probability with thermal neutrons. This paper introduces a dual system including a metal gadolinium core inserted at the center of a high-scale plastic scintillator sphere. Incident fast neutrons are thermalized by the scintillator shell and then may be captured with a significant probability by gadolinium 155 and 157 nuclei in the core. The deposition of a sufficient fraction of the capture high-energy prompt gamma signature inside the scintillator shell will then allow discrimination from background radiations by energy threshold, and therefore neutron detection. The scaling of the system with the Monte Carlo MCNPX2.7 code was carried out according to a tradeoff between the moderation of incident fast neutrons and the probability of slow neutron capture by a moderate-cost metal gadolinium core. Based on the parameters extracted from simulation, a first laboratory prototype for the assessment of the detection method principle has been synthetized. The robustness and sensitivity of the neutron detection principle are then assessed by counting measurement experiments. Experimental results confirm the potential for a stable, highly sensitive, transportable and cost-efficient neutron detector and orientate future investigation toward promising axes.
Dumazert, Jonathan; Coulon, Romain; Carrel, Frédérick; Corre, Gwenolé; Normand, Stéphane; Méchin, Laurence; Hamel, Matthieu
2016-08-01
Neutron detection forms a critical branch of nuclear-related issues, currently driven by the search for competitive alternative technologies to neutron counters based on the helium-3 isotope. The deployment of plastic scintillators shows a high potential for efficient detectors, safer and more reliable than liquids, more easily scalable and cost-effective than inorganic. In the meantime, natural gadolinium, through its 155 and mostly 157 isotopes, presents an exceptionally high interaction probability with thermal neutrons. This paper introduces a dual system including a metal gadolinium core inserted at the center of a high-scale plastic scintillator sphere. Incident fast neutrons are thermalized by the scintillator shell and then may be captured with a significant probability by gadolinium 155 and 157 nuclei in the core. The deposition of a sufficient fraction of the capture high-energy prompt gamma signature inside the scintillator shell will then allow discrimination from background radiations by energy threshold, and therefore neutron detection. The scaling of the system with the Monte Carlo MCNPX2.7 code was carried out according to a tradeoff between the moderation of incident fast neutrons and the probability of slow neutron capture by a moderate-cost metal gadolinium core. Based on the parameters extracted from simulation, a first laboratory prototype for the assessment of the detection method principle has been synthetized. The robustness and sensitivity of the neutron detection principle are then assessed by counting measurement experiments. Experimental results confirm the potential for a stable, highly sensitive, transportable and cost-efficient neutron detector and orientate future investigation toward promising axes.
In situ neutron depth profiling: A powerful method to probe lithium transport in micro-batteries
Oudenhoven, J.F.M.; Labohm, F.; Mulder, M.; Niessen, R.A.H.; Mulder, F.M.; Notten, P.H.L.
2011-01-01
In situ neutron depth profiling (NDP) offers the possibility to observe lithium transport inside micro-batteries during battery operation. It is demonstrated that NDP results are consistent with the results of electrochemical measurements, and that the use of an enriched6LiCoO2 cathode offers more i
The neutron transport code DTF-Traca users manual and input data
Energy Technology Data Exchange (ETDEWEB)
Ahnert, C.
1979-07-01
This is a users manual of the neutron transport code DTF-TRACA, which is a version of the original DTF-IV with some modifications made at JEN. A detailed input data descriptions is given. The new options developed at JEN are included too. (Author) 18 refs.
The adjoint neutron transport equation and the statistical approach for its solution
Saracco, Paolo; Ravetto, Piero
2016-01-01
The adjoint equation was introduced in the early days of neutron transport and its solution, the neutron importance, has ben used for several applications in neutronics. The work presents at first a critical review of the adjoint neutron transport equation. Afterwards, the adjont model is constructed for a reference physical situation, for which an analytical approach is viable, i.e. an infinite homogeneous scattering medium. This problem leads to an equation that is the adjoint of the slowing-down equation that is well-known in nuclear reactor physics. A general closed-form analytical solution to such adjoint equation is obtained by a procedure that can be used also to derive the classical Placzek functions. This solution constitutes a benchmark for any statistical or numerical approach to the adjoint equation. A sampling technique to evaluate the adjoint flux for the transport equation is then proposed and physically interpreted as a transport model for pseudo-particles. This can be done by introducing appr...
The Fourier transform solution for the Green's function of monoenergetic neutron transport theory
Ganapol, Barry D.
2014-01-01
Nearly 45 years ago, Ken Case published his seminal paper on the singular eigenfunction solution for the Green's function of the monoenergetic neutron transport equation with isotropic scattering. Previously, the solution had been obtained by Fourier transform. While it is apparent the two had to be equivalent, a convincing equivalence proof for general anisotropic scattering remained a challenge until now.
Post-merger evolution of a neutron star-black hole binary with neutrino transport
Foucart, Francois; Roberts, Luke; Duez, Matthew D; Haas, Roland; Kidder, Lawrence E; Ott, Christian D; Pfeiffer, Harald P; Scheel, Mark A; Szilagyi, Bela
2015-01-01
We present a first simulation of the post-merger evolution of a black hole-neutron star binary in full general relativity using an energy-integrated general relativistic truncated moment formalism for neutrino transport. We describe our implementation of the moment formalism and important tests of our code, before studying the formation phase of a disk after a black hole-neutron star merger. We use as initial data an existing general relativistic simulation of the merger of a neutron star of 1.4 solar mass with a black hole of 7 solar mass and dimensionless spin a/M=0.8. Comparing with a simpler leakage scheme for the treatment of the neutrinos, we find noticeable differences in the neutron to proton ratio in and around the disk, and in the neutrino luminosity. We find that the electron neutrino luminosity is much lower in the transport simulations, and that the remnant is less neutron-rich. The spatial distribution of the neutrinos is significantly affected by relativistic effects. Over the short timescale e...
Transport simulation and image reconstruction for fast-neutron detection of explosives and narcotics
Energy Technology Data Exchange (ETDEWEB)
Micklich, B.J.; Fink, C.L.; Sagalovsky, L.
1995-07-01
Fast-neutron inspection techniques show considerable promise for explosive and narcotics detection. A key advantage of using fast neutrons is their sensitivity to low-Z elements (carbon, nitrogen, and oxygen), which are the primary constituents of these materials. We are currently investigating two interrogation methods in detail: Fast-Neutron Transmission Spectroscopy (FNTS) and Pulsed Fast-Neutron Analysis (PFNA). FNTS is being studied for explosives and narcotics detection in luggage and small containers for which the transmission ratio is greater than about 0.01. The Monte-Carlo radiation transport code MCNP is being used to simulate neutron transmission through a series of phantoms for a few (3-5) projection angles and modest (2 cm) resolution. Areal densities along projection rays are unfolded from the transmission data. Elemental abundances are obtained for individual voxels by tomographic reconstruction, and these reconstructed elemental images are combined to provide indications of the presence or absence of explosives or narcotics. PFNA techniques are being investigated for detection of narcotics in cargo containers because of the good penetration of the fast neutrons and the low attenuation of the resulting high-energy gamma-ray signatures. Analytic models and Monte-Carlo simulations are being used to explore the range of capabilities of PFNA techniques and to provide insight into systems engineering issues. Results of studies from both FNTS and PFNA techniques are presented.
OECD/NEA benchmark for time-dependent neutron transport calculations without spatial homogenization
Energy Technology Data Exchange (ETDEWEB)
Hou, Jason, E-mail: jason.hou@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Ivanov, Kostadin N. [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Boyarinov, Victor F.; Fomichenko, Peter A. [National Research Centre “Kurchatov Institute”, Kurchatov Sq. 1, Moscow (Russian Federation)
2017-06-15
Highlights: • A time-dependent homogenization-free neutron transport benchmark was created. • The first phase, known as the kinetics phase, was described in this work. • Preliminary results for selected 2-D transient exercises were presented. - Abstract: A Nuclear Energy Agency (NEA), Organization for Economic Co-operation and Development (OECD) benchmark for the time-dependent neutron transport calculations without spatial homogenization has been established in order to facilitate the development and assessment of numerical methods for solving the space-time neutron kinetics equations. The benchmark has been named the OECD/NEA C5G7-TD benchmark, and later extended with three consecutive phases each corresponding to one modelling stage of the multi-physics transient analysis of the nuclear reactor core. This paper provides a detailed introduction of the benchmark specification of Phase I, known as the “kinetics phase”, including the geometry description, supporting neutron transport data, transient scenarios in both two-dimensional (2-D) and three-dimensional (3-D) configurations, as well as the expected output parameters from the participants. Also presented are the preliminary results for the initial state 2-D core and selected transient exercises that have been obtained using the Monte Carlo method and the Surface Harmonic Method (SHM), respectively.
Ageing of a neutron shielding used in transport/storage casks
Energy Technology Data Exchange (ETDEWEB)
Nizeyiman, Fidele; Alami, Aatif; Issard, Herve; Bellenger, Veronique [TN International, 1 rue des herons, Montigny le Bretonneux, 78054 Saint Quentin en Yvelines (France); Laboratoire PIMM, Arts and Metiers ParisTech, 151 Bd de l' Hopital, 75013 Paris (France)
2012-07-11
In radioactive materials transport/storage casks, a mineral-filled vinylester composite is used for neutron shielding which relies on its hydrogen and boron atoms content. During cask service life, this composite is mainly subjected to three types of ageing: hydrothermal ageing, thermal oxidation and neutron irradiation. The aim of this study is to investigate the effect of hydrothermal ageing on the properties and chemical composition of this polymer composite. At high temperature (120 Degree-Sign C and 140 Degree-Sign C), the main consequence is the strong decrease of mechanical properties induced by the filler/matrix debonding.
A 2-D/3-D cartesian geometry non-conforming spherical harmonic neutron transport solver
Energy Technology Data Exchange (ETDEWEB)
Van Criekingen, S. [Laboratoire J.-L. Lions, Universite Pierre et Marie Curie, 175 rue du Chevaleret, 75013 Paris (France)]. E-mail: vancriekingen@ann.jussieu.fr
2007-03-15
A new 2-D/3-D transport core solver for the time-independent Boltzmann transport equation is presented. This solver, named FIESTA, is based on the second-order even-parity form of the transport equation. The angular discretization is performed through the expansion of the angular neutron flux into spherical harmonics (P {sub N} method). The novelty of this solver is the use of non-conforming finite elements for the spatial discretization. Such elements lead to a discontinuous scalar flux approximation. This interface continuity requirement relaxation property is shared with mixed-dual formulations discretized using Raviart-Thomas finite elements. Encouraging numerical results are presented.
Garg, S; Porcar, L; Woodka, A C; Butler, P D; Perez-Salas, U
2011-07-20
Proper cholesterol transport is essential to healthy cellular activity and any abnormality can lead to several fatal diseases. However, complete understandings of cholesterol homeostasis in the cell remains elusive, partly due to the wide variability in reported values for intra- and intermembrane cholesterol transport rates. Here, we used time-resolved small-angle neutron scattering to measure cholesterol intermembrane exchange and intramembrane flipping rates, in situ, without recourse to any external fields or compounds. We found significantly slower transport kinetics than reported by previous studies, particularly for intramembrane flipping where our measured rates are several orders of magnitude slower. We unambiguously demonstrate that the presence of chemical tags and extraneous compounds employed in traditional kinetic measurements dramatically affect the system thermodynamics, accelerating cholesterol transport rates by an order of magnitude. To our knowledge, this work provides new insights into cholesterol transport process disorders, and challenges many of the underlying assumptions used in most cholesterol transport studies to date.
Reference neutron transport calculation note for Korea nuclear power plants with 3-loop PWR reactors
Energy Technology Data Exchange (ETDEWEB)
Kim, Byung Cheol; Chang, Ki Oak
1997-05-01
Reactor pressure vessel (RPV) steels are subjected to neutron irradiation at a temperature of about 290 deg C. This radiation exposure alters the mechanical properties, leading to a shift of the brittle-to-ductile transition temperature toward higher temperatures and to a diminution of the rupture energy as determined by Charpy V-notch tests. This radiation embrittlement is one of the important aging factors of nuclear power plants. U.S. NRC recommended the basic requirements for the determination of the pressure vessel fluence by regulatory guide DG-1025 in order to reduce the uncertainty in the determination of neutron fluence calculation and measurements. The determination of the pressure vessel fluence is based on both calculations and measurements. The fluence prediction is made with a calculation and the measurements are used to qualify the calculational methodology. Because of the importance and the difficulty of these calculations, the method`s qualification by comparison to measurement must be made to ensure a reliable and accurate vessel fluence determination. This reference calculation note is to provide a series of forward and adjoint neutron transport calculations for use in the evaluation of neutron dosimetry from surveillance capsule irradiations at 3-loop PWR reactor as well as for use in the determination of the neutron exposure of the reactor vessel wall in accordance with U.S Regulatory Guide DG-1025 requirements. The calculations of the pressure vessel fluence consist of the following steps; (1) Determination of the geometrical and material input data, (2) Determination of the core neutron source, and (3) Propagation of the neutron fluence from the core to the vessel and into the cavity. (author). 12 tabs., 3 figs., 7 refs.
Interfacing MCNPX and McStas for simulation of neutron transport
Energy Technology Data Exchange (ETDEWEB)
Klinkby, Esben, E-mail: esbe@dtu.dk [DTU Nutech, Technical University of Denmark, DTU Risø Campus, Frederiksborgvej 399, DK-4000 Roskilde (Denmark); ESS Design Update Programme (Denmark); Lauritzen, Bent; Nonbøl, Erik [DTU Nutech, Technical University of Denmark, DTU Risø Campus, Frederiksborgvej 399, DK-4000 Roskilde (Denmark); ESS Design Update Programme (Denmark); Kjær Willendrup, Peter [DTU Physics, Technical University of Denmark, DTU Lyngby Campus, Anker Engelunds Vej 1, DK-2800 Kgs. Lyngby (Denmark); ESS Design Update Programme (Denmark); Filges, Uwe; Wohlmuther, Michael [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); ESS Design Update Programme (Switzerland); Gallmeier, Franz X. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)
2013-02-01
Simulations of target-moderator-reflector system at spallation sources are conventionally carried out using Monte Carlo codes such as MCNPX (Waters et al., 2007 [1]) or FLUKA (Battistoni et al., 2007; Ferrari et al., 2005 [2,3]) whereas simulations of neutron transport from the moderator and the instrument response are performed by neutron ray tracing codes such as McStas (Lefmann and Nielsen, 1999; Willendrup et al., 2004, 2011a,b [4–7]). The coupling between the two simulation suites typically consists of providing analytical fits of MCNPX neutron spectra to McStas. This method is generally successful but has limitations, as it e.g. does not allow for re-entry of neutrons into the MCNPX regime. Previous work to resolve such shortcomings includes the introduction of McStas inspired supermirrors in MCNPX. In the present paper different approaches to interface MCNPX and McStas are presented and applied to a simple test case. The direct coupling between MCNPX and McStas allows for more accurate simulations of e.g. complex moderator geometries, backgrounds, interference between beam-lines as well as shielding requirements along the neutron guides.
Interfacing MCNPX and McStas for simulation of neutron transport
Klinkby, Esben; Lauritzen, Bent; Nonbøl, Erik; Kjær Willendrup, Peter; Filges, Uwe; Wohlmuther, Michael; Gallmeier, Franz X.
2013-02-01
Simulations of target-moderator-reflector system at spallation sources are conventionally carried out using Monte Carlo codes such as MCNPX (Waters et al., 2007 [1]) or FLUKA (Battistoni et al., 2007; Ferrari et al., 2005 [2,3]) whereas simulations of neutron transport from the moderator and the instrument response are performed by neutron ray tracing codes such as McStas (Lefmann and Nielsen, 1999; Willendrup et al., 2004, 2011a,b [4-7]). The coupling between the two simulation suites typically consists of providing analytical fits of MCNPX neutron spectra to McStas. This method is generally successful but has limitations, as it e.g. does not allow for re-entry of neutrons into the MCNPX regime. Previous work to resolve such shortcomings includes the introduction of McStas inspired supermirrors in MCNPX. In the present paper different approaches to interface MCNPX and McStas are presented and applied to a simple test case. The direct coupling between MCNPX and McStas allows for more accurate simulations of e.g. complex moderator geometries, backgrounds, interference between beam-lines as well as shielding requirements along the neutron guides.
Holley, A. T.
2007-10-01
The goal of the UCNA experiment is to determine the angular correlation between the electron momentum and the neutron spin (the beta-asymmetry) in neutron decay using polarized ultracold neutrons (UCN). The experimental strategy is to transport UCN into a decay volume through a 7T static magnetic field using the magnetic potential to polarize the UCN. The initial UCN spin can then be reversed via an rf adiabatic spin-flipper in a 1T field region whose gradient is tailored to optimize the adiabatic spin-flipper's performance. The spin-flipper, which also allows in situ measurement of the UCN depolarization rate, is a resonant `bird-cage' cavity capable of producing rf fields in excess of 5G at 30Mhz. In order to minimize the UCN depolarization rate, UCN guides are constructed of diamond-like carbon films on quartz tubing, a technology which has been demonstrated to produce less than 3x10-3 depolarizations per bounce. The performance of this system will be described, and compared to expectations from detailed Monte Carlo transport models. The implications for high precision measurements of polarized ultracold neutrons will also be discussed.
Energy Technology Data Exchange (ETDEWEB)
Morel, J.E.
1981-01-01
A collocation method is developed for the solution of the one-dimensional neutron transport equation in slab geometry with both symmetric and polarly asymmetric scattering. For the symmetric scattering case, it is found that the collocation method offers a combination of some of the best characteristics of the finite-element and discrete-ordinates methods. For the asymmetric scattering case, it is found that the computational cost of cross-section data processing under the collocation approach can be significantly less than that associated with the discrete-ordinates approach. A general diffusion equation treating both symmetric and asymmetric scattering is developed and used in a synthetic acceleration algorithm to accelerate the iterative convergence of collocation solutions. It is shown that a certain type of asymmetric scattering can radically alter the asymptotic behavior of the transport solution and is mathematically equivalent within the diffusion approximation to particle transport under the influence of an electric field. The method is easily extended to other geometries and higher dimensions. Applications exist in the areas of neutron transport with highly anisotropic scattering (such as that associated with hydrogenous media), charged-particle transport, and particle transport in controlled-fusion plasmas. 23 figures, 6 tables.
Monte Carlo Neutrino Transport Through Remnant Disks from Neutron Star Mergers
Richers, S; O'Connor, Evan; Fernandez, Rodrigo; Ott, Christian
2015-01-01
We present Sedonu, a new open source, steady-state, special relativistic Monte Carlo (MC) neutrino transport code, available at bitbucket.org/srichers/sedonu. The code calculates the energy- and angle-dependent neutrino distribution function on fluid backgrounds of any number of spatial dimensions, calculates the rates of change of fluid internal energy and electron fraction, and solves for the equilibrium fluid temperature and electron fraction. We apply this method to snapshots from two dimensional simulations of accretion disks left behind by binary neutron star mergers, varying the input physics and comparing to the results obtained with a leakage scheme for the case of a central black hole and a central hypermassive neutron star. Neutrinos are guided away from the densest regions of the disk and escape preferentially around 45 degrees from the equatorial plane. Neutrino heating is strengthened by MC transport a few scale heights above the disk midplane near the innermost stable circular orbit, potentiall...
Applying Advanced Neutron Transport Calculations for Improving Fuel Performance Codes
Energy Technology Data Exchange (ETDEWEB)
Botazzoli, P.; Luzzi, L. [Politecnico di Milano, Department of Energy, Nuclear Engineering Division - CeSNEF, Milano (Italy); Schubert, A.; Van Uffelen, P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, Karlsruhe (Germany); Haeck, W. [Institute de Radioprotection et de Surete Nucleaire, Fontenay-aux-Roses (France)
2009-06-15
TRANSURANUS is a computer code for the thermal and mechanical analysis of fuel rods in nuclear reactors. As part of the code, the TUBRNP model calculates the local concentration of the actinides (U, Pu, Am, Cm), the main fission products (Xe, Kr, Cs and Nd) and {sup 4}He produced during the irradiation as a function of the radial position across a fuel pellet (radial profiles). These local quantities are required for the determination of the local power density, the local burn-up, and the source term of fission products and other inert gases. In previous works the neutronic code ALEPH has been used to validate the models for the actinides and fission products concentrations in UO{sub 2} fuels. A similar approach has been adopted in the present work for verifying the Helium production. The present paper focuses on the modelling of the Helium production in PWR oxide fuels (MOX and UO{sub 2}). A reliable prediction of the Helium production and release in LWR oxide fuels is of great interest in case of increasing burn-up, linear heat generation rates and Plutonium content. The contribution of the Helium released plays a fundamental role in the gap pressure and subsequently in the mechanical behaviour of the fuel rod, in particular during the storage of the high burn-up spent fuel. Helium is produced in oxide fuels by three main paths: (i) alpha decay of the actinides (the main contribution is due to {sup 242}Cm, {sup 238}Pu and {sup 244}Cm); (ii) (n,{alpha}) reactions; and (iii) ternary fission. In the present work, the contributions due to ternary fission and the (n,{alpha}) reaction on {sup 16}O as well as some refinements in the {sup 241}Am burn-up chain have been included in TUBRNP. The VESTA neutronic code has been used for the validation of the He production model. The generic VESTA Monte Carlo depletion interface developed at IRSN allows us to couple different Monte Carlo codes with a depletion module. It currently allows for combining the ORIGEN 2.2 isotope
Yong Wang; Wenzheng Yue; Mo Zhang
2016-01-01
The anisotropic transport of thermal neutron in heterogeneous porous media is of great research interests in many fields. In this paper, it is the first time that a new model based on micron X-ray computed tomography (CT) has been proposed to simultaneously consider both the separation of matrix and pore and the distribution of mineral components. We apply the Monte Carlo method to simulate thermal neutrons transporting through the model along different directions, and meanwhile detect those ...
The FN method for anisotropic scattering in neutron transport theory: the critical slab problem.
Gülecyüz, M. C.; Tezcan, C.
1996-08-01
The FN method which has been applied to many physical problems for isotropic and anisotropic scattering in neutron transport theory is extended for problems for extremely anisotropic scattering. This method depends on the Placzek lemma and the use of the infinite medium Green's function. Here the Green's function for extremely anisotropic scattering which was expressed as a combination of the Green's functions for isotropic scattering is used to solve the critical slab problem. It is shown that the criticality condition is in agreement with the one obtained previously by reducing the transport equation for anisotropic scattering to isotropic scattering and solving using the FN method.
A portable, parallel, object-oriented Monte Carlo neutron transport code in C++
Energy Technology Data Exchange (ETDEWEB)
Lee, S.R.; Cummings, J.C. [Los Alamos National Lab., NM (United States); Nolen, S.D. [Texas A and M Univ., College Station, TX (United States)]|[Los Alamos National Lab., NM (United States)
1997-05-01
We have developed a multi-group Monte Carlo neutron transport code using C++ and the Parallel Object-Oriented Methods and Applications (POOMA) class library. This transport code, called MC++, currently computes k and {alpha}-eigenvalues and is portable to and runs parallel on a wide variety of platforms, including MPPs, clustered SMPs, and individual workstations. It contains appropriate classes and abstractions for particle transport and, through the use of POOMA, for portable parallelism. Current capabilities of MC++ are discussed, along with physics and performance results on a variety of hardware, including all Accelerated Strategic Computing Initiative (ASCI) hardware. Current parallel performance indicates the ability to compute {alpha}-eigenvalues in seconds to minutes rather than hours to days. Future plans and the implementation of a general transport physics framework are also discussed.
Žukauskaite, A; Plukiene, R; Plukis, A
2007-01-01
Particle accelerators and other high energy facilities produce penetrating ionizing radiation (neutrons and γ-rays) that must be shielded. The objective of this work was to model photon and neutron transport in various materials, usually used as shielding, such as concrete, iron or graphite. Monte Carlo method allows obtaining answers by simulating individual particles and recording some aspects of their average behavior. In this work several nuclear experiments were modeled: AVF 65 – γ-ray beams (1-10 MeV), HIMAC and ISIS-800 – high energy neutrons (20-800 MeV) transport in iron and concrete. The results were then compared with experimental data.
Neutron and gamma ray transport calculations in shielding system
Energy Technology Data Exchange (ETDEWEB)
Masukawa, Fumihiro; Sakamoto, Hiroki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1998-03-01
In the shields for radiation in nuclear facilities, the penetrating holes of various kinds and irregular shapes are made for the reasons of operation, control and others. These penetrating holes and gaps are filled with air or the substances with relatively small shielding performance, and radiation flows out through them, which is called streaming. As the calculation techniques for the shielding design or analysis related to the streaming problem, there are the calculations by simplified evaluation, transport calculation and Monte Carlo method. In this report, the example of calculation by Monte Carlo method which is represented by MCNP code is discussed. A number of variance reduction techniques which seem effective for the analysis of streaming problem were tried. As to the investigation of the applicability of MCNP code to streaming analysis, the object of analysis which are the concrete walls without hole and with horizontal hole, oblique hole and bent oblique hole, the analysis procedure, the composition of concrete, and the conversion coefficient of dose equivalent, and the results of analysis are reported. As for variance reduction technique, cell importance was adopted. (K.I.)
Improved algorithms and coupled neutron-photon transport for auto-importance sampling method
Wang, Xin; Li, Jun-Li; Wu, Zhen; Qiu, Rui; Li, Chun-Yan; Liang, Man-Chun; Zhang, Hui; Gang, Zhi; Xu, Hong
2017-01-01
The Auto-Importance Sampling (AIS) method is a Monte Carlo variance reduction technique proposed for deep penetration problems, which can significantly improve computational efficiency without pre-calculations for importance distribution. However, the AIS method is only validated with several simple examples, and cannot be used for coupled neutron-photon transport. This paper presents improved algorithms for the AIS method, including particle transport, fictitious particle creation and adjustment, fictitious surface geometry, random number allocation and calculation of the estimated relative error. These improvements allow the AIS method to be applied to complicated deep penetration problems with complex geometry and multiple materials. A Completely coupled Neutron-Photon Auto-Importance Sampling (CNP-AIS) method is proposed to solve the deep penetration problems of coupled neutron-photon transport using the improved algorithms. The NUREG/CR-6115 PWR benchmark was calculated by using the methods of CNP-AIS, geometry splitting with Russian roulette and analog Monte Carlo, respectively. The calculation results of CNP-AIS are in good agreement with those of geometry splitting with Russian roulette and the benchmark solutions. The computational efficiency of CNP-AIS for both neutron and photon is much better than that of geometry splitting with Russian roulette in most cases, and increased by several orders of magnitude compared with that of the analog Monte Carlo. Supported by the subject of National Science and Technology Major Project of China (2013ZX06002001-007, 2011ZX06004-007) and National Natural Science Foundation of China (11275110, 11375103)
Neutron, electron and photon transport in ICF tragets in direct and fast ignition
Directory of Open Access Journals (Sweden)
A. Parvazian
2005-12-01
Full Text Available Fusion energy due to inertial confinement has progressed in the last few decades. In order to increase energy efficiency in this method various designs have been presented. The standard scheme for direct ignition and fast ignition fuel targets are considered. Neutrons, electrons and photons transport in targets containing different combinations of Li and Be are calculated in both direct and fast ignition schemes. To compress spherical multilayer targets having fuel in the central part, they are irradiated by laser or heavy ion beams. Neutrons energy deposition in the target is considered using Monte Carlo method code MCNP. A significant amount of neutrons energy is deposited in the target which resulted in growing fusion reactions rates. It is found that Beryllium compared to Lithium is more important. In an introductory consideration of relativistic electron beam transport into central part of a fast ignition target, we have calculated electron energy deposition in highly dense D-T fuel and Beryllium layer of the target. It has been concluded that a fast ignition scheme is preferred to direct ignition because of the absence of hydrodynamic instability.
Energy Technology Data Exchange (ETDEWEB)
Pazianotto, Mauricio Tizziani; Carlson, Brett Vern [Instituto Tecnologico de Aeronautica (ITA), Sao Jose dos Campos, SP (Brazil); Federico, Claudio Antonio; Goncalez, Odair Lelis [Centro Tecnico Aeroespacial (CTA), Sao Jose dos Campos, SP (Brazil). Instituto de Estudos Avancados
2011-07-01
Full text: Great effort is required to understand better the cosmic radiation (CR) dose received by sensitive equipment, on-board computers and aircraft crew members at Brazil airspace, because there is a large area of South America and Brazil subject to the South Atlantic Anomaly (SAA). High energy neutrons are produced by interactions between primary cosmic ray and atmospheric atoms, and also undergo moderation resulting in a wider spectrum of energy ranging from thermal energies (0:025eV ) to energies of several hundreds of MeV. Measurements of the cosmic radiation dose on-board aircrafts need to be followed with an integral flow monitor on the ground level in order to register CR intensity variations during the measurements. The Long Counter (LC) neutron detector was designed as a directional neutron flux meter standard because it presents fairly constant response for energy under 10MeV. However we would like to use it as a ground based neutron monitor for cosmic ray induced neutron spectrum (CRINS) that presents an isotropic fluency and a wider spectrum of energy. The LC was modeled and tested using a Monte Carlo transport simulation for irradiations with known neutron sources ({sup 241}Am-Be and {sup 251}Cf) as a benchmark. Using this geometric model its efficiency was calculated to CRINS isotropic flux, introducing high energy neutron interactions models. The objective of this work is to present the model for simulation of the isotropic neutron source employing the MCNPX code (Monte Carlo N-Particle eXtended) and then access the LC efficiency to compare it with experimental results for cosmic ray neutrons measures on ground level. (author)
Geant4 simulations of the neutron production and transport in the n_TOF spallation target
Lerendegui-Marco, J.; Cortés-Giraldo, M. A.; Guerrero, C.; Quesada, J. M.
2016-11-01
The neutron production and transport in the spallation target of the n_TOF facility at CERN has been simulated with Geant4. The results obtained with the different hadronic Physics Lists provided by Geant4 have been compared with the experimental neutron flux in n_TOF-EAR1. The best overall agreement in both the absolute value and the energy dependence of the flux from thermal to 1GeV, is obtained with the INCL++ model coupled with the Fritiof Model(FTFP). This Physics List has been thus used to simulate and study the main features of the new n_TOF-EAR2 beam line, currently in its commissioning phase.
CAD-Based Monte Carlo Neutron Transport KSTAR Analysis for KSTAR
Seo, Geon Ho; Choi, Sung Hoon; Shim, Hyung Jin
2017-09-01
The Monte Carlo (MC) neutron transport analysis for a complex nuclear system such as fusion facility may require accurate modeling of its complicated geometry. In order to take advantage of modeling capability of the computer aided design (CAD) system for the MC neutronics analysis, the Seoul National University MC code, McCARD, has been augmented with a CAD-based geometry processing module by imbedding the OpenCASCADE CAD kernel. In the developed module, the CAD geometry data are internally converted to the constructive solid geometry model with help of the CAD kernel. An efficient cell-searching algorithm is devised for the void space treatment. The performance of the CAD-based McCARD calculations are tested for the Korea Superconducting Tokamak Advanced Research device by comparing with results of the conventional MC calculations using a text-based geometry input.
THE COMMISSIONING PLAN FOR THE SPALLATION NEUTRON SOURCE RING AND TRANSPORT LINES.
Energy Technology Data Exchange (ETDEWEB)
RAPARIA,D.BLASKIEWICZ,M.LEE,Y.Y.WEI,J.ET AL.
2004-03-10
The Spallation Neutron Source (SNS) accelerator systems will provide a 1 GeV, 1.44 MW proton beam to a liquid mercury target for neutron production. In order to satisfy the accelerator systems' portion of the Critical Decision 4 (CD-4) commissioning goal (which marks the completion of the construction phase of the project), a beam pulse with intensity greater than 1 x 10{sup 13} protons must be accumulated in the ring, extracted in a single turn and delivered to the target. A commissioning plan has been formulated for bringing into operation and establishing nominal operating conditions for the various ring and transport line subsystems as well as for establishing beam conditions and parameters which meet the commissioning goal.
Energy Technology Data Exchange (ETDEWEB)
Moller, J.Y.
2012-01-10
To model the nuclear reactors, the stationary linear Boltzmann equation is solved. After discretizing the energy and the angular variables, the hyperbolic equation is numerically solved with the discontinuous finite element method. The MINARET code uses this method on a triangular unstructured mesh in order to deal with complex geometries (like containing arcs of circle). However, the meshes with straight edges only approximate such geometries. With curved edges, the mesh fits exactly to the geometry, and in some cases, the number of triangles decreases. The main task of this work is the study of finite elements on curved triangles with one or several curved edges. The choice of the basis functions is one of the main points for this kind of finite elements. We obtained a convergence result under the assumption that the curved triangles are not too deformed in comparison with the associated straight triangles. Furthermore, a code has been written to treat triangles with one, two or three curved edges. Another part of this work deals with the acceleration of transport calculations. Indeed, the problem is solved iteratively, and, in some cases, can converge really slowly. A DSA (Diffusion Synthetic Acceleration) method has been implemented using a technique from interior penalty methods. A Fourier analysis in 1D and 2D allows to estimate the acceleration for infinite periodical media, and to check the stability of the numerical scheme when strong heterogeneities exist. (author) [French] La modelisation des reacteurs nucleaires repose sur la resolution de l'equation de Boltzmann lineaire. Nous nous sommes interesses a la resolution spatiale de la forme stationnaire de cette equation. Apres discretisation en energie et en angle, l'equation hyperbolique est resolue numeriquement par la methode des elements finis discontinus. Le solveur MINARET utilise cette methode sur un maillage triangulaire non structure afin de pouvoir traiter des geometries complexes
Energy Technology Data Exchange (ETDEWEB)
Bal, G.
1995-07-01
To achieve whole core calculations of the neutron transport equation, we have to follow this 2 step method: space and energy homogenization of the assemblies; resolution of the homogenized equation on the whole core. However, this is no more valid when accidents occur (for instance depressurization causing locally strong heterogeneous media). One solution consists then in coupling two kinds of resolutions: a fine computation on the damaged cell (fine mesh, high number of energy groups) coupled with a coarse one everywhere else. We only deal here with steady state solutions (which already live in 6D spaces). We present here two such methods: The coupling by transmission of homogenized sections and the coupling by transmission of boundary conditions. To understand what this coupling is, we first restrict ourselves to 1D with respect to space in one energy group. The first two chapters deal with a recall of basic properties of the neutron transport equation. We give at chapter 3 some indications of the behaviour of the flux with respect to the cross sections. We present at chapter 4 some couplings and give some properties. Chapter 5 is devoted to a presentation of some numerical applications. (author). 9 refs., 7 figs.
Wang, Guan-bo; Liu, Han-gang; Wang, Kan; Yang, Xin; Feng, Qi-jie
2012-09-01
Thermal-to-fusion neutron convertor has being studied in China Academy of Engineering Physics (CAEP). Current Monte Carlo codes, such as MCNP and GEANT, are inadequate when applied in this multi-step reactions problems. A Monte Carlo tool RSMC (Reaction Sequence Monte Carlo) has been developed to simulate such coupled problem, from neutron absorption, to charged particle ionization and secondary neutron generation. "Forced particle production" variance reduction technique has been implemented to improve the calculation speed distinctly by making deuteron/triton induced secondary product plays a major role. Nuclear data is handled from ENDF or TENDL, and stopping power from SRIM, which described better for low energy deuteron/triton interactions. As a validation, accelerator driven mono-energy 14 MeV fusion neutron source is employed, which has been deeply studied and includes deuteron transport and secondary neutron generation. Various parameters, including fusion neutron angle distribution, average neutron energy at different emission directions, differential and integral energy distributions, are calculated with our tool and traditional deterministic method as references. As a result, we present the calculation results of convertor with RSMC, including conversion ratio of 1 mm 6LiD with a typical thermal neutron (Maxwell spectrum) incidence, and fusion neutron spectrum, which will be used for our experiment.
Stability and accuracy of 3D neutron transport simulations using the 2D/1D method in MPACT
Collins, Benjamin; Stimpson, Shane; Kelley, Blake W.; Young, Mitchell T. H.; Kochunas, Brendan; Graham, Aaron; Larsen, Edward W.; Downar, Thomas; Godfrey, Andrew
2016-12-01
A consistent "2D/1D" neutron transport method is derived from the 3D Boltzmann transport equation, to calculate fuel-pin-resolved neutron fluxes for realistic full-core Pressurized Water Reactor (PWR) problems. The 2D/1D method employs the Method of Characteristics to discretize the radial variables and a lower order transport solution to discretize the axial variable. This paper describes the theory of the 2D/1D method and its implementation in the MPACT code, which has become the whole-core deterministic neutron transport solver for the Consortium for Advanced Simulations of Light Water Reactors (CASL) core simulator VERA-CS. Several applications have been performed on both leadership-class and industry-class computing clusters. Results are presented for whole-core solutions of the Watts Bar Nuclear Power Station Unit 1 and compared to both continuous-energy Monte Carlo results and plant data.
Stability and accuracy of 3D neutron transport simulations using the 2D/1D method in MPACT
Energy Technology Data Exchange (ETDEWEB)
Collins, Benjamin, E-mail: collinsbs@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Stimpson, Shane, E-mail: stimpsonsg@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Kelley, Blake W., E-mail: kelleybl@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Young, Mitchell T.H., E-mail: youngmit@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Kochunas, Brendan, E-mail: bkochuna@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Graham, Aaron, E-mail: aarograh@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Larsen, Edward W., E-mail: edlarsen@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Downar, Thomas, E-mail: downar@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Godfrey, Andrew, E-mail: godfreyat@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Rd., Oak Ridge, TN 37831 (United States)
2016-12-01
A consistent “2D/1D” neutron transport method is derived from the 3D Boltzmann transport equation, to calculate fuel-pin-resolved neutron fluxes for realistic full-core Pressurized Water Reactor (PWR) problems. The 2D/1D method employs the Method of Characteristics to discretize the radial variables and a lower order transport solution to discretize the axial variable. This paper describes the theory of the 2D/1D method and its implementation in the MPACT code, which has become the whole-core deterministic neutron transport solver for the Consortium for Advanced Simulations of Light Water Reactors (CASL) core simulator VERA-CS. Several applications have been performed on both leadership-class and industry-class computing clusters. Results are presented for whole-core solutions of the Watts Bar Nuclear Power Station Unit 1 and compared to both continuous-energy Monte Carlo results and plant data.
Preliminary study on CAD-based method of characteristics for neutron transport calculation
Chen, Zhen-Ping; Sun, Guang-Yao; Song, Jing; Hao, Li-Juan; Hu, Li-Qin; Wu, Yi-Can
2013-01-01
The method of characteristics (MOC) is widely used for neutron transport calculation in recent decades. However, the key problem determining whether MOC can be applied in highly heterogeneous geometry is how to combine an effective geometry modeling method with it. Most of the existing MOC codes conventionally describe the geometry model just by lines and arcs with extensive input data. Thus they have difficulty in geometry modeling and ray tracing for complicated geometries. In this study, a new method making use of a CAD-based automatic modeling tool MCAM which is a CAD/Image-based Automatic Modeling Program for Neutronics and Radiation Transport developed by FDS Team in China was introduced for geometry modeling and ray tracing of particle transport to remove those limitations. The diamond -difference scheme was applied to MOC to reduce the spatial discretization errors of the flat flux approximation. Based on MCAM and MOC, a new MOC code was developed and integrated into SuperMC system, whic h is a Super ...
Hybrid Parallel Programming Models for AMR Neutron Monte-Carlo Transport
Dureau, David; Poëtte, Gaël
2014-06-01
This paper deals with High Performance Computing (HPC) applied to neutron transport theory on complex geometries, thanks to both an Adaptive Mesh Refinement (AMR) algorithm and a Monte-Carlo (MC) solver. Several Parallelism models are presented and analyzed in this context, among them shared memory and distributed memory ones such as Domain Replication and Domain Decomposition, together with Hybrid strategies. The study is illustrated by weak and strong scalability tests on complex benchmarks on several thousands of cores thanks to the petaflopic supercomputer Tera100.
Žukauskaitėa, A; Plukienė, R; Ridikas, D
2007-01-01
Particle accelerators and other high energy facilities produce penetrating ionizing radiation (neutrons and γ-rays) that must be shielded. The objective of this work was to model photon and neutron transport in various materials, usually used as shielding, such as concrete, iron or graphite. Monte Carlo method allows obtaining answers by simulating individual particles and recording some aspects of their average behavior. In this work several nuclear experiments were modeled: AVF 65 (AVF cyclotron of Research Center of Nuclear Physics, Osaka University, Japan) – γ-ray beams (1-10 MeV), HIMAC (heavy-ion synchrotron of the National Institute of Radiological Sciences in Chiba, Japan) and ISIS-800 (ISIS intensive spallation neutron source facility of the Rutherford Appleton laboratory, UK) – high energy neutron (20-800 MeV) transport in iron and concrete. The calculation results were then compared with experimental data.compared with experimental data.
Zimmer, Oliver
2016-01-01
A neutron optical transport system is proposed which comprises nested short elliptical mirrors located halfway between two common focal points M and M'. It images cold neutrons from a diverging beam or a source with finite size at M by single reflections onto a spot of similar size at M'. Direct view onto the neutron source is blocked by a central absorber with little impact on the transported solid angle. Geometric neutron losses due to source size can be kept small using modern supermirrors and distances M-M' of a few tens of metres. Very short flat mirrors can be used in practical implementations. Transport with a minimum of reflections remedies losses due to multiple reflections that are common in long elliptical neutron guides. Moreover, well-defined reflection angles lead to new possibilities for enhancing the spectral quality of primary beams, such as clear-cut discrimination of short neutron wavelengths or beam monochromation using bandpass supermirrors. Multi-mirror imaging systems may thus complemen...
A Deterministic-Monte Carlo Hybrid Method for Time-Dependent Neutron Transport Problems
Energy Technology Data Exchange (ETDEWEB)
Justin Pounders; Farzad Rahnema
2001-10-01
A new deterministic-Monte Carlo hybrid solution technique is derived for the time-dependent transport equation. This new approach is based on dividing the time domain into a number of coarse intervals and expanding the transport solution in a series of polynomials within each interval. The solutions within each interval can be represented in terms of arbitrary source terms by using precomputed response functions. In the current work, the time-dependent response function computations are performed using the Monte Carlo method, while the global time-step march is performed deterministically. This work extends previous work by coupling the time-dependent expansions to space- and angle-dependent expansions to fully characterize the 1D transport response/solution. More generally, this approach represents and incremental extension of the steady-state coarse-mesh transport method that is based on global-local decompositions of large neutron transport problems. An example of a homogeneous slab is discussed as an example of the new developments.
MONTE CARLO NEUTRINO TRANSPORT THROUGH REMNANT DISKS FROM NEUTRON STAR MERGERS
Energy Technology Data Exchange (ETDEWEB)
Richers, Sherwood; Ott, Christian D. [TAPIR, Mailcode 350-17, Walter Burke Institute for Theoretical Physics, California Institute of Technology, Pasadena, CA 91125 (United States); Kasen, Daniel; Fernández, Rodrigo [Department of Astronomy and Theoretical Astrophysics Center, University of California, Berkeley, CA 94720 (United States); O’Connor, Evan [Department of Physics, Campus Code 8202, North Carolina State University, Raleigh, NC 27695 (United States)
2015-11-01
We present Sedonu, a new open source, steady-state, special relativistic Monte Carlo (MC) neutrino transport code, available at bitbucket.org/srichers/sedonu. The code calculates the energy- and angle-dependent neutrino distribution function on fluid backgrounds of any number of spatial dimensions, calculates the rates of change of fluid internal energy and electron fraction, and solves for the equilibrium fluid temperature and electron fraction. We apply this method to snapshots from two-dimensional simulations of accretion disks left behind by binary neutron star mergers, varying the input physics and comparing to the results obtained with a leakage scheme for the cases of a central black hole and a central hypermassive neutron star. Neutrinos are guided away from the densest regions of the disk and escape preferentially around 45° from the equatorial plane. Neutrino heating is strengthened by MC transport a few scale heights above the disk midplane near the innermost stable circular orbit, potentially leading to a stronger neutrino-driven wind. Neutrino cooling in the dense midplane of the disk is stronger when using MC transport, leading to a globally higher cooling rate by a factor of a few and a larger leptonization rate by an order of magnitude. We calculate neutrino pair annihilation rates and estimate that an energy of 2.8 × 10{sup 46} erg is deposited within 45° of the symmetry axis over 300 ms when a central BH is present. Similarly, 1.9 × 10{sup 48} erg is deposited over 3 s when an HMNS sits at the center, but neither estimate is likely to be sufficient to drive a gamma-ray burst jet.
Energy Technology Data Exchange (ETDEWEB)
Moraes, Pedro Gabriel B.; Leite, Michel C.A.; Barros, Ricardo C., E-mail: pgbmoraes@gmail.com, E-mail: chell_leite@hotmail.com, E-mail: rcbarros@pq.cnpq.br [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Instituto Politecnico. Departamento de Modelagem Computacional
2013-07-01
In this work we developed a software to model and generate results in tables and graphs of one-dimensional neutron transport problems in multi-group formulation of energy. The numerical method we use to solve the problem of neutron diffusion is analytic, thus eliminating the truncation errors that appear in classical numerical methods, e.g., the method of finite differences. This numerical analytical method increases the computational efficiency, since they are not refined spatial discretization necessary because for any spatial discretization grids used, the numerical result generated for the same point of the domain remains unchanged unless the rounding errors of computational finite arithmetic. We chose to develop a computational application in MatLab platform for numerical computation and program interface is simple and easy with knobs. We consider important to model this neutron transport problem with a fixed source in the context of shielding calculations of radiation that protects the biosphere, and could be sensitive to ionizing radiation.
Thomas, Justin W.
2006-12-01
The Numerical Nuclear Reactor (NNR) is a code suite that is being developed to provide high-fidelity multi-physics capability for the analysis of light water nuclear reactors. The focus of the work here is to extend the capability of the NNR by incorporation of the neutronics module, DeCART, for Boiling Water Reactor (BWR) applications. The DeCART code has been coupled to the NNR fluid mechanics and heat transfer module STAR-CD for light water reactor applications. The coupling has been accomplished via an interface program, which is responsible for mapping the STAR-CD and DeCART meshes, managing communication, and monitoring convergence. DeCART obtains the solution of the 3-D Boltzmann transport equation by performing a series of 2-D modular ray tracing-based method of characteristics problems that are coupled within the framework of 3-D coarse-mesh finite difference. The relatively complex geometry and increased axial heterogeneity found in BWRs are beyond the modeling capability of the original version of DeCART. In this work, DeCART is extended in three primary areas. First, the geometric capability is generalized by extending the modular ray tracing scheme and permitting an unstructured mesh in the global finite difference kernel. Second, numerical instabilities, which arose as a result of the severe axial heterogeneity found in BWR cores, have been resolved. Third, an advanced nodal method has been implemented to improve the accuracy of the axial flux distribution. In this semi-analytic nodal method, the analytic solution to the transverse-integrated neutron diffusion equation is obtained, where the nonhomogeneous neutron source was first approximated by a quartic polynomial. The successful completion of these three tasks has allowed the application of the coupled DeCART/STAR-CD code to practical BWR problems.
Global Error Bounds for the Petrov-Galerkin Discretization of the Neutron Transport Equation
Energy Technology Data Exchange (ETDEWEB)
Chang, B; Brown, P; Greenbaum, A; Machorro, E
2005-01-21
In this paper, we prove that the numerical solution of the mono-directional neutron transport equation by the Petrov-Galerkin method converges to the true solution in the L{sup 2} norm at the rate of h{sup 2}. Since consistency has been shown elsewhere, the focus here is on stability. We prove that the system of Petrov-Galerkin equations is stable by showing that the 2-norm of the inverse of the matrix for the system of equations is bounded by a number that is independent of the order of the matrix. This bound is equal to the length of the longest path that it takes a neutron to cross the domain in a straight line. A consequence of this bound is that the global error of the Petrov-Galerkin approximation is of the same order of h as the local truncation error. We use this result to explain the widely held observation that the solution of the Petrov-Galerkin method is second accurate for one class of problems, but is only first order accurate for another class of problems.
Using the transportable, computer-operated, liquid-scintillator fast-neutron spectrometer system
Energy Technology Data Exchange (ETDEWEB)
Thorngate, J.H.
1988-11-01
When a detailed energy spectrum is needed for radiation-protection measurements from approximately 1 MeV up to several tens of MeV, organic-liquid scintillators make good neutron spectrometers. However, such a spectrometer requires a sophisticated electronics system and a computer to reduce the spectrum from the recorded data. Recently, we added a Nuclear Instrument Module (NIM) multichannel analyzer and a lap-top computer to the NIM electronics we have used for several years. The result is a transportable fast-neutron spectrometer system. The computer was programmed to guide the user through setting up the system, calibrating the spectrometer, measuring the spectrum, and reducing the data. Measurements can be made over three energy ranges, 0.6--2 MeV, 1.1--8 MeV, or 1.6--16 MeV, with the spectrum presented in 0.1-MeV increments. Results can be stored on a disk, presented in a table, and shown in graphical form. 5 refs., 51 figs.
Beam transient analyses of Accelerator Driven Subcritical Reactors based on neutron transport method
Energy Technology Data Exchange (ETDEWEB)
He, Mingtao; Wu, Hongchun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China); Zheng, Youqi, E-mail: yqzheng@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China); Wang, Kunpeng [Nuclear and Radiation Safety Center, PO Box 8088, Beijing 100082 (China); Li, Xunzhao; Zhou, Shengcheng [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China)
2015-12-15
Highlights: • A transport-based kinetics code for Accelerator Driven Subcritical Reactors is developed. • The performance of different kinetics methods adapted to the ADSR is investigated. • The impacts of neutronic parameters deteriorating with fuel depletion are investigated. - Abstract: The Accelerator Driven Subcritical Reactor (ADSR) is almost external source dominated since there is no additional reactivity control mechanism in most designs. This paper focuses on beam-induced transients with an in-house developed dynamic analysis code. The performance of different kinetics methods adapted to the ADSR is investigated, including the point kinetics approximation and space–time kinetics methods. Then, the transient responds of beam trip and beam overpower are calculated and analyzed for an ADSR design dedicated for minor actinides transmutation. The impacts of some safety-related neutronics parameters deteriorating with fuel depletion are also investigated. The results show that the power distribution varying with burnup leads to large differences in temperature responds during transients, while the impacts of kinetic parameters and feedback coefficients are not very obvious. Classification: Core physic.
Institute of Scientific and Technical Information of China (English)
无
2006-01-01
The single-group,steadystate,isotropic for mofthe neutron transport equationis given by[1]Ω·+σtI-σsPψ(x,Ω)=q(x,Ω)(x,Ω)∈D×Sψ(x,Ω)=g(x,Ω)x∈Din={x∈D,γ(x)·Ω<0(1)whereσtis the total cross section,σSis the scatteringcross section,andψ(x,Ω)is the angular flux to bedeter mined for all pointsx∈D,D Rn(n=2,3)and all possible travel directionsΩ,ΩS(Sis a u-nit disk or a unit sphere),γ(x)denotes the out wardunit nor mal atx∈D,Idenotes the identity opera-tor,the operatorPis defined by[Pψ](x)=∫Sψ(x,Ω)dΩ(2)Whenσt→∞,andσσ...
GPU-based high performance Monte Carlo simulation in neutron transport
Energy Technology Data Exchange (ETDEWEB)
Heimlich, Adino; Mol, Antonio C.A.; Pereira, Claudio M.N.A. [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. de Inteligencia Artificial Aplicada], e-mail: cmnap@ien.gov.br
2009-07-01
Graphics Processing Units (GPU) are high performance co-processors intended, originally, to improve the use and quality of computer graphics applications. Since researchers and practitioners realized the potential of using GPU for general purpose, their application has been extended to other fields out of computer graphics scope. The main objective of this work is to evaluate the impact of using GPU in neutron transport simulation by Monte Carlo method. To accomplish that, GPU- and CPU-based (single and multicore) approaches were developed and applied to a simple, but time-consuming problem. Comparisons demonstrated that the GPU-based approach is about 15 times faster than a parallel 8-core CPU-based approach also developed in this work. (author)
Design of the low energy beam transport line for the China spallation neutron source
Institute of Scientific and Technical Information of China (English)
LI Jin-Hai; OUYANG Hua-Fu; FU Shi-Nian; ZHANG Sua-Shun; HE Wei
2008-01-01
The design of the China Spallation Neutron Source (CSNS) low-energy beam transport (LEBT) line, which locates between the ion source and the radio-frequency quadrupole (RFQ), has been completed with the TRACE3D code. The design aims at perfect matching, primary chopping, a small emittance growth and sufficient space for beam diagnostics. The line consists of three solenoids, three vacuum chambers, two steering magnets and a pre-chopper. The total length of LEBT is about 1.74 m. This LEBT is designed to transfer 20 mA of H-pulsed beam from the ion source to the RFQ. An induction cavity is adopted as the pre-chopper.The electrostatic octupole steerer is discussed as a candidate. A four-quadrant aperture for beam scraping and beam position monitoring is designed.
Wang, Yi; He, Xi; Mukherjee, T.; Fitzsimmons, M. R.; Sahoo, S.; Binek, Ch.
2011-11-01
Exchange coupled magnetic hard layer/soft layer thin films show a variety of complex magnetization reversal mechanisms depending on the hierarchy of interaction strengths within and between the films. Magnetization reversal can include uniform rotation, soft layer biasing, as well as exchange spring behavior. We investigate the magnetization reversal of a CoPt/Permalloy/Ta/Permalloy heterostructure. Here, Stoner-Wohlfarth-type uniform magnetization rotation of the virtually free Permalloy layer and exchange spring behavior of the strongly pinned Permalloy layer are found in the same sample. We investigate the complex magnetization reversal by polarized neutron reflectometry, magnetometry, and magneto-transport. The synergy of combining these experimental methods together with theoretical modeling is key to obtain the complete quantitative depth resolved information of the magnetization reversal processes for a multilayer of mesoscopic thickness.
Energy Technology Data Exchange (ETDEWEB)
Wang Yi; He Xi; Mukherjee, T.; Binek, Ch. [Department of Physics and Astronomy and Nebraska Center for Materials and Nanoscience, Jorgenson Hall, University of Nebraska, Lincoln, Nebraska 68588-0111 (United States); Fitzsimmons, M. R. [Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States); Sahoo, S. [Seagate Technology, Minneapolis, Minnesota 55435 (United States)
2011-11-15
Exchange coupled magnetic hard layer/soft layer thin films show a variety of complex magnetization reversal mechanisms depending on the hierarchy of interaction strengths within and between the films. Magnetization reversal can include uniform rotation, soft layer biasing, as well as exchange spring behavior. We investigate the magnetization reversal of a CoPt/Permalloy/Ta/Permalloy heterostructure. Here, Stoner-Wohlfarth-type uniform magnetization rotation of the virtually free Permalloy layer and exchange spring behavior of the strongly pinned Permalloy layer are found in the same sample. We investigate the complex magnetization reversal by polarized neutron reflectometry, magnetometry, and magneto-transport. The synergy of combining these experimental methods together with theoretical modeling is key to obtain the complete quantitative depth resolved information of the magnetization reversal processes for a multilayer of mesoscopic thickness.
TART97 a coupled neutron-photon 3-D, combinatorial geometry Monte Carlo transport code
Energy Technology Data Exchange (ETDEWEB)
Cullen, D.E.
1997-11-22
TART97 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo transport code. This code can on any modern computer. It is a complete system to assist you with input preparation, running Monte Carlo calculations, and analysis of output results. TART97 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART97 is distributed on CD. This CD contains on- line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART97 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART97 and its data riles.
Anisotropic scattering treatment for the neutron transport equation with primal finite elements
Energy Technology Data Exchange (ETDEWEB)
Akherraz, B.; Fedon-Magnaud, C.; Lautard, J.J.; Sanchez, R. [Commissariat a l`Energie Atomique, Gif-sur-Yvette (France)
1995-07-01
Three approaches are presented to treat anisotropic scattering in neutron transport. The approaches are based on the even-odd-parity flux formalism and yield three different second-order equations for the even-parity flux. The first one is based on the total elimination of the odd-parity flux of the second-order equation. In the other two approaches, anisotropic scattering contributions are homogenized and incorporated into the collision term. The numerical solutions of these equations are implemented in the CRONOS code for pressurized water reactor core calculations and are done with a finite element spatial approximation and the discrete ordinates methods (S{sub N}) for the angular variable. Numerical results are presented for critical problems (k{sub eff}) in x-y geometry. Comparisons with the APOLLO2 assembly code show the accuracy and the efficiency of the proposed algorithms.
Wang, Yong; Yue, Wenzheng; Zhang, Mo
2016-06-01
The anisotropic transport of thermal neutron in heterogeneous porous media is of great research interests in many fields. In this paper, it is the first time that a new model based on micron X-ray computed tomography (CT) has been proposed to simultaneously consider both the separation of matrix and pore and the distribution of mineral components. We apply the Monte Carlo method to simulate thermal neutrons transporting through the model along different directions, and meanwhile detect those unreacted thermal neutrons by an array detector on the other side of the model. Therefore, the anisotropy of pore structure can be imaged by the amount of received thermal neutrons, due to the difference of rock matrix and pore-filling fluids in the macroscopic reaction cross section (MRCS). The new model has been verified by the consistent between the simulated data and the pore distribution from X-ray CT. The results show that the evaluation of porosity can be affected by the anisotropy of media. Based on the research, a new formula is developed to describe the correlation between the resolution of array detectors and the quality of imaging. The formula can be further used to analyze the critical resolution and the suitable number of thermal neutrons emitted in each simulation. Unconventionally, we find that a higher resolution cannot always lead to a better image.
Energy Technology Data Exchange (ETDEWEB)
Attaya, H.
1995-01-01
The primary goal of this task is to provide the capabilities in the activation code RACC, to treat pulsed operation modes. In addition, it is required that the code utilizes the same spatial mesh and geometrical models as employed in the one or multidimensional neutron transport codes used in ITER design. This would ensure the use of the same neutron flux generated by those codes to calculate the different activation parameters. It is also required to have the capabilities for generating graphical outputs for the calculated activation parameters.
On-the-fly Neutron Tomography of Water Transport into Lupine Roots
Zarebanadkouki, Mohsen; Carminati, Andrea; Kaestner, Anders; Mannes, David; Morgano, Manuel; Peetermans, Steven; Lehmann, Eberhard; Trtik, Pavel
Measurement and visualization of water flow in soil and roots is essential for understanding of how roots take up water from soils. Such information would allow for the optimization of irrigation practices and for the identification of the optimal traits for the capture of water, in particular when water is scarce. However, measuring water flow in roots growing in soil is challenging. The previous 2D experiments (Zarebanadkouki et al., 2012) have not been sufficient for understanding the water transport across the root and therefore we employed an on-the-fly tomography technique with temporal resolution of three minutes. In this paper, we show that the series of on-the-fly neutron tomographic experiments performed on the same sample allow for monitoring the three-dimensional spatial distribution of D2O across the root tissue. The obtained data will allow us to calculate the convective and diffusive transport properties across root tissue and to estimate the relative importance of different pathways of water across the root tissue.
The TORT three-dimensional discrete ordinates neutron/photon transport code (TORT version 3)
Energy Technology Data Exchange (ETDEWEB)
Rhoades, W.A.; Simpson, D.B.
1997-10-01
TORT calculates the flux or fluence of neutrons and/or photons throughout three-dimensional systems due to particles incident upon the system`s external boundaries, due to fixed internal sources, or due to sources generated by interaction with the system materials. The transport process is represented by the Boltzman transport equation. The method of discrete ordinates is used to treat the directional variable, and a multigroup formulation treats the energy dependence. Anisotropic scattering is treated using a Legendre expansion. Various methods are used to treat spatial dependence, including nodal and characteristic procedures that have been especially adapted to resist numerical distortion. A method of body overlay assists in material zone specification, or the specification can be generated by an external code supplied by the user. Several special features are designed to concentrate machine resources where they are most needed. The directional quadrature and Legendre expansion can vary with energy group. A discontinuous mesh capability has been shown to reduce the size of large problems by a factor of roughly three in some cases. The emphasis in this code is a robust, adaptable application of time-tested methods, together with a few well-tested extensions.
MCNP: a general Monte Carlo code for neutron and photon transport
Energy Technology Data Exchange (ETDEWEB)
Forster, R.A.; Godfrey, T.N.K.
1985-01-01
MCNP is a very general Monte Carlo neutron photon transport code system with approximately 250 person years of Group X-6 code development invested. It is extremely portable, user-oriented, and a true production code as it is used about 60 Cray hours per month by about 150 Los Alamos users. It has as its data base the best cross-section evaluations available. MCNP contains state-of-the-art traditional and adaptive Monte Carlo techniques to be applied to the solution of an ever-increasing number of problems. Excellent user-oriented documentation is available for all facets of the MCNP code system. Many useful and important variants of MCNP exist for special applications. The Radiation Shielding Information Center (RSIC) in Oak Ridge, Tennessee is the contact point for worldwide MCNP code and documentation distribution. A much improved MCNP Version 3A will be available in the fall of 1985, along with new and improved documentation. Future directions in MCNP development will change the meaning of MCNP to Monte Carlo N Particle where N particle varieties will be transported.
Analytical calculations of neutron slowing down and transport in the constant-cross-section problem
Energy Technology Data Exchange (ETDEWEB)
Cacuci, D.G.
1978-04-01
Aspects of the problem of neutron slowing down and transport in an infinite medium consisting of a single nuclide that scatters elastically and isotropically and has energy-independent cross sections were investigated. The method of singular eigenfunctions was applied to the Boltzmann Equation governing the Laplace transform (with respect to the lethargy variable) of the neutron flux. A new sufficient condition for the convergence of the coefficients of the expansion of the scattering kernel in Legendre polynomials was rigorously derived for this energy-dependent problem. Formulas were obtained for the lethargy-dependent spatial moments of the scalar flux that are valid for medium to large lethargies. Use was made of the well-known connection between the spatial moments of the Laplace-transformed scalar flux and the moments of the flux in the ''eigenvalue space.'' The calculations were aided by the construction of a closed general expression for these ''eigenvalue space'' moments. Extensive use was also made of the methods of combinatorial analysis and of computer evaluation of complicated sequences of manipulations. For the case of no absorption it was possible to obtain for materials of any atomic weight explicit corrections to the age-theory formulas for the spatial moments M/sub 2n/(u) of the scalar flux that are valid through terms of the order of u/sup -5/. The evaluation of the coefficients of the powers of n, as explicit functions of the nuclear mass, represent one of the end products of this investigation. In addition, an exact expression for the second spatial moment, M/sub 2/(u), valid for arbitrary (constant) absorption, was derived. It is now possible to calculate analytically and rigorously the ''age'' for the constant-cross-section problem for arbitrary (constant) absorption and nuclear mass. 5 figures, 1 table.
A Complex-Geometry Validation Experiment for Advanced Neutron Transport Codes
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David W. Nigg; Anthony W. LaPorta; Joseph W. Nielsen; James Parry; Mark D. DeHart; Samuel E. Bays; William F. Skerjanc
2013-11-01
The Idaho National Laboratory (INL) has initiated a focused effort to upgrade legacy computational reactor physics software tools and protocols used for support of core fuel management and experiment management in the Advanced Test Reactor (ATR) and its companion critical facility (ATRC) at the INL.. This will be accomplished through the introduction of modern high-fidelity computational software and protocols, with appropriate new Verification and Validation (V&V) protocols, over the next 12-18 months. Stochastic and deterministic transport theory based reactor physics codes and nuclear data packages that support this effort include MCNP5[1], SCALE/KENO6[2], HELIOS[3], SCALE/NEWT[2], and ATTILA[4]. Furthermore, a capability for sensitivity analysis and uncertainty quantification based on the TSUNAMI[5] system has also been implemented. Finally, we are also evaluating the Serpent[6] and MC21[7] codes, as additional verification tools in the near term as well as for possible applications to full three-dimensional Monte Carlo based fuel management modeling in the longer term. On the experimental side, several new benchmark-quality code validation measurements based on neutron activation spectrometry have been conducted using the ATRC. Results for the first four experiments, focused on neutron spectrum measurements within the Northwest Large In-Pile Tube (NW LIPT) and in the core fuel elements surrounding the NW LIPT and the diametrically opposite Southeast IPT have been reported [8,9]. A fifth, very recent, experiment focused on detailed measurements of the element-to-element core power distribution is summarized here and examples of the use of the measured data for validation of corresponding MCNP5, HELIOS, NEWT, and Serpent computational models using modern least-square adjustment methods are provided.
Dhital, Chetan
The work performed within this thesis is divided into two parts, each focusing primarily on the study of magnetic phase behavior using neutron scattering techniques. In first part, I present transport, magnetization, and neutron scattering studies of materials within the iridium oxide-based Ruddelsden-Popper series [Srn+1IrnO3n+1] compounds Sr 3Ir2O7 (n=2) and Sr2IrO4 (n=1). This includes a comprehensive study of the doped bilayer system Sr 3(Ir1-xRux )2O7. In second part, I present my studies of the effect of uniaxial pressure on magnetic and structural phase behavior of the iron-based high temperature superconductor Ba(Fe1-xCox)2As2. Iridium-based 5d transition metal oxides host rather unusual electronic/magnetic ground states due to strong interplay between electronic correlation, lattice structure and spin-orbit effects. Out of the many oxides containing iridium, the Ruddelsden-Popper series [Srn+1IrnO 3n+1] oxides are some of the most interesting systems to study both from the point of view of physics as well as from potential applications. My work is focused on two members of this series Sr3Ir2O 7 (n=2) and Sr2IrO4 (n=1). In particular, our combined transport, magnetization and neutron scattering studies of Sr 3Ir2O7 (n=2) showed that this system exhibits a complex coupling between charge transport and magnetism. The spin magnetic moments form a G-type antiferromagnetic structure with moments oriented along the c-axis, with an ordered moment of 0.35+/-0.06 muB/Ir. I also performed experiments doping holes in this bilayer Sr3(Ir1-xRu x)2O7 system in order to study the role of electronic correlation in these materials. Our results show that the ruthenium-doped holes remain localized within the Jeff=1/2 Mott insulating background of Sr3Ir2O7, suggestive of 'Mott blocking' and the presence of strong electronic correlation in these materials. Antiferromagnetic order however survives deep into the metallic regime with the same ordering q-vector, suggesting an
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Escobar, M.; Meyerovich, A. E., E-mail: Alexander-Meyerovich@uri.edu [University of Rhode Island, Department of Physics (United States)
2014-12-15
We discuss transport of particles along random rough surfaces in quantum size effect conditions. As an intriguing application, we analyze gravitationally quantized ultracold neutrons in rough waveguides in conjunction with GRANIT experiments (ILL, Grenoble). We present a theoretical description of these experiments in the biased diffusion approximation for neutron mirrors with both one- and two-dimensional (1D and 2D) roughness. All system parameters collapse into a single constant which determines the depletion times for the gravitational quantum states and the exit neutron count. This constant is determined by a complicated integral of the correlation function (CF) of surface roughness. The reliable identification of this CF is always hindered by the presence of long fluctuation-driven correlation tails in finite-size samples. We report numerical experiments relevant for the identification of roughness of a new GRANIT waveguide and make predictions for ongoing experiments. We also propose a radically new design for the rough waveguide.
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Azmy, Yousry
2014-06-10
We employ the Integral Transport Matrix Method (ITMM) as the kernel of new parallel solution methods for the discrete ordinates approximation of the within-group neutron transport equation. The ITMM abandons the repetitive mesh sweeps of the traditional source iterations (SI) scheme in favor of constructing stored operators that account for the direct coupling factors among all the cells' fluxes and between the cells' and boundary surfaces' fluxes. The main goals of this work are to develop the algorithms that construct these operators and employ them in the solution process, determine the most suitable way to parallelize the entire procedure, and evaluate the behavior and parallel performance of the developed methods with increasing number of processes, P. The fastest observed parallel solution method, Parallel Gauss-Seidel (PGS), was used in a weak scaling comparison with the PARTISN transport code, which uses the source iteration (SI) scheme parallelized with the Koch-baker-Alcouffe (KBA) method. Compared to the state-of-the-art SI-KBA with diffusion synthetic acceleration (DSA), this new method- even without acceleration/preconditioning-is completitive for optically thick problems as P is increased to the tens of thousands range. For the most optically thick cells tested, PGS reduced execution time by an approximate factor of three for problems with more than 130 million computational cells on P = 32,768. Moreover, the SI-DSA execution times's trend rises generally more steeply with increasing P than the PGS trend. Furthermore, the PGS method outperforms SI for the periodic heterogeneous layers (PHL) configuration problems. The PGS method outperforms SI and SI-DSA on as few as P = 16 for PHL problems and reduces execution time by a factor of ten or more for all problems considered with more than 2 million computational cells on P = 4.096.
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Petzoldt, R.W.; Perkins, L.J.
1995-06-16
The authors performed 1-D coupled, neutron-gamma transport calculations for lithium-vanadium blankets and lithium-sodium cauldron pot blankets in cylindrical and spherical geometries. Parametric fits to the data are supplied for subsequent use in systems code models. Scaling relationships are given for various neutronics parameters of interest, including: tritium breeding ratio, neutron energy multiplication, magnet dose rates, magnet heating rates, and integrated magnet fluence.
Pollock, Rachel A.
Mesoporous materials are interesting as catalyst supports, because molecules can move efficiently in and out of the pore network, but they must be stable in water if they are to be used for the production of biofuels. Before investigating hydrothermal stability and transport properties, the pore structure of SBA-15 was characterized using small angle neutron scattering (SANS) and non-local density functional theory (NLDFT) analysis of nitrogen sorption isotherms. A new Contrast Matching SANS method, using a range of probe molecules to directly probe the micropore size, gave a pore size distribution onset of 6 ± 0.2 Å, consistent with cylindrical pores formed from polymer template strands that unravel into the silica matrix. Diffraction intensity analysis of SANS measurements, combined with pore size distributions calculated from NLDFT, showed that the secondary pores are distributed relatively uniformly throughout the silica framework. The hydrothermal stability of SBA-15 was evaluated using a post-calcination hydrothermal treatment in both liquid and vapor phase water. The results were consistent with a degradation mechanism in which silica dissolves from regions of small positive curvature, e.g. near the entrance to the secondary pores, and is re-deposited deeper into the framework. Under water treatment at 115 °C, the mesopore diameter increases and the intra-wall void fraction decreases significantly. The behavior is similar for steam treatment, but occurs more slowly, suggesting that transport is faster when condensation occurs in the pores. Quasielastic neutron scattering (QENS) measurements of methane in SBA-15 probed the rotational and translational motion as a function of temperature and loading. A qualitative analysis of the QENS data suggested that for the initial dose of methane at 100 K, the self diffusion constant is similar in magnitude to literature values for methane in ZSM-5 and Y-zeolite, showing that the secondary pores trap methane and limit
Hoffman, Adam J.; Lee, John C.
2016-02-01
A new time-dependent Method of Characteristics (MOC) formulation for nuclear reactor kinetics was developed utilizing angular flux time-derivative propagation. This method avoids the requirement of storing the angular flux at previous points in time to represent a discretized time derivative; instead, an equation for the angular flux time derivative along 1D spatial characteristics is derived and solved concurrently with the 1D transport characteristic equation. This approach allows the angular flux time derivative to be recast principally in terms of the neutron source time derivatives, which are approximated to high-order accuracy using the backward differentiation formula (BDF). This approach, called Source Derivative Propagation (SDP), drastically reduces the memory requirements of time-dependent MOC relative to methods that require storing the angular flux. An SDP method was developed for 2D and 3D applications and implemented in the computer code DeCART in 2D. DeCART was used to model two reactor transient benchmarks: a modified TWIGL problem and a C5G7 transient. The SDP method accurately and efficiently replicated the solution of the conventional time-dependent MOC method using two orders of magnitude less memory.
Transport and mixing of r-process elements in neutron star binary merger blast waves
Montes, Gabriela; Naiman, Jill; Shen, Sijing; Lee, William H
2016-01-01
The r-process nuclei are robustly synthesized in the material ejected during a neutron star binary merger (NSBM), as tidal torques transport angular momentum and energy through the outer Lagrange point in the form of a vast tidal tail. If NSBM are indeed solely responsible for the solar system r- process abundances, a galaxy like our own would require to host a few NSBM per million years, with each event ejecting, on average, about 5x10^{-2} M_sun of r-process material. Because the ejecta velocities in the tidal tail are significantly larger than in ordinary supernovae, NSBM deposit a comparable amount of energy into the interstellar medium (ISM). In contrast to extensive efforts studying spherical models for supernova remnant evolution, calculations quantifying the impact of NSBM ejecta in the ISM have been lacking. To better understand their evolution in a cosmological context, we perform a suite of three-dimensional hydrodynamic simulations with optically-thin radiative cooling of isolated NSBM ejecta expa...
Inversion of Source and Transport Parameters of Relativistic SEPs from Neutron Monitor Data
Agueda, Neus; Bütikofer, Rolf; Vainio, Rami; Heber, Bernd; Afanasiev, Alexander; Malandraki, Olga E.
2016-04-01
We present a new methodology to study the release processes of relativistic solar energetic particles (SEPs) based on the direct inversion of Ground Level Enhancements (GLEs) observed by the worldwide network of neutron monitors (NMs). The new approach makes use of several models, including: the propagation of relativistic SEPs from the Sun to the Earth, their transport in the Earth's magnetosphere and atmosphere, as well as the detection of the nucleon component of the secondary cosmic rays by ground based NMs. The combination of these models allows us to compute the expected ground-level NM counting rates for a series of instantaneous releases from the Sun. The amplitudes of the source components are then inferred by fitting the NM observations with the modeled NM counting rate increases. Within the HESPERIA project, we will develop the first software package for the direct inversion of GLEs and we will make it freely available for the solar and heliospheric communities. Acknowledgement: This work has received funding from the European Union's Horizon 2020 research and innovation programme under grant agreement No 637324.
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Vergnaud, T.; Nimal, J.C. (CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France))
1990-01-01
The three-dimensional polycinetic Monte Carlo particle transport code TRIPOLI has been under development in the French Shielding Laboratory at Saclay since 1965. TRIPOLI-1 began to run in 1970 and became TRIPOLI-2 in 1978: since then its capabilities have been improved and many studies have been performed. TRIPOLI can treat stationary or time dependent problems in shielding and in neutronics. Some examples of solved problems are presented to demonstrate the many possibilities of the system. (author).
The application of the Monte-Carlo neutron transport code MCNP to a small "nuclear battery" system
Puigdellívol Sadurní, Roger
2009-01-01
The project consist in calculate the keff to a small nuclear battery. The code Monte- Carlo neutron transport code MCNP is used to calculate the keff. The calculations are done at the beginning of life to know the capacity of the core becomes critical in different conditions. These conditions are the study parameters that determine the criticality of the core. These parameters are the uranium enrichment, the coated particles (TRISO) packing factor and the size of the core. More...
Ding, M.; Hjelm, R.; Sussman, A. J.
2016-12-01
Low-permeability geomedia are prevalent in subsurface environments. They have become increasingly important in a wide range of applications such as CO2-sequestration, hydrocarbon recovery, enhanced geothermal systems, legacy waste stewardship, high-level radioactive waste disposal, and global security. The flow and transport characteristics of low-permeability geomedia are dictated by their exceedingly low permeability values ranging from 10-6 to 10-12 darcy with porosities dominated by nanoscale pores. Developing new characterization methods and robust computational models that allow estimation of transport properties of low-permeability geomedia has been identified as a critical basic research and technology development need for controlling subsurface and fluids flow. Due to its sensibility to hydrogen and flexible sample environment, neutron based elastic and inelastic scattering can, through various techniques, interrogate all the nanoscale pores in the sample whether they are fluid accessible or not, and readily characterize interfacial waters. In this presentation, we will present two studies revealing the effects of nanoscale pore confinement on fluid dynamics in geomedia. In one study, we use combined (ultra-small)/small-angle elastic neutron scatterings to probe nanoporous features responses in geological materials to transport processes. In the other study, incoherent inelastic neutron scattering was used to distingwish between intergranular pore water and fluid inclusion moisture in bedded rock salt, and to explore their thermal stablibility. Our work demonstrates that neutron based elastic and inelastic scatterings are techniques of choice for in situ probing hydrocarbon and water behavior in nanoporous materials, providing new insights into water-rock interaction and fluids transport in low-permeability geomaterials.
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Duerigen, Susan
2013-05-15
The superior advantage of a nodal method for reactor cores with hexagonal fuel assemblies discretized as cells consisting of equilateral triangles is its mesh refinement capability. In this thesis, a diffusion and a simplified P{sub 3} (or SP{sub 3}) neutron transport nodal method are developed based on trigonal geometry. Both models are implemented in the reactor dynamics code DYN3D. As yet, no other well-established nodal core analysis code comprises an SP{sub 3} transport theory model based on trigonal meshes. The development of two methods based on different neutron transport approximations but using identical underlying spatial trigonal discretization allows a profound comparative analysis of both methods with regard to their mathematical derivations, nodal expansion approaches, solution procedures, and their physical performance. The developed nodal approaches can be regarded as a hybrid NEM/AFEN form. They are based on the transverse-integration procedure, which renders them computationally efficient, and they use a combination of polynomial and exponential functions to represent the neutron flux moments of the SP{sub 3} and diffusion equations, which guarantees high accuracy. The SP{sub 3} equations are derived in within-group form thus being of diffusion type. On this basis, the conventional diffusion solver structure can be retained also for the solution of the SP{sub 3} transport problem. The verification analysis provides proof of the methodological reliability of both trigonal DYN3D models. By means of diverse hexagonal academic benchmark and realistic detailed-geometry full-transport-theory problems, the superiority of the SP{sub 3} transport over the diffusion model is demonstrated in cases with pronounced anisotropy effects, which is, e.g., highly relevant to the modeling of fuel assemblies comprising absorber material.
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Bal, G. [Electricite de France (EDF), Direction des Etudes et Recherches, 92 - Clamart (France)
1997-12-31
Neutron transport in nuclear reactors is well modeled by the linear Boltzmann transport equation. Its resolution is relatively easy but very expensive. To achieve whole core calculations, one has to consider simpler models, such as diffusion or homogeneous transport equations. However, the solutions may become inaccurate in particular situations (as accidents for instance). That is the reason why we wish to solve the equations on small area accurately and more coarsely on the remaining part of the core. It is than necessary to introduce some links between different discretizations or modelizations. In this note, we give some results on the coupling of different discretizations of all degrees of freedom of the integral-differential neutron transport equation (two degrees for the angular variable, on for the energy component, and two or three degrees for spatial position respectively in 2D (cylindrical symmetry) and 3D). Two chapters are devoted to the coupling of discrete ordinates methods (for angular discretization). The first one is theoretical and shows the well posing of the coupled problem, whereas the second one deals with numerical applications of practical interest (the results have been obtained from the neutron transport code developed at the R and D, which has been modified for introducing the coupling). Next, we present the nodal scheme RTN0, used for the spatial discretization. We show well posing results for the non-coupled and the coupled problems. At the end, we deal with the coupling of energy discretizations for the multigroup equations obtained by homogenization. Some theoretical results of the discretization of the velocity variable (well-posing of problems), which do not deal directly with the purposes of coupling, are presented in the annexes. (author). 34 refs.
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Teixeira, Paulo Cleber Mendonca
2002-12-01
In this study, an analytical solution of the neutron transport equation in an annular reactor is presented with a short and rotating neutron source of the type S(x) {delta} (x- Vt), where V is the speed of annular pulsed reactor. The study is an extension of a previous study by Williams [12] carried out with a pulsed source of the type S(x) {delta} (t). In the new concept of annular pulsed reactor designed to produce continuous high flux, the core consists of a subcritical annular geometry pulsed by a rotating modulator, producing local super prompt critical condition, thereby giving origin to a rotating neutron pulse. An analytical solution is obtained by opening up of the annular geometry and applying one energy group transport theory in one dimension using applied mathematical techniques of Laplace transform and Complex Variables. The general solution for the flux consists of a fundamental mode, a finite number of harmonics and a transient integral. A condition which limits the number of harmonics depending upon the circumference of the annular geometry has been obtained. Inverse Laplace transform technique is used to analyse instability condition in annular reactor core. A regenerator parameter in conjunction with perimeter of the ring and nuclear properties is used to obtain stable and unstable harmonics and to verify if these exist. It is found that the solution does not present instability in the conditions stated in the new concept of annular pulsed reactor. (author)
Imanaka, T
2001-09-01
A transport calculation of the neutrons leaked to the environment by the JCO criticality accident was carried out based on three-dimensional geometrical models of the buildings within the JCO territory. Our work started from an initial step to simulate the leakage process of neutrons from the precipitation tank, and proceeded to a step to calculate the neutron propagation throughout the JCO facilities. The total fission number during the accident in the precipitation tank was evaluated to be 2.5 x 10(18) by comparing the calculated neutron-induced activities per 235U fission with the measured values in a stainless-steel net sample taken 2 m from the precipitation tank. Shield effects by various structures within the JCO facilities were evaluated by comparing the present results with a previous calculation using two-dimensional models which suppose a point source of the fission spectrum in the air above the ground without any shield structures. The shield effect by the precipitation tank, itself, was obtained to be a factor of 3. The shield factor by the conversion building varied between 1.1 and 2, depending on the direction from the building. The shield effect by the surrounding buildings within the JCO territory was between I and 5, also depending on the direction.
Transport in fuel cells: Electrochemical impedance spectroscopy and neutron imaging studies
Aaron, Douglas Scott
This dissertation focuses on two powerful methods of performing in-situ studies of transport limitations in fuel cells. The first is electrochemical impedance spectroscopy (EIS) while the second is neutron imaging. Three fuel cell systems are studied in this work: polymer electrolyte membrane fuel cells (PEMFCs), microbial fuel cells (MFCs) and enzyme fuel cells (EFCs). The first experimental section of this dissertation focuses on application of EIS and neutron imaging to an operating PEMFC. The effects of cathode-side humidity and flow rate, as well as cell temperature and a transient response to cathode-side humidity, were studied for a PEMFC via EIS. It was found that increased air humidity in the cathode resulted in greatly reduced cathode resistance as well as a significant reduction in membrane resistance. The anode resistance was only slightly reduced in this case. Increased air flow rate was observed to have little effect on any resistance in the PEMFC, though slight reductions in both the anode and the cathode were observed. Increased cell temperature resulted in decreased cathode and anode resistances. Finally, the transient response to increased humidity exhibited unstable behavior for both the anode and the cathode resistances and the PEMFC power output. Neutron imaging allowed the calculation of water content throughout the PEMFC, showing a maximum in water content at the cathode gas diffusion layer - membrane interface. The second experimental section of this dissertation delves into the world of microbial fuel cells. Multiple long-term observations of changes in internal resistances were performed and illustrated the reduction in anode resistance as the bacterial community was established. Over this same time period, the cathode resistance was observed to have increased; these two phenomena suggest that the anode improved over time while the cathode suffered from degradation. Increased anode fluid ionic strength and flow rate both led to significant
A 2D/1D coupling neutron transport method based on the matrix MOC and NEM methods
Energy Technology Data Exchange (ETDEWEB)
Zhang, H.; Zheng, Y.; Wu, H.; Cao, L. [School of Nuclear Science and Technology, Xi' an Jiaotong University, No. 28, Xianning West Road, Xi' an, Shaanxi 710049 (China)
2013-07-01
A new 2D/1D coupling method based on the matrix MOC method (MMOC) and nodal expansion method (NEM) is proposed for solving the three-dimensional heterogeneous neutron transport problem. The MMOC method, used for radial two-dimensional calculation, constructs a response matrix between source and flux with only one sweep and then solves the linear system by using the restarted GMRES algorithm instead of the traditional trajectory sweeping process during within-group iteration for angular flux update. Long characteristics are generated by using the customization of commercial software AutoCAD. A one-dimensional diffusion calculation is carried out in the axial direction by employing the NEM method. The 2D and ID solutions are coupled through the transverse leakage items. The 3D CMFD method is used to ensure the global neutron balance and adjust the different convergence properties of the radial and axial solvers. A computational code is developed based on these theories. Two benchmarks are calculated to verify the coupling method and the code. It is observed that the corresponding numerical results agree well with references, which indicates that the new method is capable of solving the 3D heterogeneous neutron transport problem directly. (authors)
White, M C
2000-01-01
The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron tran...
Ignatovich, V K
2005-01-01
A new, algebraic, method is applied to calculation of neutron albedo from substance to check the claim that use of ultradispersive fuel and moderator of an active core can help to gain in size and mass of the reactor. In a model of isotropic distribution of incident and reflected neutrons it is shown that coherent scattering on separate grains in the case of thermal neutrons increases transport cross section negligibly, however it decreases albedo from a wall of finite thickness because of decrease of substance density. A visible increase of albedo takes place only for neutrons with wave length of the order of the size of a single grain.
Energy Technology Data Exchange (ETDEWEB)
Thomas, Sarah [University of Alabama, Birmingham; Uhoya, Walter [University of Alabama, Birmingham; Tsoi, Georgiy [University of Alabama, Birmingham; Wenger, Lowell E [University of Alabama, Birmingham; Vohra, Yogesh [University of Alabama, Birmingham; Chesnut, Gary Neal [University of Alabama, Birmingham; Weir, S. T. [Lawrence Livermore National Laboratory (LLNL); Tulk, Christopher A [ORNL; Moreira Dos Santos, Antonio F [ORNL
2012-01-01
Neutron diffraction and electrical transport measurements have been made on the heavy rare earth metal holmium at high pressures and low temperatures in order to elucidate its transition from a paramagnetic (PM) to a helical antiferromagnetic (AFM) ordered phase as a function of pressure. The electrical resistance measurements show a change in the resistance slope as the temperature is lowered through the antiferromagnetic Neel temperature. The temperature of this antiferromagnetic transition decreases from approximately 122 K at ambient pressure at a rate of -4.9 K GPa(-1) up to a pressure of 9 GPa, whereupon the PM-to-AFM transition vanishes for higher pressures. Neutron diffraction measurements as a function of pressure at 89 and 110 K confirm the incommensurate nature of the phase transition associated with the antiferromagnetic ordering of the magnetic moments in a helical arrangement and that the ordering occurs at similar pressures as determined from the resistance results for these temperatures.
Zarebanadkouki, Mohsen; Kroener, Eva; Kaestner, Anders; Carminati, Andrea
2014-10-01
Our understanding of soil and plant water relations is limited by the lack of experimental methods to measure water fluxes in soil and plants. Here, we describe a new method to noninvasively quantify water fluxes in roots. To this end, neutron radiography was used to trace the transport of deuterated water (D2O) into roots. The results showed that (1) the radial transport of D2O from soil to the roots depended similarly on diffusive and convective transport and (2) the axial transport of D2O along the root xylem was largely dominated by convection. To quantify the convective fluxes from the radiographs, we introduced a convection-diffusion model to simulate the D2O transport in roots. The model takes into account different pathways of water across the root tissue, the endodermis as a layer with distinct transport properties, and the axial transport of D2O in the xylem. The diffusion coefficients of the root tissues were inversely estimated by simulating the experiments at night under the assumption that the convective fluxes were negligible. Inverse modeling of the experiment at day gave the profile of water fluxes into the roots. For a 24-d-old lupine (Lupinus albus) grown in a soil with uniform water content, root water uptake was higher in the proximal parts of lateral roots and decreased toward the distal parts. The method allows the quantification of the root properties and the regions of root water uptake along the root systems.
Institute of Scientific and Technical Information of China (English)
张坤明; 张雄杰; 瞿金辉; 汤彬
2015-01-01
利用MCNP程序模拟研究脉冲中子－裂变中子探测铀黄饼，采用脉冲式中子源，利用氦三管中子探测器记录裂变中子，得到铀黄饼中的铀含量信息。通过对14 MeV脉冲中子源和产生的裂变中子在不同铀含量模型中的输运计算，分析了裂变中子与铀含量的关系。结果表明：利用裂变超热中子衰减时间谱，可以确定铀黄饼中的铀含量；通过对热中子衰减时间谱进行校正，可以提高铀黄饼中铀含量计算结果的准确度。%The Monte Carlo N particle transport code ( MCNP ) is used to simulate how to explore the uranium yel⁃lowcake by using the pulsed neutron⁃fission neutron ( PNFN) method. In order to obtain uranium yellowcake quan⁃titation, pulsed neutron source was used, prompt fission neutrons were detected by using the neutron detector. Un⁃der the condition of different uranium quantitation models, the transport of the 14 MeV pulsed neutron source and the released fission neutron were calculated. On the basis of these, the relationship between fission neutron and ura⁃nium quantitation was studied. The results show that using the epithermal neutron time decay spectrum, the urani⁃um yellowcake quantitation can be determined; the precision of the uranium yellowcake quantitation could be in⁃creased by the correction of thermal neutron time decay spectrum.
3-D Deep Penetration Neutron Imaging of Thick Absorgin and Diffusive Objects Using Transport Theory
Energy Technology Data Exchange (ETDEWEB)
Ragusa, Jean; Bangerth, Wolfgang
2011-08-01
here explores the inverse problem of optical tomography applied to heterogeneous domains. The neutral particle transport equation was used as the forward model for how neutral particles stream through and interact within these heterogeneous domains. A constrained optimization technique that uses Newtons method served as the basis of the inverse problem. Optical tomography aims at reconstructing the material properties using (a) illuminating sources and (b) detector readings. However, accurate simulations for radiation transport require that the particle (gamma and/or neutron) energy be appropriate discretize in the multigroup approximation. This, in turns, yields optical tomography problems where the number of unknowns grows (1) about quadratically with respect to the number of energy groups, G, (notably to reconstruct the scattering matrix) and (2) linearly with respect to the number of unknown material regions. As pointed out, a promising approach could rely on algorithms to appropriately select a material type per material zone rather than G2 values. This approach, though promising, still requires further investigation: (a) when switching from cross-section values unknowns to material type indices (discrete integer unknowns), integer programming techniques are needed since derivative information is no longer available; and (b) the issue of selecting the initial material zoning remains. The work reported here proposes an approach to solve the latter item, whereby a material zoning is proposed using one-group or few-groups transport approximations. The capabilities and limitations of the presented method were explored; they are briefly summarized next and later described in fuller details in the Appendices. The major factors that influenced the ability of the optimization method to reconstruct the cross sections of these domains included the locations of the sources used to illuminate the domains, the number of separate experiments used in the reconstruction, the
Energy Technology Data Exchange (ETDEWEB)
Filho, J. F. P. [Institute de Matematica, Estatistica e Fisica, Universidade Federal do Rio Grande, Av. Italia, s/n, 96203-900 Rio Grande, RS (Brazil); Barichello, L. B. [Institute de Matematica, Universidade Federal do Rio Grande do Sul, Av. Bento Goncalves, 9500, 91509-900 Porto Alegre, RS (Brazil)
2013-07-01
In this work, an analytical discrete ordinates method is used to solve a nodal formulation of a neutron transport problem in x, y-geometry. The proposed approach leads to an important reduction in the order of the associated eigenvalue systems, when combined with the classical level symmetric quadrature scheme. Auxiliary equations are proposed, as usually required for nodal methods, to express the unknown fluxes at the boundary introduced as additional unknowns in the integrated equations. Numerical results, for the problem defined by a two-dimensional region with a spatially constant and isotropically emitting source, are presented and compared with those available in the literature. (authors)
TARTNP: a coupled neutron--photon Monte Carlo transport code. [10-/sup 9/ to 20 MeV; in LLL FORTRAN
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Plechaty, E.F.; Kimlinger, J.R.
1976-07-04
A Monte Carlo code was written that calculates the transport of neutrons, photons, and neutron-induced photons. The cross sections of these particles are derived from TARTNP's data base, the Evaluated Nuclear Data Library. The energy range of the neutron data in the Library is 10/sup -9/ MeV to 20 MeV; the photon energy range is 1 keV to 20 MeV. One of the chief advantages of the code is its flexibility: it allows up to 17 different kinds of output to be evaluated in the same problem.
Košťál, Michal; Cvachovec, František; Milčák, Ján; Mravec, Filip
2013-05-01
The paper is intended to show the effect of a biological shielding simulator on fast neutron and photon transport in its vicinity. The fast neutron and photon fluxes were measured by means of scintillation spectroscopy using a 45×45 mm(2) and a 10×10 mm(2) cylindrical stilbene detector. The neutron spectrum was measured in the range of 0.6-10 MeV and the photon spectrum in 0.2-9 MeV. The results of the experiment are compared with calculations. The calculations were performed with various nuclear data libraries.
Heterogeneity of solid neutron-star matter: transport coefficients and neutrino emissivity
Jones, P B
2004-01-01
Calculations of weak-interaction transition rates and of nuclear formation enthalpies show that in isolated neutron stars, the solid phase, above the neutron-drip threshold, is amorphous and heterogeneous in nuclear charge. The neutrino emissivities obtained are very dependent on the effects of proton shell structure but may be several orders of magnitude larger than the electron bremsstrahlung neutrino-pair emissivity at temperatures of 10^9 K. In this phase, electrical and thermal conductivities are much smaller than for a homogeneous bcc lattice. In particular, the reduced electrical conductivity, which is also temperature-independent, must have significant consequences for the evolution of high-multipole magnetic fields in neutron stars.
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White, Morgan C. [Univ. of Florida, Gainesville, FL (United States)
2000-07-01
The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V&V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second
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Zardini, D.M.
1996-12-31
The feasibility of neutron transport problems parallel resolution by CRONOS code`s SN module is here studied. In this report we give the first data about the parallel resolution by angular variable decomposition of the transport equation. Problems about parallel resolution by spatial variable decomposition and memory stage limits are also explained here. (author). 9 refs.
Wongthai, Printip; Hagiwara, Kohei; Miyoshi, Yurika; Wiriyasermkul, Pattama; Wei, Ling; Ohgaki, Ryuichi; Kato, Itsuro; Hamase, Kenji; Nagamori, Shushi; Kanai, Yoshikatsu
2015-03-01
The efficacy of boron neutron capture therapy relies on the selective delivery of boron carriers to malignant cells. p-Boronophenylalanine (BPA), a boron delivery agent, has been proposed to be localized to cells through transporter-mediated mechanisms. In this study, we screened aromatic amino acid transporters to identify BPA transporters. Human aromatic amino acid transporters were functionally expressed in Xenopus oocytes and examined for BPA uptake and kinetic parameters. The roles of the transporters in BPA uptake were characterized in cancer cell lines. For the quantitative assessment of BPA uptake, HPLC was used throughout the study. Among aromatic amino acid transporters, ATB(0,+), LAT1 and LAT2 were found to transport BPA with Km values of 137.4 ± 11.7, 20.3 ± 0.8 and 88.3 ± 5.6 μM, respectively. Uptake experiments in cancer cell lines revealed that the LAT1 protein amount was the major determinant of BPA uptake at 100 μM, whereas the contribution of ATB(0,+) became significant at 1000 μM, accounting for 20-25% of the total BPA uptake in MCF-7 breast cancer cells. ATB(0,+), LAT1 and LAT2 transport BPA at affinities comparable with their endogenous substrates, suggesting that they could mediate effective BPA uptake in vivo. The high and low affinities of LAT1 and ATB(0,+), respectively, differentiate their roles in BPA uptake. ATB(0,+), as well as LAT1, could contribute significantly to the tumor accumulation of BPA at clinical dose.
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Gleicher, Frederick N.; Williamson, Richard L.; Ortensi, Javier; Wang, Yaqi; Spencer, Benjamin W.; Novascone, Stephen R.; Hales, Jason D.; Martineau, Richard C.
2014-10-01
The MOOSE neutron transport application RATTLESNAKE was coupled to the fuels performance application BISON to provide a higher fidelity tool for fuel performance simulation. This project is motivated by the desire to couple a high fidelity core analysis program (based on the self-adjoint angular flux equations) to a high fidelity fuel performance program, both of which can simulate on unstructured meshes. RATTLESNAKE solves self-adjoint angular flux transport equation and provides a sub-pin level resolution of the multigroup neutron flux with resonance treatment during burnup or a fast transient. BISON solves the coupled thermomechanical equations for the fuel on a sub-millimeter scale. Both applications are able to solve their respective systems on aligned and unaligned unstructured finite element meshes. The power density and local burnup was transferred from RATTLESNAKE to BISON with the MOOSE Multiapp transfer system. Multiple depletion cases were run with one-way data transfer from RATTLESNAKE to BISON. The eigenvalues are shown to agree well with values obtained from the lattice physics code DRAGON. The one-way data transfer of power density is shown to agree with the power density obtained from an internal Lassman-style model in BISON.
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Nimal, J.C.; Vergnaud, T. (CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France))
1990-01-01
This paper describes the most important features of the Monte Carlo code TRIPOLI-2. This code solves the Boltzmann equation in three-dimensional geometries for coupled neutron and gamma rays problems. A particular emphasis is devoted to the biasing techniques, which are very important for deep penetration. Future developments in TRIPOLI are described in the conclusion. (author).
3-D Deep Penetration Neutron Imaging of Thick Absorgin and Diffusive Objects Using Transport Theory
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Ragusa, Jean; Bangerth, Wolfgang
2011-08-01
here explores the inverse problem of optical tomography applied to heterogeneous domains. The neutral particle transport equation was used as the forward model for how neutral particles stream through and interact within these heterogeneous domains. A constrained optimization technique that uses Newtons method served as the basis of the inverse problem. Optical tomography aims at reconstructing the material properties using (a) illuminating sources and (b) detector readings. However, accurate simulations for radiation transport require that the particle (gamma and/or neutron) energy be appropriate discretize in the multigroup approximation. This, in turns, yields optical tomography problems where the number of unknowns grows (1) about quadratically with respect to the number of energy groups, G, (notably to reconstruct the scattering matrix) and (2) linearly with respect to the number of unknown material regions. As pointed out, a promising approach could rely on algorithms to appropriately select a material type per material zone rather than G2 values. This approach, though promising, still requires further investigation: (a) when switching from cross-section values unknowns to material type indices (discrete integer unknowns), integer programming techniques are needed since derivative information is no longer available; and (b) the issue of selecting the initial material zoning remains. The work reported here proposes an approach to solve the latter item, whereby a material zoning is proposed using one-group or few-groups transport approximations. The capabilities and limitations of the presented method were explored; they are briefly summarized next and later described in fuller details in the Appendices. The major factors that influenced the ability of the optimization method to reconstruct the cross sections of these domains included the locations of the sources used to illuminate the domains, the number of separate experiments used in the reconstruction, the
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Arreola V, G. [IPN, Escuela Superior de Fisica y Matematicas, Posgrado en Ciencias Fisicomatematicas, area en Ingenieria Nuclear, Unidad Profesional Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07730 Mexico D. F. (Mexico); Vazquez R, R.; Guzman A, J. R., E-mail: energia.arreola.uam@gmail.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)
2012-10-15
In this work a comparative analysis of the results for the neutrons dispersion in a not multiplicative semi-infinite medium is presented. One of the frontiers of this medium is located in the origin of coordinates, where a neutrons source in beam form, i.e., {mu}{omicron}=1 is also. The neutrons dispersion is studied on the statistical method of Monte Carlo and through the unidimensional transport theory and for an energy group. The application of transport theory gives a semi-analytic solution for this problem while the statistical solution for the flow was obtained applying the MCNPX code. The dispersion in light water and heavy water was studied. A first remarkable result is that both methods locate the maximum of the neutrons distribution to less than two mean free trajectories of transport for heavy water, while for the light water is less than ten mean free trajectories of transport; the differences between both methods is major for the light water case. A second remarkable result is that the tendency of both distributions is similar in small mean free trajectories, while in big mean free trajectories the transport theory spreads to an asymptote value and the solution in base statistical method spreads to zero. The existence of a neutron current of low energy and toward the source is demonstrated, in contrary sense to the neutron current of high energy coming from the own source. (Author)
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Rahnema, Farzad; Garimeela, Srinivas; Ougouag, Abderrafi; Zhang, Dingkang
2013-11-29
This project will develop a 3D, advanced coarse mesh transport method (COMET-Hex) for steady- state and transient analyses in advanced very high-temperature reactors (VHTRs). The project will lead to a coupled neutronics and thermal hydraulic (T/H) core simulation tool with fuel depletion capability. The computational tool will be developed in hexagonal geometry, based solely on transport theory without (spatial) homogenization in complicated 3D geometries. In addition to the hexagonal geometry extension, collaborators will concurrently develop three additional capabilities to increase the 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.
Košťál, Michal; Milčák, Ján; Cvachovec, František; Jánský, Bohumil; Rypar, Vojtěch; Juříček, Vlastimil; Novák, Evžen; Egorov, Alexander; Zaritskiy, Sergey
2016-02-01
A set of benchmark experiments was carried out in the full scale VVER-1000 mock-up on the reactor LR-0 in order to validate neutron transport calculation methodologies and to perform the optimization of the shape and locations of neutron flux operation monitors channels inside the shielding of the new VVER-1000 type reactors. Compared with previous experiments on the VVER-1000 mock-up on the reactor LR-0, the fast neutron spectra were measured in the extended neutron energy interval (0.1-10 MeV) and new calculations were carried out with the MCNPX code using various nuclear data libraries (ENDF/B VII.0, JEFF 3.1, JENDL 3.3, JENDL 4, ROSFOND 2009, and CENDL 3.1). Measurements and calculations were carried out at different points in the mock-up. The calculation and experimental data are compared.
Directory of Open Access Journals (Sweden)
Košťál Michal
2016-01-01
Full Text Available A set of benchmark experiments was carried out in the full scale VVER-1000 mock-up on the reactor LR-0 in order to validate neutron transport calculation methodologies and to perform the optimization of the shape and locations of neutron flux operation monitors channels inside the shielding of the new VVER-1000 type reactors. Compared with previous experiments on the VVER-1000 mock-up on the reactor LR-0, the fast neutron spectra were measured in the extended neutron energy interval (0.1–10 MeV and new calculations were carried out with the MCNPX code using various nuclear data libraries (ENDF/B VII.0, JEFF 3.1, JENDL 3.3, JENDL 4, ROSFOND 2009, and CENDL 3.1. Measurements and calculations were carried out at different points in the mock-up. The calculation and experimental data are compared.
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Girardi, E
2004-12-15
A new methodology for the solution of the neutron transport equation, based on domain decomposition has been developed. This approach allows us to employ different numerical methods together for a whole core calculation: a variational nodal method, a discrete ordinate nodal method and a method of characteristics. These new developments authorize the use of independent spatial and angular expansion, non-conformal Cartesian and unstructured meshes for each sub-domain, introducing a flexibility of modeling which is not allowed in today available codes. The effectiveness of our multi-domain/multi-method approach has been tested on several configurations. Among them, one particular application: the benchmark model of the Phebus experimental facility at Cea-Cadarache, shows why this new methodology is relevant to problems with strong local heterogeneities. This comparison has showed that the decomposition method brings more accuracy all along with an important reduction of the computer time.
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Pinchedez, K
1999-06-01
Parallel computing meets the ever-increasing requirements for neutronic computer code speed and accuracy. In this work, two different approaches have been considered. We first parallelized the sequential algorithm used by the neutronics code CRONOS developed at the French Atomic Energy Commission. The algorithm computes the dominant eigenvalue associated with PN simplified transport equations by a mixed finite element method. Several parallel algorithms have been developed on distributed memory machines. The performances of the parallel algorithms have been studied experimentally by implementation on a T3D Cray and theoretically by complexity models. A comparison of various parallel algorithms has confirmed the chosen implementations. We next applied a domain sub-division technique to the two-group diffusion Eigen problem. In the modal synthesis-based method, the global spectrum is determined from the partial spectra associated with sub-domains. Then the Eigen problem is expanded on a family composed, on the one hand, from eigenfunctions associated with the sub-domains and, on the other hand, from functions corresponding to the contribution from the interface between the sub-domains. For a 2-D homogeneous core, this modal method has been validated and its accuracy has been measured. (author)
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Thomas, Sarah [University of Alabama, Birmingham; Montgomery, Jeffrey M [University of Alabama, Birmingham; Tsoi, Georgiy [University of Alabama, Birmingham; Vohra, Yogesh [University of Alabama, Birmingham; Chesnut, Gary Neal [University of Alabama, Birmingham; Weir, S. T. [Lawrence Livermore National Laboratory (LLNL); Tulk, Christopher A [ORNL; Moreira Dos Santos, Antonio F [ORNL
2013-01-01
Neutron diffraction and electrical transport measurements have been carried out on the heavy rare-earth metal terbium at high pressures and low temperatures in order to elucidate the onset of ferromagnetic (FM) order as a function of pressure. The electrical resistance measurements show a change in slope as the temperature is lowered through the FM Curie temperature. The temperature of this FM transition decreases at a rate of-16.7 K/GPa up to a pressure of 3.6 GPa, at which point the onset of FM order is suppressed. The neutron diffraction measurements as a function of pressure at temperatures ranging from 90 to 290 K confirm that the change of slope in the resistance is associated with the FM ordering, since this occurs at pressures similar to those determined from the resistance results at these temperatures. A disappearance of FM ordering was observed as the pressure is increased above 3.6 GPa and is correlated with the phase transition from the ambient hexagonal close packed structure to an -Sm-type structure at high pressures.
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Biondo, Elliott D [ORNL; Ibrahim, Ahmad M [ORNL; Mosher, Scott W [ORNL; Grove, Robert E [ORNL
2015-01-01
Detailed radiation transport calculations are necessary for many aspects of the design of fusion energy systems (FES) such as ensuring occupational safety, assessing the activation of system components for waste disposal, and maintaining cryogenic temperatures within superconducting magnets. Hybrid Monte Carlo (MC)/deterministic techniques are necessary for this analysis because FES are large, heavily shielded, and contain streaming paths that can only be resolved with MC. The tremendous complexity of FES necessitates the use of CAD geometry for design and analysis. Previous ITER analysis has required the translation of CAD geometry to MCNP5 form in order to use the AutomateD VAriaNce reducTion Generator (ADVANTG) for hybrid MC/deterministic transport. In this work, ADVANTG was modified to support CAD geometry, allowing hybrid (MC)/deterministic transport to be done automatically and eliminating the need for this translation step. This was done by adding a new ray tracing routine to ADVANTG for CAD geometries using the Direct Accelerated Geometry Monte Carlo (DAGMC) software library. This new capability is demonstrated with a prompt dose rate calculation for an ITER computational benchmark problem using both the Consistent Adjoint Driven Importance Sampling (CADIS) method an the Forward Weighted (FW)-CADIS method. The variance reduction parameters produced by ADVANTG are shown to be the same using CAD geometry and standard MCNP5 geometry. Significant speedups were observed for both neutrons (as high as a factor of 7.1) and photons (as high as a factor of 59.6).
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Cullen, D E
1998-11-22
TART98 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input preparation, running Monte Carlo calculations, and analysis of output results. TART98 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART98 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART98 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART98 and its data files.
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Thatar Vento, V., E-mail: Vladimir.ThatarVento@gmail.com [Gerencia de Investigacion y Aplicaciones, CNEA, Av. Gral. Paz 1499 (1650), San Martin, Buenos Aires (Argentina)] [CONICET, Av. Rivadavia 1917 (1033), Ciudad Autonoma de Buenos Aires (Argentina); Bergueiro, J.; Cartelli, D. [Gerencia de Investigacion y Aplicaciones, CNEA, Av. Gral. Paz 1499 (1650), San Martin, Buenos Aires (Argentina)] [CONICET, Av. Rivadavia 1917 (1033), Ciudad Autonoma de Buenos Aires (Argentina); Valda, A.A. [Gerencia de Investigacion y Aplicaciones, CNEA, Av. Gral. Paz 1499 (1650), San Martin, Buenos Aires (Argentina)] [Escuela de Ciencia y Tecnologia, UNSAM, M. Irigoyen 3100 (1650), San Martin, Buenos Aires (Argentina); Kreiner, A.J. [Gerencia de Investigacion y Aplicaciones, CNEA, Av. Gral. Paz 1499 (1650), San Martin, Buenos Aires (Argentina)] [CONICET, Av. Rivadavia 1917 (1033), Ciudad Autonoma de Buenos Aires (Argentina)] [Escuela de Ciencia y Tecnologia, UNSAM, M. Irigoyen 3100 (1650), San Martin, Buenos Aires (Argentina)
2011-12-15
Within the frame of an ongoing project to develop a folded Tandem-Electrostatic-Quadrupole (TESQ) accelerator facility for Accelerator-Based Boron Neutron Capture Therapy (AB-BNCT), we discuss here the electrostatic design of the machine, including the accelerator tubes with electrostatic quadrupoles and the simulations for the transport and acceleration of a high intensity beam.
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Mugica R, C.A.; Valle G, E. del [IPN, ESFM, Departamento de Ingenieria Nuclear, 07738 Mexico D.F. (Mexico)]. e-mail: cmugica@ipn.mx
2005-07-01
In 2002, E. del Valle and Ernest H. Mund developed a technique to solve numerically the Neutron transport equations in discrete ordinates and hexagonal geometry using two nodal schemes type finite element weakly discontinuous denominated WD{sub 5,3} and WD{sub 12,8} (of their initials in english Weakly Discontinuous). The technique consists on representing each hexagon in the union of three rhombuses each one of which it is transformed in a square in the one that the methods WD{sub 5,3} and WD{sub 12,8} were applied. In this work they are solved the mentioned equations of transport using the same discretization technique by hexagon but using two nodal schemes type finite element strongly discontinuous denominated SD{sub 3} and SD{sub 8} (of their initials in english Strongly Discontinuous). The application in each case as well as a reference problem for those that results are provided for the effective multiplication factor is described. It is carried out a comparison with the obtained results by del Valle and Mund for different discretization meshes so much angular as spatial. (Author)
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Chang, B
2004-03-22
This paper contains three analytical solutions of transport problems which can be used to test ray-effect errors in the numerical solutions of the Boltzmann Transport Equation (BTE). We derived the first two solutions and the third was shown to us by M. Prasad. Since this paper is intended to be an internal LLNL report, no attempt was made to find the original derivations of the solutions in the literature in order to cite the authors for their work.
Trade Study for Neutron Transport at Low Earth Orbit: Adding Fidelity to DIORAMA
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McClanahan, Tucker Caden [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wakeford, Daniel Tyler [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2017-08-22
The Distributed Infrastructure Offering Real-Time Access to Modeling and Analysis (DIORAMA) software provides performance modeling capabilities of the United States Nuclear Detonation Detection System (USNDS) with a focus on the characterization of Space-Based Nuclear Detonation Detection (SNDD) instrument performance [1]. A case study was done to add the neutron propagation capabilities of DIORAMA to low earth orbit (LEO), and compare the back-calculated incident energy from the time-of- ight (TOF) spectrum with the scored incident energy spectrum. As the scoring altitude lowers, the time increase due to scattering takes up much more of the fraction of total TOF; whereas at geosynchronous earth orbit (GEO), the time increase due to scattering is a negligible fraction of the total TOF [2]. The scattering smears out the TOF enough to make the back-calculation of the initial energy spectrum from the TOF spectrum very convoluted.
Shi, Xue-Ming; Peng, Xian-Jue
2016-09-01
Fusion science and technology has made progress in the last decades. However, commercialization of fusion reactors still faces challenges relating to higher fusion energy gain, irradiation-resistant material, and tritium self-sufficiency. Fusion Fission Hybrid Reactors (FFHR) can be introduced to accelerate the early application of fusion energy. Traditionally, FFHRs have been classified as either breeders or transmuters. Both need partition of plutonium from spent fuel, which will pose nuclear proliferation risks. A conceptual design of a Fusion Fission Hybrid Reactor for Energy (FFHR-E), which can make full use of natural uranium with lower nuclear proliferation risk, is presented. The fusion core parameters are similar to those of the International Thermonuclear Experimental Reactor. An alloy of natural uranium and zirconium is adopted in the fission blanket, which is cooled by light water. In order to model blanket burnup problems, a linkage code MCORGS, which couples MCNP4B and ORIGEN-S, is developed and validated through several typical benchmarks. The average blanket energy Multiplication and Tritium Breeding Ratio can be maintained at 10 and 1.15 respectively over tens of years of continuous irradiation. If simple reprocessing without separation of plutonium from uranium is adopted every few years, FFHR-E can achieve better neutronic performance. MCORGS has also been used to analyze the ultra-deep burnup model of Laser Inertial Confinement Fusion Fission Energy (LIFE) from LLNL, and a new blanket design that uses Pb instead of Be as the neutron multiplier is proposed. In addition, MCORGS has been used to simulate the fluid transmuter model of the In-Zinerater from Sandia. A brief comparison of LIFE, In-Zinerater, and FFHR-E will be given.
Design of a high-current low-energy beam transport line for an intense D-T/D-D neutron generator
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Lu, Xiaolong, E-mail: luxl@lzu.edu.cn [School of Nuclear Science and Technology, Lanzhou University, Lanzhou 730000 (China); Engineering Research Center for Neutron Application, Ministry of Education, Lanzhou University, Lanzhou 730000 (China); Wang, Junrun [School of Nuclear Science and Technology, Lanzhou University, Lanzhou 730000 (China); Engineering Research Center for Neutron Application, Ministry of Education, Lanzhou University, Lanzhou 730000 (China); Zhang, Yu; Li, Jianyi; Xia, Li; Zhang, Jie; Ding, Yanyan; Jiang, Bing; Huang, Zhiwu; Ma, Zhanwen; Wei, Zheng [School of Nuclear Science and Technology, Lanzhou University, Lanzhou 730000 (China); Qian, Xiangping; Xu, Dapeng; Lan, Changlin [School of Nuclear Science and Technology, Lanzhou University, Lanzhou 730000 (China); Engineering Research Center for Neutron Application, Ministry of Education, Lanzhou University, Lanzhou 730000 (China); Yao, Zeen, E-mail: zeyao@lzu.edu.cn [School of Nuclear Science and Technology, Lanzhou University, Lanzhou 730000 (China); Engineering Research Center for Neutron Application, Ministry of Education, Lanzhou University, Lanzhou 730000 (China)
2016-03-01
An intense D-T/D-D neutron generator is currently being developed at the Lanzhou University. The Cockcroft–Walton accelerator, as a part of the neutron generator, will be used to accelerate and transport the high-current low-energy beam from the duoplasmatron ion source to the rotating target. The design of a high-current low-energy beam transport (LEBT) line and the dynamics simulations of the mixed beam were carried out using the TRACK code. The results illustrate that the designed beam line facilitates smooth transportation of a deuteron beam of 40 mA, and the number of undesired ions can be reduced effectively using two apertures.
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Kosaka, Shinya; Saji, Etsuro [In-Core Fuel Management System Department, Toden Software, Inc., Tokyo (Japan)
2000-12-01
A characteristics transport theory code, CHAPLET, has been developed for the purpose of making it practical to perform a whole LWR core calculation with the same level of calculational model and accuracy as that of an ordinary single assembly calculation. The characteristics routine employs the CACTUS algorithm for drawing ray tracing lines, which assists the two key features of the flux solution in the CHAPLET code. One is the direct neutron path linking (DNPL) technique which strictly connects angular fluxes at each assembly interface in the flux solution separated between assemblies. Another is to reduce the required memory storage by sharing the data related to ray tracing among assemblies with the same configuration. For faster computation, the coarse mesh rebalance (CMR) method and the Aitken method were incorporated in the code and the combined use of both methods showed the most promising acceleration performance among the trials. In addition, the parallelization of the flux solution was attempted, resulting in a significant reduction in the wall-clock time of the calculation. By all these efforts, coupled with the results of many verification studies, a whole LWR core heterogeneous transport theory calculation finally became practical. CHAPLET is thought to be a useful tool which can produce the reference solutions for analyses of an LWR (author)
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Fournier, D.; Le Tellier, R.; Suteau, C., E-mail: damien.fournier@cea.fr, E-mail: romain.le-tellier@cea.fr, E-mail: christophe.suteau@cea.fr [CEA, DEN, DER/SPRC/LEPh, Cadarache, Saint Paul-lez-Durance (France); Herbin, R., E-mail: raphaele.herbin@cmi.univ-mrs.fr [Laboratoire d' Analyse et de Topologie de Marseille, Centre de Math´ematiques et Informatique (CMI), Universit´e de Provence, Marseille Cedex (France)
2011-07-01
The solution of the time-independent neutron transport equation in a deterministic way invariably consists in the successive discretization of the three variables: energy, angle and space. In the SNATCH solver used in this study, the energy and the angle are respectively discretized with a multigroup approach and the discrete ordinate method. A set of spatial coupled transport equations is obtained and solved using the Discontinuous Galerkin Finite Element Method (DGFEM). Within this method, the spatial domain is decomposed into elements and the solution is approximated by a hierarchical polynomial basis in each one. This approach is time and memory consuming when the mesh becomes fine or the basis order high. To improve the computational time and the memory footprint, adaptive algorithms are proposed. These algorithms are based on an error estimation in each cell. If the error is important in a given region, the mesh has to be refined (h−refinement) or the polynomial basis order increased (p−refinement). This paper is related to the choice between the two types of refinement. Two ways to estimate the error are compared on different benchmarks. Analyzing the differences, a hp−refinement method is proposed and tested. (author)
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Kramer, K J; Latkowski, J F; Abbott, R P; Boyd, J K; Powers, J J; Seifried, J E
2008-10-24
Lawrence Livermore National Laboratory is currently developing a hybrid fusion-fission nuclear energy system, called LIFE, to generate power and burn nuclear waste. We utilize inertial confinement fusion to drive a subcritical fission blanket surrounding the fusion chamber. It is composed of TRISO-based fuel cooled by the molten salt flibe. Low-yield (37.5 MJ) targets and a repetition rate of 13.3 Hz produce a 500 MW fusion source that is coupled to the subcritical blanket, which provides an additional gain of 4-8, depending on the fuel. In the present work, we describe the neutron transport and nuclear burnup analysis. We utilize standard analysis tools including, the Monte Carlo N-Particle (MCNP) transport code, ORIGEN2 and Monteburns to perform the nuclear design. These analyses focus primarily on a fuel composed of depleted uranium not requiring chemical reprocessing or enrichment. However, other fuels such as weapons grade plutonium and highly-enriched uranium are also under consideration. In addition, we have developed a methodology using {sup 6}Li as a burnable poison to replace the tritium burned in the fusion targets and to maintain constant power over the lifetime of the engine. The results from depleted uranium analyses suggest up to 99% burnup of actinides is attainable while maintaining full power at 2GW for more than five decades.
Directory of Open Access Journals (Sweden)
Das Basanta Kumar
2014-01-01
Full Text Available The study of an ion beam transport mechanism contributes to the production of a good quality ion beam with a higher current and better beam emittance. The simulation of an ion beam provides the basis for optimizing the extraction system and the acceleration gap for the ion source. In order to extract an ion beam from an ion source, a carefully designed electrode system for the required beam energy must be used. In our case, a self-extracted penning ion source is used for ion generation, extraction and acceleration with a single accelerating gap for the production of neutrons. The characteristics of the ion beam extracted from this ion source were investigated using computer code SIMION 8.0. The ion trajectories from different locations of the plasma region were investigated. The simulation process provided a good platform for a study on optimizing the extraction and focusing system of the ion beam transported to the required target position without any losses and provided an estimation of beam emittance.
Energy Technology Data Exchange (ETDEWEB)
Ahnert, C.; Aragones, J. M.
1981-07-01
This Is a users manual of the neutron transport code TWOTRAN-TRACA, which is a version of the original TWOTRAN-GG from the Los Alamos Laboratory, with some modifications made at JEN. A detailed input data description is given as well as the new modifications developed at JEN. (Author) 8 refs.
Energy Technology Data Exchange (ETDEWEB)
Frankel, J.I.
1997-09-01
This investigation used sysmbolic manipulation in developing analytical methods and general computational strategies for solving both linear and nonlinear, regular and singular integral and integro-differential equations which appear in radiative and mixed-mode energy transport. Contained in this report are seven papers which present the technical results as individual modules.
Zheng, Weixiong
2016-01-01
In this paper, we present an accurate and robust scaling operator based on material optical thickness (OT) for the least-squares spherical harmonics (LSP$_N$) method for solving neutron transport problems. LSP$_N$ without proper scaling is known to be erroneous in highly scattering medium, if the optical thickness of the material is large. A previously presented scaling developed by Manteuffel, et al.\\ does improve the accuracy of LSP$_N$, in problems where the material is optically thick. With the method, however, essentially no scaling is applied in optically thin materials, which can lead to an erroneous solution with presence of highly scattering medium. Another scaling approach, called the reciprocal-removal (RR) scaled LSP$_N$, which is equivalent to the self-adjoint angular flux (SAAF) equation, has numerical issues in highly-scattering materials due to a singular weighting. We propose a scaling based on optical thickness that improves the solution in optically thick media while avoiding the singularit...
Limits on the Efficiency of Event-Based Algorithms for Monte Carlo Neutron Transport
Energy Technology Data Exchange (ETDEWEB)
Romano, Paul K.; Siegel, Andrew R.
2017-04-16
The traditional form of parallelism in Monte Carlo particle transport simulations, wherein each individual particle history is considered a unit of work, does not lend itself well to data-level parallelism. Event-based algorithms, which were originally used for simulations on vector processors, may offer a path toward better utilizing data-level parallelism in modern computer architectures. In this study, a simple model is developed for estimating the efficiency of the event-based particle transport algorithm under two sets of assumptions. Data collected from simulations of four reactor problems using OpenMC was then used in conjunction with the models to calculate the speedup due to vectorization as a function of two parameters: the size of the particle bank and the vector width. When each event type is assumed to have constant execution time, the achievable speedup is directly related to the particle bank size. We observed that the bank size generally needs to be at least 20 times greater than vector size in order to achieve vector efficiency greater than 90%. When the execution times for events are allowed to vary, however, the vector speedup is also limited by differences in execution time for events being carried out in a single event-iteration. For some problems, this implies that vector effciencies over 50% may not be attainable. While there are many factors impacting performance of an event-based algorithm that are not captured by our model, it nevertheless provides insights into factors that may be limiting in a real implementation.
Energy Technology Data Exchange (ETDEWEB)
Fournier, D.
2011-10-10
The different neutronic parameters have to be calculated with a higher accuracy in order to design the 4. generation reactor cores. As memory storage and computation time are limited, adaptive methods are a solution to solve the neutron transport equation. The neutronic flux, solution of this equation, depends on the energy, angle and space. The different variables are successively discretized. The energy with a multigroup approach, considering the different quantities to be constant on each group, the angle by a collocation method called SN approximation. Once the energy and angle variable are discretized, a system of spatially-dependent hyperbolic equations has to be solved. Discontinuous finite elements are used to make possible the development of hp-refinement methods. Thus, the accuracy of the solution can be improved by spatial refinement (h-refinement), consisting into subdividing a cell into sub-cells, or by order refinement (p-refinement), by increasing the order of the polynomial basis. In this thesis, the properties of this methods are analyzed showing the importance of the regularity of the solution to choose the type of refinement. Thus, two error estimators are used to lead the refinement process. Whereas the first one requires high regularity hypothesis (analytical solution), the second one supposes only the minimal hypothesis required for the solution to exist. The comparison of both estimators is done on benchmarks where the analytic solution is known by the method of manufactured solutions. Thus, the behaviour of the solution as a regard of the regularity can be studied. It leads to a hp-refinement method using the two estimators. Then, a comparison is done with other existing methods on simplified but also realistic benchmarks coming from nuclear cores. These adaptive methods considerably reduces the computational cost and memory footprint. To further improve these two points, an approach with energy-dependent meshes is proposed. Actually, as the
Energy Technology Data Exchange (ETDEWEB)
Burns, Kimberly A. [Georgia Inst. of Technology, Atlanta, GA (United States)
2009-08-01
The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples.
Energy Technology Data Exchange (ETDEWEB)
Zwermann, W.; Aures, A.; Bernnat, W.; and others
2013-06-15
This report documents the status of the research and development goals reached within the reactor safety research project RS1503 ''Development and Application of Neutron Transport Methods and Uncertainty Analyses for Reactor Core Calculations'' as of the 1{sup st} quarter of 2013. The superordinate goal of the project is the development, validation, and application of neutron transport methods and uncertainty analyses for reactor core calculations. These calculation methods will mainly be applied to problems related to the core behaviour of light water reactors and innovative reactor concepts. The contributions of this project towards achieving this goal are the further development, validation, and application of deterministic and stochastic calculation programmes and of methods for uncertainty and sensitivity analyses, as well as the assessment of artificial neutral networks, for providing a complete nuclear calculation chain. This comprises processing nuclear basis data, creating multi-group data for diffusion and transport codes, obtaining reference solutions for stationary states with Monte Carlo codes, performing coupled 3D full core analyses in diffusion approximation and with other deterministic and also Monte Carlo transport codes, and implementing uncertainty and sensitivity analyses with the aim of propagating uncertainties through the whole calculation chain from fuel assembly, spectral and depletion calculations to coupled transient analyses. This calculation chain shall be applicable to light water reactors and also to innovative reactor concepts, and therefore has to be extensively validated with the help of benchmarks and critical experiments.
Verma, Rishi; Mishra, Ekansh; Dhang, Prosenjit; Sagar, Karuna; Meena, Manraj; Shyam, Anurag
2016-09-01
The results of characterization experiments carried out on a newly developed dense plasma focus device based intense pulsed neutron source with efficient and compact pulsed power system are reported. Its high current sealed pseudospark switch based low inductance capacitor bank with maximum stored energy of ˜10 kJ is segregated into four modules of ˜2.5 kJ each and it cumulatively delivers peak current in the range of 400 kA-600 kA (corresponding to charging voltage range of 14 kV-18 kV) in a quarter time period of ˜2 μs. The neutron yield performance of this device has been optimized by discretely varying deuterium filling gas pressure in the range of 6 mbar-11 mbar at ˜17 kV/550 kA discharge. At ˜7 kJ/8.5 mbar operation, the average neutron yield has been measured to be in the order of ˜4 × 109 neutrons/pulse which is the highest ever reported neutron yield from a plasma focus device with the same stored energy. The average forward to radial anisotropy in neutron yield is found to be ˜2. The entire system is contained on a moveable trolley having dimensions 1.5 m × 1 m × 0.7 m and its operation and control (up to the distance of 25 m) are facilitated through optically isolated handheld remote console. The overall compactness of this system provides minimum proximity to small as well as large samples for irradiation. The major intended application objective of this high neutron yield dense plasma focus device development is to explore the feasibility of active neutron interrogation experiments by utilization of intense pulsed neutron sources.
S. Z. Kalantari; H Tavakoli; Nami, M.
2015-01-01
One of the important neutron sources for Boron Neutron Capture Therapy (BNCT) is a nuclear reactor. It needs a high flux of epithermal neutrons. The optimum conditions of the neutron spectra for BNCT are provided by the International Atomic Energy Agency (IAEA). In this paper, Miniature Neutron Source Reactor (MNSR) as a neutron source for BNCT was investigated. For this purpose, we designed a Beam Shaping Assembly (BSA) for the reactor and the neutron transport from the core of the reactor t...
Energy Technology Data Exchange (ETDEWEB)
Tomaschewski, Fernanda K.; Segatto, Cynthia F., E-mail: fernandasls_89@hotmail.com, E-mail: cynthia.segatto@ufrgs.br [Universidade Federal do Rio Grande do Sul (UFRS), Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Matematica Aplicada; Barros, Ricardo C., E-mail: rcbarros@pq.cnpq.br [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Departamento de Modelagem Computacional
2015-07-01
Presented here is a decomposition method based on series representation of the group angular fluxes and delayed neutron precursors in smoothly continuous functions for energy multigroups, slab-geometry discrete ordinates kinetics equations supplemented with a prescribed number of delayed neutron precursors. Numerical results to a non-reflected sub-critical slab stabilized by steady-state sources are given to illustrate the accuracy and efficiency of the o offered method. (author)
Energy Technology Data Exchange (ETDEWEB)
Fevotte, F. [CEA Saclay, Dept. Modelisation de Systemes et Structures (DEN/DANS/DM2S/SERMA), 91 - Gif sur Yvette (France)
2008-07-01
At the various stages of a nuclear reactor's life, numerous studies are needed to guaranty the safety and efficiency of the design, analyse the fuel cycle, prepare the dismantlement, and so on. Due to the extreme difficulty to take extensive and accurate measurements in the reactor core, most of these studies are numerical simulations. The complete numerical simulation of a nuclear reactor involves many types of physics: neutronics, thermal hydraulics, materials, control engineering, Among these, the neutron transport simulation is one of the fundamental steps, since it allows computation - among other things - of various fundamental values such as the power density (used in thermal hydraulics computations) or fuel burn-up. The neutron transport simulation is based on the Boltzmann equation, which models the neutron population inside the reactor core. Among the various methods allowing its numerical solution, much interest has been devoted in the past few years to the Method of Characteristics in unstructured meshes (MOC), since it offers a good accuracy and operates in complicated geometries. The aim of this work is to propose improvements of the calculation scheme bound on the two dimensions MOC, in order to decrease the needed resources number. (A.L.B.)
Rubinson, Kenneth A; Faraone, Antonio
2016-05-14
X-ray and neutron scattering have been used to provide insight into the structures of ionic solutions for over a century, but the probes have covered distances shorter than 8 Å. For the non-hydrolyzing salt SrI2 in aqueous solution, a locally ordered lattice of ions exists that scatters slow neutrons coherently down to at least 0.1 mol L(-1) concentration, where the measured average distance between scatterers is over 18 Å. To investigate the motions of these scatterers, coherent quasielastic neutron scattering (CQENS) data on D2O solutions with SrI2 at 1, 0.8, 0.6, and 0.4 mol L(-1) concentrations was obtained to provide an experimental measure of the diffusive transport rate for the motion between pairs of ions relative to each other. Because CQENS measures the motion of one ion relative to another, the frame of reference is centered on an ion, which is unique among all diffusion measurement methods. We call the measured quantity the pairwise diffusive transport rate Dp. In addition to this ion centered frame of reference, the diffusive transport rate can be measured as a function of the momentum transfer q, where q = (4π/λ)sin θ with a scattering angle of 2θ. Since q is related to the interion distance (d = 2π/q), for the experimental range 0.2 Å(-1)≤q≤ 1.0 Å(-1), Dp is, then, measured over interion distances from 40 Å to ≈6 Å. We find the measured diffusional transport rates increase with increasing distance between scatterers over the entire range covered and interpret this behavior to be caused by dynamic coupling among the ions. Within the model of Fickian diffusion, at the longer interionic distances Dp is greater than the Nernst-Hartley value for an infinitely dilute solution. For these nm-distance diffusional transport rates to conform with the lower, macroscopically measured diffusion coefficients, we propose that local, coordinated counter motion of at least pairs of ions is part of the transport process.
Indian Academy of Sciences (India)
Charles Simon; Alain Pautrat; Christophe Goupil; Joseph Scola; Patrice Mathieu; Annie Brûlet; Antoine Ruyter; M J Higgins; Shobo Bhattacharya; D Plessis
2006-01-01
The existence of a peak effect in transport properties (a maximum of the critical current as function of magnetic field) is a well-known but still intriguing feature of Type II superconductors such as NbSe2 and Bi-2212. Using a model of pinning by surface irregularities in anisotropic superconductors, we have developed a calculation of the critical current which allows estimating quantitatively the critical current in both the high critical current phase and the low critical current phase. The only adjustable parameter of this model is the angle of the vortices at the surface. The agreement between the measurements and the model is really very impressive. In this framework, the anomalous dynamical properties close to the peak effect is due to coexistence of two different vortex states with different critical currents. Recent neutron diffraction data in NbSe2 crystals in the presence of transport current support this point of view.
Energy Technology Data Exchange (ETDEWEB)
Basso Barichello, Liliane; Dias da Cunha, Rudnei [Universidade Federal do Rio Grande do Sul, Porto Alegre, RS (Brazil). Inst. de Matematica; Becker Picoloto, Camila [Universidade Federal do Rio Grande do Sul, Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica; Tres, Anderson [Universidade Federal do Rio Grande do Sul, Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Matematica Aplicada
2015-05-15
A nodal formulation of a fixed-source two-dimensional neutron transport problem, in Cartesian geometry, defined in a heterogeneous medium, is solved by an analytical approach. Explicit expressions, in terms of the spatial variables, are derived for averaged fluxes in each region in which the domain is subdivided. The procedure is an extension of an analytical discrete ordinates method, the ADO method, for the solution of the two-dimensional homogeneous medium case. The scheme is developed from the discrete ordinates version of the two-dimensional transport equation along with the level symmetric quadrature scheme. As usual for nodal schemes, relations between the averaged fluxes and the unknown angular fluxes at the contours are introduced as auxiliary equations. Numerical results are in agreement with results available in the literature.
Energy Technology Data Exchange (ETDEWEB)
Pazianotto, Mauricio Tizziani; Goncalez, Odair Lelis; Federico, Claudio Antonio [Centro Tecnico Aeroespacial (IEAv/CTA), Sao Jose dos Campos, SP (Brazil). Inst. de Estudos Avancados; Carlson, Brett Vern [Centro Tecnico Aeroespacial (ITA/CTA), Sao Jose dos Campos, SP (Brazil). Inst. Tecnologico de Aeronautica
2010-07-01
Full text: The Institute for Advanced Studies (IEAv) is developing activities to study the dose levels of ionizing radiation from cosmic rays (CR) received by aircraft crews, sensitive equipment (on-board computers, for example) and embedded electronics in Brazilian airspace. Neutrons generated by the interaction of CR with the atmosphere are the dominant particles in the dose accumulation in electronic circuits and aircraft crews at flight altitude. Their production has a very broad energy spectrum, ranging from thermal neutrons (0.025eV ) to neutrons of several hundreds of MeV , making their detection a very difficult process. To observe the temporal variation in flow during the measurements, a detector of the Long Counter (LC) type is being used. This detector is designed to measure the one-way flow of neutrons with constant response over a wide energy range (thermal to 20 MeV ). However, to measure cosmic rays, the flow of which is non-directional, the dependence of the response on the angle of incidence, as well as energy, should be properly investigated. The objective of this study is to assess the angular response of the neutron detector (Long Counter) using the code MCNP5 (Monte Carlo N-Particle) and to compare it with the experimental data previously obtained with a {sup 241}Am-Be source at a distance of 1.66 m from the geometric center of the detector, varying the angle of incidence from 00 to 3600 in intervals of 150. The simulation was performed by modeling in detail the structure and materials of the LC, as well as the experimental arrangement for irradiation. The results of the simulation present reasonable agreement with the experimental data. This agreement shows that the modeling of the geometry of the source-detector system is adequate. The next step is to develop a model of neutron detection for the higher energy present in cosmic radiation fields, for which the experimental calibration is not so easily achievable. (author)
Lucero, Catherine L.; Bentz, Dale P.; Hussey, Daniel S.; Jacobson, David L.; Weiss, W. Jason
Air entrainment is commonly added to concrete to help in reducing the potential for freeze thaw damage. It is hypothesized that the entrained air voids remain unsaturated or partially saturated long after the smaller pores fill with water. Small gel and capillary pores in the cement matrix fill quickly on exposure to water, but larger pores (entrapped and entrained air voids) require longer times or other methods to achieve saturation. As such, it is important to quantitatively determine the water content and degree of saturation in air entrained cementitious materials. In order to further investigate properties of cement-based mortar, a model based on Beer's Law has been developed to interpret neutron radiographs. This model is a powerful tool for analyzing images acquired from neutron radiography. A mortar with a known volume of aggregate, water to cement ratio and degree of hydration can be imaged and the degree of saturation can be estimated.
Indian Academy of Sciences (India)
V Wagner; A Krása; M Majerla; F Křížek; O Svoboda; A Kugler; J Adam; V M Tsoupko-Sitnikov; M I Krivopustov; I V Zhuk; W Westmeier
2007-02-01
The set-up `energy plus transmutation', consisting of a thick lead target and a natural uranium blanket, was irradiated by relativistic proton beams with the energy from 0.7 GeV up to 2 GeV. Neutron field was measured in different places of this set-up using different activation detectors. The possibilities of using the obtained data for benchmark studies are analyzed in this paper. Uncertainties of experimental data are shown and discussed. The experimental data are compared with results of simulation with MCNPX code.
Iwamoto, Yosuke; Ogawa, Tatsuhiko
2017-04-01
Because primary knock-on atoms (PKAs) create point defects and clusters in materials that are irradiated with neutrons, it is important to validate the calculations of recoil cross section spectra that are used to estimate radiation damage in materials. Here, the recoil cross section spectra of fission- and fusion-relevant materials were calculated using the Event Generator Mode (EGM) of the Particle and Heavy Ion Transport code System (PHITS) and also using the data processing code NJOY2012 with the nuclear data libraries TENDL2015, ENDF/BVII.1, and JEFF3.2. The heating number, which is the integral of the recoil cross section spectra, was also calculated using PHITS-EGM and compared with data extracted from the ACE files of TENDL2015, ENDF/BVII.1, and JENDL4.0. In general, only a small difference was found between the PKA spectra of PHITS + TENDL2015 and NJOY + TENDL2015. From analyzing the recoil cross section spectra extracted from the nuclear data libraries using NJOY2012, we found that the recoil cross section spectra were incorrect for 72Ge, 75As, 89Y, and 109Ag in the ENDF/B-VII.1 library, and for 90Zr and 55Mn in the JEFF3.2 library. From analyzing the heating number, we found that the data extracted from the ACE file of TENDL2015 for all nuclides were problematic in the neutron capture region because of incorrect data regarding the emitted gamma energy. However, PHITS + TENDL2015 can calculate PKA spectra and heating numbers correctly.
Energy Technology Data Exchange (ETDEWEB)
Iwamoto, Yosuke, E-mail: iwamoto.yosuke@jaea.go.jp; Ogawa, Tatsuhiko
2017-04-01
Because primary knock-on atoms (PKAs) create point defects and clusters in materials that are irradiated with neutrons, it is important to validate the calculations of recoil cross section spectra that are used to estimate radiation damage in materials. Here, the recoil cross section spectra of fission- and fusion-relevant materials were calculated using the Event Generator Mode (EGM) of the Particle and Heavy Ion Transport code System (PHITS) and also using the data processing code NJOY2012 with the nuclear data libraries TENDL2015, ENDF/BVII.1, and JEFF3.2. The heating number, which is the integral of the recoil cross section spectra, was also calculated using PHITS-EGM and compared with data extracted from the ACE files of TENDL2015, ENDF/BVII.1, and JENDL4.0. In general, only a small difference was found between the PKA spectra of PHITS + TENDL2015 and NJOY + TENDL2015. From analyzing the recoil cross section spectra extracted from the nuclear data libraries using NJOY2012, we found that the recoil cross section spectra were incorrect for {sup 72}Ge, {sup 75}As, {sup 89}Y, and {sup 109}Ag in the ENDF/B-VII.1 library, and for {sup 90}Zr and {sup 55}Mn in the JEFF3.2 library. From analyzing the heating number, we found that the data extracted from the ACE file of TENDL2015 for all nuclides were problematic in the neutron capture region because of incorrect data regarding the emitted gamma energy. However, PHITS + TENDL2015 can calculate PKA spectra and heating numbers correctly.
Using MCNP in the design of neutron sources and neutron beams
Energy Technology Data Exchange (ETDEWEB)
Hergenreder, Daniel F.; Lecot, Carlos A.; Lovotti, Osvaldo P. [INVAP S.A., San Carlos de Bariloche (Argentina). Nuclear Projects Department. Nuclear Engineering Division
2002-07-01
The calculation methodology used to design cold, thermal and hot neutron sources and their associated neutron beam transport systems is presented. The design goal is to evaluate the performance of the neutron sources, their beam tubes and neutron guides at specific experimental locations in the reactor hall as well as in the neutron hall. The Monte Carlo method is a unique and powerful tool to transport neutrons. Its use in a bootstrap scheme appears to be an appropriate solution for this type of system. The proper use of MCNP as the main tool leads to a fast and reliable method to perform calculations in a relatively short time with low statistical errors. (author)
General Design for CARR Neutron Guide System
Institute of Scientific and Technical Information of China (English)
无
2011-01-01
A neutron guide system has been designed and partly installed at the China Advanced Research Reactor (CARR) to transport cold neutrons from the cold neutron source (CNS) to several instruments,which are situated in a separate guide hall of 30 m×60 m.
Energy Technology Data Exchange (ETDEWEB)
Masiello, E
2006-07-01
The principal goal of this manuscript is devoted to the investigation of a new type of heterogeneous mesh adapted to the shape of the fuel pins (fuel-clad-moderator). The new heterogeneous mesh guarantees the spatial modelling of the pin-cell with a minimum of regions. Two methods are investigated for the spatial discretization of the transport equation: the discontinuous finite element method and the method of characteristics for structured cells. These methods together with the new representation of the pin-cell result in an appreciable reduction of calculation points. They allow an exact modelling of the fuel pin-cell without spatial homogenization. A new synthetic acceleration technique based on an angular multigrid is also presented for the speed up of the inner iterations. These methods are good candidates for transport calculations for a nuclear reactor core. A second objective of this work is the application of method of characteristics for non-structured geometries to the study of double heterogeneity problem. The letters is characterized by fuel material with a stochastic dispersion of heterogeneous grains, and until now was solved with a model based on collision probabilities. We propose a new statistical model based on renewal-Markovian theory, which makes possible to take into account the stochastic nature of the problem and to avoid the approximations of the collision probability model. The numerical solution of this model is guaranteed by the method of characteristics. (author)
Constitutive laws for the neutron density current
Energy Technology Data Exchange (ETDEWEB)
Espinosa-Paredes, Gilberto [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Av. San Rafael Atlixco 186 Col. Vicentina, Mexico, D.F., 09340 (Mexico)], E-mail: gepe@xanum.uam.mx; Morales-Sandoval, Jaime B. [Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Mor. (Mexico); Vazquez-Rodriguez, Rodolfo [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Av. San Rafael Atlixco 186 Col. Vicentina, Mexico, D.F., 09340 (Mexico); Espinosa-Martinez, Erick-G. [Retorno Quebec 6, Col. Burgos de Cuernavaca 62580, Temixco, Mor. (Mexico)
2008-10-15
In this technical note, a fractional wave equation for the average neutron motion in nuclear reactor is considered. This representation covers the full spectrum of the average neutron transport behavior, i.e., Fickian and non-Fickian effects. The fractional diffusion model retains the main dynamic characteristics of the neutron motion in which the relaxation time associated with a rapid variation in the neutron flux contains a fractional exponent that can be manipulated to obtain the best representation of the neutron transport phenomena. The detrended fluctuation analysis (DFA) method is presented in this paper to estimate the fractional exponent.
Energy Technology Data Exchange (ETDEWEB)
Palau, J.M. [CEA Cadarache, Service de Physique des Reacteurs et du Cycle, Lab. de Projets Nucleaires, 13 - Saint-Paul-lez-Durance (France)
2005-07-01
This paper presents how Monte-Carlo calculations (French TRIPOLI4 poly-kinetic code with an appropriate pre-processing and post-processing software called OVNI) are used in the case of 3-dimensional heterogeneous benchmarks (slab reactor cores) to reduce model biases and enable a thorough and detailed analysis of the performances of deterministic methods and their associated data libraries with respect to key neutron parameters (reactivity, local power). Outstanding examples of application of these tools are presented regarding the new numerical methods implemented in the French lattice code APOLLO2 (advanced self-shielding models, new IDT characteristics method implemented within the discrete-ordinates flux solver model) and the JEFF3.1 nuclear data library (checked against JEF2.2 previous file). In particular we have pointed out, by performing multigroup/point-wise TRIPOLI4 (assembly and core) calculations, the efficiency (in terms of accuracy and computation time) of the new IDT method developed in APOLLO2. In addition, by performing 3-dimensional TRIPOLI4 calculations of the whole slab core (few millions of elementary volumes), the high quality of the new JEFF3.1 nuclear data files and revised evaluations (U{sup 235}, U{sup 238}, Hf) for reactivity prediction of slab cores critical experiments has been stressed. As a feedback of the whole validation process, improvements in terms of nuclear data (mainly Hf capture cross-sections) and numerical methods (advanced quadrature formulas accounting validation results, validation of new self-shielding models, parallelization) are suggested to improve even more the APOLLO2-CRONOS2 standard calculation route. (author)
Manuel, Oliver K.
2011-01-01
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...
Neutron logging tool readings and neutron parameters of formations
Czubek, Jan A.
1995-03-01
A case history of the calibration of neutron porosity tools is given in the paper. The calibration of neutron porosity tools is one of the most difficult, complicated, and time consuming tasks in the well logging operations in geophysics. A semi empirical approach to this problem is given in the paper. It is based on the correlation of the tool readings observed in known environments with the apparent neutron parameters sensed by the tools. The apparent neutron parameters are functions of the true neutron parameters of geological formations and of the borehole material, borehole diameter, and the tool position inside the borehole. The true integral neutron transport parameters are obtained by the multigroup diffusion approximation for slowing down of neutrons and by one thermal neutron group for the diffusion. In the latter, the effective neutron temperature is taken into account. The problem of the thermal neutron absorption cross section of rocks is discussed in detail from the point of view of its importance for the well logging results and for the experimental techniques being used.
Owens, A. R.; Welch, J. A.; Kópházi, J.; Eaton, M. D.
2016-06-01
In this paper two discontinuous Galerkin isogeometric analysis methods are developed and applied to the first-order form of the neutron transport equation with a discrete ordinate (SN) angular discretisation. The discontinuous Galerkin projection approach was taken on both an element level and the patch level for a given Non-Uniform Rational B-Spline (NURBS) patch. This paper describes the detailed dispersion analysis that has been used to analyse the numerical stability of both of these schemes. The convergence of the schemes for both smooth and non-smooth solutions was also investigated using the method of manufactured solutions (MMS) for multidimensional problems and a 1D semi-analytical benchmark whose solution contains a strongly discontinuous first derivative. This paper also investigates the challenges posed by strongly curved boundaries at both the NURBS element and patch level with several algorithms developed to deal with such cases. Finally numerical results are presented both for a simple pincell test problem as well as the C5G7 quarter core MOX/UOX small Light Water Reactor (LWR) benchmark problem. These numerical results produced by the isogeometric analysis (IGA) methods are compared and contrasted against linear and quadratic discontinuous Galerkin finite element (DGFEM) SN based methods.
Energy Technology Data Exchange (ETDEWEB)
Magat, Ph
1997-04-01
Today neutron transport in PWR's core is routinely computed through the transport-diffusion(2 groups) scheme. This method gives satisfactory results for reactors operating in normal conditions but the 2 group diffusion approximation is unable to take into account interface effects or anisotropy. The improvement of this scheme is logically possible through the use of a simplified P{sub N} method (SP{sub N}) for the modeling of the core. The comparison between S{sub N} calculations and SP{sub N} calculations shows an excellent agreement on eigenvalues as well as on power maps. We can notice that: -) it is no use extending the development beyond P{sub 3}, there is no effect; -) the P{sub 1} development is adequate; and -) the P{sub 0} development is totally inappropriate. Calculations performed on the N4 core of the Chooz power plant have enabled us to compare diffusion operators with transport operators (SP{sub 1}, SP{sub 3}, SP{sub 5} and SP{sub 7}). These calculations show that the implementation of the SP{sub N} method is feasible but the extra-costs in computation times and memory are important. We recommend: SP{sub 5}P{sub 1} calculations for heterogeneous 2-dimension geometry and SP{sub 3}P{sub 1} calculations for the homogeneous 3-dimension geometry. (A.C.)
Energy Technology Data Exchange (ETDEWEB)
Valle G, E. del; Mugica R, C.A. [IPN, ESFM, Departamento de Ingenieria Nuclear, 07738 Mexico D.F. (Mexico)]. e-mail: cmugica@ipn.mx
2005-07-01
In our country, in last congresses, Gomez et al carried out reactivity calculations based on the solution of the diffusion equation for an energy group using nodal methods in one dimension and the TPL approach (Lineal Perturbation Theory). Later on, Mugica extended the application to the case of multigroup so much so much in one as in two dimensions (X Y geometry) with excellent results. Presently work is carried out similar calculations but this time based on the solution of the neutron transport equation in X Y geometry using nodal methods and again the TPL approximation. The idea is to provide a calculation method that allows to obtain in quick form the reactivity solving the direct problem as well as the enclosed problem of the not perturbed problem. A test problem for the one that results are provided for the effective multiplication factor is described and its are offered some conclusions. (Author)
Energy Technology Data Exchange (ETDEWEB)
Oujidi, B.
1996-09-19
The TDT code solves the multigroup transport equation by the interface current method for unstructured 2D geometries. This works presents the extension of TDT to the treatment of 3D geometries obtained by axial displacement of unstructured 2D geometries. Three-dimensional trajectories are obtained by lifting the 2D trajectories. The code allows for the definition of macro-domains in the axial direction to be used in the interface-current method. Specular and isotropic reflection or translations boundary conditions can be applied to the horizontal boundaries of the domain. Numerical studies have shown the need for longer trajectory cutoffs for trajectories intersecting horizontal boundaries. Numerical applications to the calculation of local power peaks are given in a second part for: the local destruction of a Pyrex absorbent and inter-assembly (UO{sub 2}-MOX) power distortion due to pellet collapsing at the top of the core. Calculations with 16 groups were performed by coupling TDT to the spectral code APOLLO2. One-group comparisons with the Monte Carlo code TRIMARAN2 are also given. (author). 30 refs.
Energy Technology Data Exchange (ETDEWEB)
Oujidi, B
1996-09-19
The TDT code solves the multigroup transport equation by the interface-current method for unstructured 2D geometries. This works presents the extension of TDT to the treatment of 3D geometries obtained by axial displacement of unstructured 2D geometries. Three-dimensional trajectories are obtained by lifting the 2D trajectories. The code allows for the definition of macro-domains in the axial direction to be used in interface-current method. Specular and isotropic reflection or translations boundary conditions can be applied to the horizontal boundaries of the domain. Numerical studies have shown the need for longer trajectory cutoffs for trajectories intersecting horizontal boundaries. Numerical applications to the calculation of local power peaks are given in a second part for: the local destruction of a Pyrex absorbent, inter-assembly (U02-MOX) power distortion due to pellet collapsing at the top of the core. Calculations with 16 groups were performed by coupling TDT to the spectral code APOLLO2. One-group comparisons with the Monte Carlo code TRIMARAN2 are also given. (author) 30 refs.
Heller, A. K.; Brenizer, J. S.
Neutron radiography and its related two-dimensional (2D) neutron imaging techniques have been established as invaluable nondestructive inspection methods and quantitative measurement tools. They have been used in a wide variety of applications ranging from inspection of aircraft engine turbine blades to study of two-phase fluid flow in operating proton exchange membrane fuel cells. Neutron radiography is similar to X-ray radiography in that the method produces a 2D attenuation map of neutron radiation that has penetrated the object being examined. However, the images produced differ and are often complementary due to the differences between X-ray and neutron interaction mechanisms. The uses and types of 2D neutron imaging have expanded over the past 15 years as a result of advances in imaging technology and improvements in neutron generators/sources and computers. Still, high-intensity sources such as those from reactors and spallation neutron sources, together with conventional film radiography, remain the mainstay of high-resolution, large field-of-view neutron imaging. This chapter presents a summary of the history, methods, and related variations of neutron radiography techniques.
Wu, Yican
2017-01-01
This book provides a systematic and comprehensive introduction to fusion neutronics, covering all key topics from the fundamental theories and methodologies, as well as a wide range of fusion system designs and experiments. It is the first-ever book focusing on the subject of fusion neutronics research. Compared with other nuclear devices such as fission reactors and accelerators, fusion systems are normally characterized by their complex geometry and nuclear physics, which entail new challenges for neutronics such as complicated modeling, deep penetration, low simulation efficiency, multi-physics coupling, etc. The book focuses on the neutronics characteristics of fusion systems and introduces a series of theories and methodologies that were developed to address the challenges of fusion neutronics, and which have since been widely applied all over the world. Further, it introduces readers to neutronics design’s unique principles and procedures, experimental methodologies and technologies for fusion systems...
Performance of an elliptically tapered neutron guide
Mühlbauer, Sebastian; Stadlbauer, Martin; Böni, Peter; Schanzer, Christan; Stahn, Jochen; Filges, Uwe
2006-11-01
Supermirror coated neutron guides are used at all modern neutron sources for transporting neutrons over large distances. In order to reduce the transmission losses due to multiple internal reflection of neutrons, ballistic neutron guides with linear tapering have been proposed and realized. However, these systems suffer from an inhomogeneous illumination of the sample. Moreover, the flux decreases significantly with increasing distance from the exit of the neutron guide. We propose using elliptically tapered guides that provide a more homogeneous phase space at the sample position as well as a focusing at the sample. Moreover, the design of the guide system is simplified because ellipses are simply defined by their long and short axes. In order to prove the concept we have manufactured a doubly focusing guide and investigated its properties with neutrons. The experiments show that the predicted gains using the program package McStas are realized. We discuss several applications of elliptic guides in various fields of neutron physics.
Performance of an elliptically tapered neutron guide
Energy Technology Data Exchange (ETDEWEB)
Muehlbauer, Sebastian [Physik-Department E21, Technische Universitaet Muenchen, D-85747 Garching (Germany)]. E-mail: sebastian.muehlbauer@frm2.tum.de; Stadlbauer, Martin [Physik-Department E21, Technische Universitaet Muenchen, D-85747 Garching (Germany); Boeni, Peter [Physik-Department E21, Technische Universitaet Muenchen, D-85747 Garching (Germany); Schanzer, Christan [Labor fuer Neutronenstreuung, Paul Scherrer Institut, CH-5232 Villingen PSI (Switzerland); Stahn, Jochen [Labor fuer Neutronenstreuung, Paul Scherrer Institut, CH-5232 Villingen PSI (Switzerland); Filges, Uwe [Labor fuer Neutronenstreuung, Paul Scherrer Institut, CH-5232 Villingen PSI (Switzerland)
2006-11-15
Supermirror coated neutron guides are used at all modern neutron sources for transporting neutrons over large distances. In order to reduce the transmission losses due to multiple internal reflection of neutrons, ballistic neutron guides with linear tapering have been proposed and realized. However, these systems suffer from an inhomogeneous illumination of the sample. Moreover, the flux decreases significantly with increasing distance from the exit of the neutron guide. We propose using elliptically tapered guides that provide a more homogeneous phase space at the sample position as well as a focusing at the sample. Moreover, the design of the guide system is simplified because ellipses are simply defined by their long and short axes. In order to prove the concept we have manufactured a doubly focusing guide and investigated its properties with neutrons. The experiments show that the predicted gains using the program package McStas are realized. We discuss several applications of elliptic guides in various fields of neutron physics.
Energy Technology Data Exchange (ETDEWEB)
Gerhard Strydom; Cristian Rabiti; Andrea Alfonsi
2012-10-01
PHISICS is a neutronics code system currently under development at the Idaho National Laboratory (INL). Its goal is to provide state of the art simulation capability to reactor designers. The different modules for PHISICS currently under development are a nodal and semi-structured transport core solver (INSTANT), a depletion module (MRTAU) and a cross section interpolation (MIXER) module. The INSTANT module is the most developed of the mentioned above. Basic functionalities are ready to use, but the code is still in continuous development to extend its capabilities. This paper reports on the effort of coupling the nodal kinetics code package PHISICS (INSTANT/MRTAU/MIXER) to the thermal hydraulics system code RELAP5-3D, to enable full core and system modeling. This will enable the possibility to model coupled (thermal-hydraulics and neutronics) problems with more options for 3D neutron kinetics, compared to the existing diffusion theory neutron kinetics module in RELAP5-3D (NESTLE). In the second part of the paper, an overview of the OECD/NEA MHTGR-350 MW benchmark is given. This benchmark has been approved by the OECD, and is based on the General Atomics 350 MW Modular High Temperature Gas Reactor (MHTGR) design. The benchmark includes coupled neutronics thermal hydraulics exercises that require more capabilities than RELAP5-3D with NESTLE offers. Therefore, the MHTGR benchmark makes extensive use of the new PHISICS/RELAP5-3D coupling capabilities. The paper presents the preliminary results of the three steady state exercises specified in Phase I of the benchmark using PHISICS/RELAP5-3D.
Neutron dosimetry; Dosimetria de neutrons
Energy Technology Data Exchange (ETDEWEB)
Fratin, Luciano
1993-12-31
A neutron irradiation facility was designed and built in order to establish a procedure for calibrating neutron monitors and dosemeters. A 185 GBq {sup 241} Am Be source of known is used as a reference source. The irradiation facility using this source in the air provides neutron dose rates between 9 nSv s{sup -1} and 0,5 {sup {mu}}Sv s{sup -1}. A calibrated 50 nSv s{sup -1} thermal neutron field is obtained by using a specially designed paraffin block in conjunction with the {sup 241} Am Be source. A Bonner multisphere spectrometer was calibrated, using a procedure based on three methods proposed by international standards. The unfold {sup 241} Am Be neutron spectrum was determined from the Bonner spheres data and resulted in a good agreement with expected values for fluence rate, dose rate and mean energy. A dosimetric system based on the electrochemical etching of CR-39 was developed for personal dosimetry. The dosemeter badge using a (n,{alpha}) converter, the etching chamber and high frequency power supply were designed and built specially for this project. The electrochemical etching (ECE) parameters used were: a 6N KOH solution, 59 deg C, 20 kV{sub pp} cm{sup -1}, 2,0 kHz, 3 hours of ECE for thermal and intermediate neutrons and 6 hours for fast neutrons. The calibration factors for thermal, intermediate and fast neutrons were determined for this personal dosemeter. The sensitivities determined for the developed dosimetric system were (1,46{+-} 0,09) 10{sup 4} tracks cm{sup -2} mSv{sup -1} for thermal neutrons, (9{+-}3) 10{sup 2} tracks cm{sup -2} mSV{sup -1} for intermediate neutrons and (26{+-}4) tracks cm{sup -2} mSv{sup -1} for fast neutrons. The lower and upper limits of detection were respectively 0,002 mSv and 0,6 mSv for thermal neutrons, 0,04 mSv and 8 mSv for intermediate neutrons and 1 mSv and 12 mSv for fast neutrons. In view of the 1990`s ICRP recommendations, it is possible to conclude that the personal dosemeter described in this work is
Energy Technology Data Exchange (ETDEWEB)
Ondis, L.A., II; Tyburski, L.J.; Moskowitz, B.S.
2000-03-01
The RCP01 Monte Carlo program is used to analyze many geometries of interest in nuclear design and analysis of light water moderated reactors such as the core in its pressure vessel with complex piping arrangement, fuel storage arrays, shipping and container arrangements, and neutron detector configurations. Written in FORTRAN and in use on a variety of computers, it is capable of estimating steady state neutron or photon reaction rates and neutron multiplication factors. The energy range covered in neutron calculations is that relevant to the fission process and subsequent slowing-down and thermalization, i.e., 20 MeV to 0 eV. The same energy range is covered for photon calculations.
DEFF Research Database (Denmark)
Klösgen-Buchkremer, Beate Maria
2014-01-01
films or films with magnetic properties. The reason is the peculiar property of neutron light since the mass of a neutron is close to the one of a proton, and since it bears a magnetic moment. The optical properties of matter, when interacting with neutrons, are described by a refractive index......Neutron (and X-ray) reflectometry constitute complementary interfacially sensitive techniques that open access to studying the structure within thin films of both soft and hard condensed matter. Film thickness starts oxide surfaces on bulk substrates, proceeding to (pauci-)molecular layers and up...... to hundreds of nanometers. Thickness resolution for flat surfaces is in the range of few Ǻngstrøm, and as a peculiar benefit, the presence and properties of buried interfaces are accessible. Focus here will be on neutron reflectometry, a technique that is unique in applications involving composite organic...
Advanced Neutron Source (ANS) Project
Campbell, J. H.
1992-01-01
This report discusses the following about the Advanced Neutron Source: Project Management; Research and Development; Fuel Development; Corrosion Loop Tests and Analyses; Thermal-Hydraulic Loop Tests; Reactor Control and Shutdown 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; Design; and Safety.
Neutron Transmission through Sapphire Crystals
DEFF Research Database (Denmark)
Sapphire crystals are excellent filters of fast neutrons, while at the same time exhibit moderate to very little absorption at smaller energies. We have performed an extensive series of measurements in order to quantify the above effect. Alongside our experiments, we have performed a series...... of simulations, in order to reproduce the transmission of cold neutrons through sapphire crystals. Those simulations were part of the effort of validating and improving the newly developed interface between the Monte-Carlo neutron transport code MCNP and the Monte Carlo ray-tracing code McStas....
Neutron scattering and hydrogen storage
Directory of Open Access Journals (Sweden)
A.J. Ramirez-Cuesta
2009-11-01
Full Text Available Hydrogen has been identified as a fuel of choice for providing clean energy for transport and other applications across the world and the development of materials to store hydrogen efficiently and safely is crucial to this endeavour. Hydrogen has the largest scattering interaction with neutrons of all the elements in the periodic table making neutron scattering ideal for studying hydrogen storage materials. Simultaneous characterisation of the structure and dynamics of these materials during hydrogen uptake is straightforward using neutron scattering techniques. These studies will help us to understand the fundamental properties of hydrogen storage in realistic conditions and hence design new hydrogen storage materials.
Neutron transport in random media
Energy Technology Data Exchange (ETDEWEB)
Makai, M. [KFKI Atomic Energy Research Institute, Budapest (Hungary)
1996-08-01
The survey reviews the methods available in the literature which allow a discussion of corium recriticality after a severe accident and a characterization of the corium. It appears that to date no one has considered the eigenvalue problem, though for the source problem several approaches have been proposed. The mathematical formulation of a random medium may be approached in different ways. Based on the review of the literature, we can draw three basic conclusions. The problem of static, random perturbations has been solved. The static case is tractable by the Monte Carlo method. There is a specific time dependent case for which the average flux is given as a series expansion.
Calculations of Neutron and y Transport through Rare-earth Doped Polymer%中子、γ射线在稀土-高分子材料中的输运
Institute of Scientific and Technical Information of China (English)
呼延雪莹; 胡碧涛
2011-01-01
A series of shielding analyses have been performed to estimate the material composition and optimum thickness required for a new radiation shield with various rare-earth doped polymer and heavy metal mixtures.The neutron and y photon fluxes have been calculated by Monte Carlo N-Particle(MCNP) transport code.The results indicate that the relative fluxes of y photon and neutron in both traditional and new composite materials follow an exponential decay rule with the distance of penetration.It can be seen that the composite material consisting of rare-earth doped polymer and heavy metal has stronger neutron shielding performance than lead-boron polyethylene,but weaker y shielding effectiveness than W-Ni alloy.It isalso found that materials with more components of rare earth elements don't always provide better neutron shielding performance.%为研究新型复合屏蔽材料的最佳厚度与各种成分最佳配比,用MCNP计算了中子、γ射线在稀土-高分子与重金属复合材料中的通量.对中子、γ射线在屏蔽体中变化规律进行了深入探索,同传统复合屏蔽材料的屏蔽性能进行了对比.结果表明,中子和γ射线通过屏蔽体时,其强度遵循指数衰减规律.新型屏蔽材料对中子的屏蔽效果均优于铅硼聚乙烯,对γ射线的屏蔽效果均劣于W-Ni合金,且并非稀土含量越高,材料对中子辐射屏蔽能力越强.
Energy Technology Data Exchange (ETDEWEB)
ElAgib, I. [College of Science, King Saud University, P.O. Box 2455 (Saudi Arabia)], E-mail: elagib@ksu.edu.sa; Elsheikh, N. [College of Applied and Industrial Science, University of Juba, Khartoum, P.O. Box 321 (Sudan); AlSewaidan, H. [College of Science, King Saud University, P.O. Box 2455 (Saudi Arabia); Habbani, F. [Faculty of Science, Physics Department, University of Khartoum, Khartoum, P.O. Box 321 (Sudan)
2009-01-15
Calculations of elastically backscattered (EBS) neutrons from hidden explosives buried in soil were performed using Monte-Carlo N-particle transport code MCNP5. Three different neutron sources were used in the study. The study re-examines the performance of the neutron backscattering methods in providing identification of hidden explosives through their chemical composition. The EBS neutron energy spectra of fast and slow neutrons of the major constituent elements in soil and an explosive material in form of TNT have shown definite structures that can be used for the identification of a buried landmine.
Cyclotron-based neutron source for BNCT
Mitsumoto, T.; Yajima, S.; Tsutsui, H.; Ogasawara, T.; Fujita, K.; Tanaka, H.; Sakurai, Y.; Maruhashi, A.
2013-04-01
Kyoto University Research Reactor Institute (KURRI) and Sumitomo Heavy Industries, Ltd. (SHI) have developed a cyclotron-based neutron source for Boron Neutron Capture Therapy (BNCT). It was installed at KURRI in Osaka prefecture. The neutron source consists of a proton cyclotron named HM-30, a beam transport system and an irradiation & treatment system. In the cyclotron, H- ions are accelerated and extracted as 30 MeV proton beams of 1 mA. The proton beams is transported to the neutron production target made by a beryllium plate. Emitted neutrons are moderated by lead, iron, aluminum and calcium fluoride. The aperture diameter of neutron collimator is in the range from 100 mm to 250 mm. The peak neutron flux in the water phantom is 1.8×109 neutrons/cm2/sec at 20 mm from the surface at 1 mA proton beam. The neutron source have been stably operated for 3 years with 30 kW proton beam. Various pre-clinical tests including animal tests have been done by using the cyclotron-based neutron source with 10B-p-Borono-phenylalanine. Clinical trials of malignant brain tumors will be started in this year.
Cyclotron-based neutron source for BNCT
Energy Technology Data Exchange (ETDEWEB)
Mitsumoto, T.; Yajima, S.; Tsutsui, H.; Ogasawara, T.; Fujita, K. [Sumitomo Heavy Industries, Ltd (Japan); Tanaka, H.; Sakurai, Y.; Maruhashi, A. [Kyoto University Research Reactor Institute (Japan)
2013-04-19
Kyoto University Research Reactor Institute (KURRI) and Sumitomo Heavy Industries, Ltd. (SHI) have developed a cyclotron-based neutron source for Boron Neutron Capture Therapy (BNCT). It was installed at KURRI in Osaka prefecture. The neutron source consists of a proton cyclotron named HM-30, a beam transport system and an irradiation and treatment system. In the cyclotron, H- ions are accelerated and extracted as 30 MeV proton beams of 1 mA. The proton beams is transported to the neutron production target made by a beryllium plate. Emitted neutrons are moderated by lead, iron, aluminum and calcium fluoride. The aperture diameter of neutron collimator is in the range from 100 mm to 250 mm. The peak neutron flux in the water phantom is 1.8 Multiplication-Sign 109 neutrons/cm{sup 2}/sec at 20 mm from the surface at 1 mA proton beam. The neutron source have been stably operated for 3 years with 30 kW proton beam. Various pre-clinical tests including animal tests have been done by using the cyclotron-based neutron source with {sup 10}B-p-Borono-phenylalanine. Clinical trials of malignant brain tumors will be started in this year.
Low energy neutron propagation in MCNPX and GEANT4
Lemrani, R; Gerbier, G; Kudryavtsev, V A; Robinson, M; Spooner, N J C
2006-01-01
Simulations of neutron background from rock for underground experiments are presented. Neutron propagation through two types of rock, lead and hydrocarbon material is discussed. The results show a reasonably good agreement between GEANT4, MCNPX and GEANT3 in transporting low-energy neutrons.
Energy Technology Data Exchange (ETDEWEB)
Fevotte, F
2008-10-15
In the past years, the Method of Characteristics (MOC) has become a popular tool for the numerical solution of the neutron transport equation. Among its most interesting advantages are its good precision over computing time ratio, as well as its ability to accurately describe complicated geometries using non structured meshes. In order to reduce the need for computing resources in the method of characteristics, we propose in this dissertation two lines of improvement. The first axis of development is based on an analysis of the transverse integration technique in the method of characteristics. Various limitations have been discerned in this regard, which we intend to correct by proposing a new variant of the method of characteristics. Through a better treatment of material discontinuities in the geometry, our aim is to increase the accuracy of the transverse integration formula in order to decrease the computing resources without sacrificing the quality of the results. This method has been numerically tested in order to show its interest. Analysing the numerical results obtained with this new method also allows better understanding of the transverse integration approximations. Another improvement comes from the observation that industrial reactor cores exhibit very complex structures, but are often partly composed of a lattice of geometrically identical cells or assemblies. We propose a systematic method taking advantage of repetitions in the geometry to reduce the storage requirements for geometric data. Based on the group theory, this method can be employed for all lattice geometries. We present some numerical results showing the interest of the method in industrial contexts. (author)
Manuel, Oliver K
2011-01-01
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...
Energy Technology Data Exchange (ETDEWEB)
Heger, G. [Rheinisch-Westfaelische Technische Hochschule Aachen, Inst. fuer Kristallographie, Aachen (Germany)
1996-12-31
X-ray diffraction using conventional laboratory equipment and/or synchrotron installations is the most important method for structure analyses. The purpose of this paper is to discuss special cases, for which, in addition to this indispensable part, neutrons are required to solve structural problems. Even though the huge intensity of modern synchrotron sources allows in principle the study of magnetic X-ray scattering the investigation of magnetic structures is still one of the most important applications of neutron diffraction. (author) 15 figs., 1 tab., 10 refs.
Neutronics of pulsed spallation neutron sources
Watanabe, N
2003-01-01
Various topics and issues on the neutronics of pulsed spallation neutron sources, mainly for neutron scattering experiments, are reviewed to give a wide circle of readers a better understanding of these sources in order to achieve a high neutronic performance. Starting from what neutrons are needed, what the spallation reaction is and how to produce slow-neutrons more efficiently, the outline of the target and moderator neutronics are explained. Various efforts with some new concepts or ideas have already been devoted to obtaining the highest possible slow-neutron intensity with desired pulse characteristics. This paper also reviews the recent progress of such efforts, mainly focused on moderator neutronics, since moderators are the final devices of a neutron source, which determine the source performance. Various governing parameters for neutron-pulse characteristics such as material issues, geometrical parameters (shape and dimensions), the target-moderator coupling scheme, the ortho-para-hydrogen ratio, po...
The measurement of neutron and neutron induced photon spectra in fusion reactor related assemblies
Unholzer, S; Klein, H; Seidel, K
2002-01-01
The spectral neutron and photon fluence (or flux) measured outside and inside of assemblies related to fusion reactor constructions are basic quantities of fusion neutronics. The comparison of measured spectra with the results of MCNP neutron and photon transport calculations allows a crucial test of evaluated nuclear data as generally used in fusion applications to be carried out. The experiments concern mixed neutron/photon fields with about the same intensity of the two components. An NE-213 scintillation spectrometer, well described by response matrices for both neutrons and photons, is used as proton-recoil and Compton spectrometer. The experiments described here in more detail address the background problematic of two applications, an iron benchmark experiment with an ns-pulsed neutron source and a deep penetration mock-up experiment for the investigation of the ITER in-board shield system. The measured spectral neutron and photon fluences are compared with spectra calculated with the MCNP code on the b...
Energy Technology Data Exchange (ETDEWEB)
Martinez, Aquilino Senra; Silva, Fernando Carvalho da; Cardoso, Carlos Eduardo Santos [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear
2000-07-01
In some applications of perturbation theory, it is necessary know the adjoint neutron flux, which is obtained by the solution of adjoint neutron diffusion equation. However, the multigroup constants used for this are weighted in only the direct neutron flux, from the solution of direct P1 equations. In this work, this procedure is questioned and the adjoint P1 equations are derived by the neutron transport equation, the reversion operators rules and analogies between direct and adjoint parameters. (author)
Neutron noise computation using panda deterministic code
Energy Technology Data Exchange (ETDEWEB)
Humbert, Ph. [CEA Bruyeres le Chatel (France)
2003-07-01
PANDA is a general purpose discrete ordinates neutron transport code with deterministic and non deterministic applications. In this paper we consider the adaptation of PANDA to stochastic neutron counting problems. More specifically we consider the first two moments of the count number probability distribution. In a first part we will recall the equations for the single neutron and source induced count number moments with the corresponding expression for the excess of relative variance or Feynman function. In a second part we discuss the numerical solution of these inhomogeneous adjoint time dependent transport coupled equations with discrete ordinate methods. Finally, numerical applications are presented in the third part. (author)
Crump, James C., III; Richards, Wade J.; Shields, Kevin C.
1995-07-01
The McClellan Nuclear Radiation Center's (MNRC) staff in conjunction with a Cooperative Research and Development Agreement (CRDA) with the U.C. Santa Barbara facility has developed a system that can be used for aircraft inspection of jet engine blades. The problem was to develop an inspection system that can detect very low concentrations of hydrogen (i.e., greater than 100 ppm) in metal matricies. Specifically in Titanium alloy jet engine blades. Entrapment and precipitation of hydrogen in metals is an undesirable phenomenon which occurs in many alloys of steel and titanium. In general, metals suffer a loss of mechanical properties after long exposures to hydrogen, especially at high temperatures and pressures, thereby becoming embrittled. Neutron radiography has been used as a nondestructive testing technique for many years. Neutrons, because of their unique interactions with materials, are especially useful in the detection of hydrogen. They have an extremely high interaction cross section for low atomic number nuclei (i.e., hydrogen). Thus hydrogen in a metal matrix can be visualized using neutrons. Traditional radiography is sensitive to the total attenuation integrated over the path of radiation through the material. Increased sensitivity and quantitative cross section resolution can be obtained using three-dimensional volumetric imaging techniques such as tomography. The solution used to solve the problem was to develop a neutron tomography system. The neutron source is the McClellan Nuclear Radiation Center's 1 MW TRIGA reactor. This paper describes the hardware used in the system as well as some of the preliminary results.
DEFF Research Database (Denmark)
Klösgen-Buchkremer, Beate Maria
2014-01-01
to hundreds of nanometers. Thickness resolution for flat surfaces is in the range of few Ǻngstrøm, and as a peculiar benefit, the presence and properties of buried interfaces are accessible. Focus here will be on neutron reflectometry, a technique that is unique in applications involving composite organic...... of desired information. In the course, an introduction into the method and an overview on selected instruments at large scale facilities will be presented. Examples will be given that illustrate the potential of the method, mostly based on organic films. Results from the investigation of layered films...... and the detection on nanoscopic roughnesses will be shown. The potential of neutron reflectometry is not only of academic origin. It may turn out to be useful in the design and development of new functional materials even though it will never develop into a standard method to be applied in the product control...
Li, Guanjia; Ma, Zhongjian; Guo, Siming; Yan, Mingyang; Shi, Haoyu; Xu, Chao
2015-01-01
This paper described a measurement of accelerator neutron radiation field at a transport beam line of Beijing-TBF. The experiment place was be selected around a Faraday Cup with a graphite target impacted by electron beam at 2.5GeV. First of all, we simulated the neutron radiation experiment by FLUKA. Secondly, we chose six appropriate ERNMS according to their neutron fluence response function to measure the neutron count rate. Then the U_M_G package program was be utilized to unfolding experiment data. Finally, we drew a comparison between the unfolding with the simulation spectrum and made an analysis about the result.
Cooling neutrons using non-dispersive magnetic excitations
Zimmer, Oliver
2014-01-01
A new method is proposed for cooling neutrons by inelastic magnetic scattering in weakly absorbing, cold paramagnetic systems. Kinetic neutron energy is removed in constant decrements determined by the Zeeman energy of paramagnetic atoms or ions in an external magnetic field, or by zero-field level splittings in magnetic molecules. Analytical solutions of the stationary neutron transport equation are given using inelastic neutron scattering cross sections derived in an appendix. They neglect any inelastic process except the paramagnetic scattering and hence still underestimate very-cold neutron densities. Molecular oxygen with its triplet ground state appears particularly promising, notably as a host in fully deuterated oxygen-clathrate hydrate, or more exotically, in dry oxygen-He4 van der Waals clusters. At a neutron temperature about 6 K, for which neutron conversion to ultra-cold neutrons by single-phonon emission in pure superfluid He4 works best, conversion rates due to paramagnetic scattering in the cl...
Neutron shielding performance of water-extended polyester
Energy Technology Data Exchange (ETDEWEB)
Vega Carrillo, H.R.; Manzanares-Acuna, E.; Hernandez-Davila, V.M. [Zacatecas Univ. Autonoma, Nuclear Studies (Mexico); Vega Carrillo, H.R.; Hernandez-Davila, V.M. [Zacatecas Univ. Autonoma, Electric Engineering Academic Units (Mexico); Gallego, E.; Lorente, A. [Madrid Univ. Politecnica, cNuclear Engineering Department (Mexico)
2006-07-01
A Monte Carlo study to determine the shielding features to neutrons of water-extended polyester (WEP) was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through elastic and inelastic collisions. In addition to neutron attenuation properties, other desirable properties for neutron shielding materials include mechanical strength, stability, low cost, and ease of handling. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide induced by neutron activation must be considered. In this investigation the Monte Carlo method (MCNP code) was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a {sup 252}Cf isotopic neutron source, for comparison the calculations were extended to water shielding, the bare source in vacuum and in air. (authors)
Spallation Neutron Source (SNS)
Federal Laboratory Consortium — The SNS at Oak Ridge National Laboratory is a next-generation spallation neutron source for neutron scattering that is currently the most powerful neutron source in...
Energy Technology Data Exchange (ETDEWEB)
Shin, Chang Hwan; Seo, Kyong Won; Chun, Tae Hyun; Kim, Kang Seog
2005-03-15
Code coupling activities have so far focused on coupling the neutronics modules with the CFD module. An interface module for the CFD-ACE/DeCART coupling was established as an alternative to the original STAR-CD/DeCART interface. The interface module for DeCART/CFD-ACE was validated by single-pin model. The optimized CFD mesh was decided through the calculation of multi-pin model. It was important to consider turbulent mixing of subchannels for calculation of fuel temperature. For the parallel calculation, the optimized decompose process was necessary to reduce the calculation costs and setting of the iteration and convergence criterion for each code was important, too.
Doa, Changwoo; Stanley, Christopher; Gallmeier, Franz X; Doucet, Mathieu; Smith, Gregory S
2013-01-01
It is generally assumed by most of the small-angle neutron scattering (SANS) user community that a neutrons energy is unchanged during SANS measurements. Here, the scattering from water, specifically light water, was measured on the EQ-SANS instrument, a time-of-flight SANS instrument located at the Spallation Neutron Source of Oak Ridge National Laboratory. A significant inelastic process was observed in the TOF spectra of neutrons scattered from water. Analysis of the TOF spectra from the sample showed that the scattered neutrons have energies consistent with room-temperature thermal energies (~20 meV) regardless of the incident neutron energy. With the aid of Monte Carlo particle transport simulations, we conclude that the thermalization process within the sample results in faster neutrons that arrive at the detector earlier than expected based on the incident neutron energies. This thermalization process impacts the measured SANS intensities in a manner that will ultimately be sample- and temperature-depe...
Energy Technology Data Exchange (ETDEWEB)
Shin Yunchang, E-mail: yunchang.shin@yale.ed [Department of Physics, Indiana University Bloomington, IN 47408 (United States); Department of Physics, Yale University, New Haven, CT 06511 (United States); Lavelle, C.M.; Mike Snow, W.; Baxter, David V.; Tong Xin; Yan Haiyang [Department of Physics, Indiana University Bloomington, IN 47408 (United States); Leuschner, Mark [ProCure 420 North Walnut Street Bloomington, IN 47404 (United States)
2010-08-21
Measurements of the neutron brightness from a solid methane moderator were performed at the Low Energy Neutron Source (LENS) at the Indiana University Cyclotron Facility (IUCF) to characterize the source and to test our new neutron scattering model of phase II solid methane . A time-of-flight method was used to measure the neutron energy spectrum from the moderator in the energy range of 0.1 meV {approx}1eV. Neutrons were counted with a high efficiency {sup 3}He detector. The solid methane in the moderator occupied phase II and the energy spectra were measured at 20 K and 4 K. We tested our newly developed scattering kernels for phase II solid methane by calculating the neutron brightness expected from the methane moderator at the LENS neutron source using MCNP (Monte Carlo N-particle Transport Code). Within the accuracy of our approximate approach, our model correctly predicts the neutron brightness at both temperatures.
Energy Technology Data Exchange (ETDEWEB)
Brueckel, Thomas; Heger, Gernot; Richter, Dieter; Roth, Georg; Zorn, Reiner (eds.)
2010-07-01
The following topics are dealt with: Neutron sources, neutron properties and elastic scattering, correlation functions measured by scattering experiments, symmetry of crystals, applications of neutron scattering, polarized-neutron scattering and polarization analysis, structural analysis, magnetic and lattice excitation studied by inelastic neutron scattering, macromolecules and self-assembly, dynamics of macromolecules, correlated electrons in complex transition-metal oxides, surfaces, interfaces, and thin films investigated by neutron reflectometry, nanomagnetism. (HSI)
Federal Laboratory Consortium — The Neutron Therapy Facility provides a moderate intensity, broad energy spectrum neutron beam that can be used for short term irradiations for radiobiology (cells)...
Development of compact accelerator neutron source
Directory of Open Access Journals (Sweden)
Letourneau Alain
2017-01-01
Full Text Available There is a worldwide growing interest for small-scale and reduced-cost neutron sources not based on nuclear fission. High-intensity proton or deuteron beams impinging on light materials could be used to produce such neutron sources with intensities or brightness comparable to nuclear reactor for dedicated experiments. To develop such technologies several key technological issues have to be addressed. Among them the neutron production and the maximization of the neutron extraction and transport to the instrument is a key parameter for the design of high-brightness sources adapted for the required application. This issue have to be addressed with validated and predictive Monte-Carlo simulations. In this paper we present preliminary results on the use of Geant4 in the context of Compact Accelerator based Neutron Source (CANS developments.
Advanced Neutron Source (ANS) Project progress report
Energy Technology Data Exchange (ETDEWEB)
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))
1990-04-01
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.
Utilization of low voltage D-T neutron generators in neutron physics studies
Energy Technology Data Exchange (ETDEWEB)
Singkarat, S.
1995-08-01
In a small nuclear laboratory of a developing country a low voltage D-T neutron generator can be a very useful scientific apparatus. Such machines have been used successfully for more than 40 years in teaching and scientific research. The original continuous mode 150-kV D-T neutron generator has been modified to have also a capability of producing 2-ns pulsed neutrons. Together with a carefully designed 10 m long flight path collimator and shielding of a 25 cm diameter {center_dot} 10 cm thick BC-501 neutron detector, the pulsing system was successfully used for measuring the double differential cross-section (DDX) of natural iron for 14.1-MeV neutron from the angle of 30 deg to 150 deg in 10 deg steps. In order to extend the utility of the generator, two methods for converting the almost monoenergetic 14-MeV neutrons to monoenergetic neutrons of lower energy were proposed and tested. The first method uses a pulsed neutron generator and the second method uses an ordinary continuous mode generator. The latter method was successfully used to measure the scintillation light output of a 1.4 cm diameter spherical NE-213 scintillation detector. The neutron generator has also been used in the continuous search for improved neutron detection techniques. There is a proposal, based on Monte Carlo calculations, of using a scintillation fiber for a fast neutron spectrometer. Due to the slender shape of the fiber, the pattern of produced light gives a peak in the pulse height spectrum instead of the well-known rectangular-like distribution, when the fiber is bombarded end-on by a beam of 14-MeV neutrons. Experimental investigations were undertaken. Detailed investigations on the light transportation property of a short fiber were performed. The predicted peak has not yet been found but the fiber detector may be developed as a directional discrimination fast neutron detector. 18 refs.
Energy Technology Data Exchange (ETDEWEB)
Candelore, N R; Gast, R C; Ondis, II, L A
1978-08-01
The RCP01 Monte Carlo program for the CDC-7600 and CDC-6600 performs fixed source or eigenfunction neutron reaction rate calculations, or photon reaction rate calculations, for complex geometries. The photon calculations may be linked to the neutron reaction rate calculations. For neutron calculations, the full energy range is treated as required for neutron birth by the fission process and the subsequent neutron slowing down and thermalization, i.e., 10 MeV to 0 eV; for photon calculations the same energy range is treated. The detailed cross sections required for the neutron or photon collision processes are provided by RCPL1. This report provides details of the various types of neutron and photon starts and collisions, the common geometry tracking, and the input required. 37 figures, 1 table.
Wilson, Stephen Christian
Small, highly enriched reactors designed for weapons effects simulations undergo extreme thermal transients during pulsed operations. The primary shutdown mechanism of these reactors---thermal expansion of fuel material---experiences an inertial delay resulting in a different value for the fuel temperature coefficient of reactivity during pulse operation as compared to the value appropriate for steady-state operation. The value appropriate for pulsed operation may further vary as a function of initial reactivity addition. Here we design and implement a finite element numerical method to predict the pulse operation behavior of Sandia Pulsed Reactor (SPR) II, SPR III, and a hypothetical spherical assembly with identical fuel properties without using operationally observed data in our model. These numerical results are compared to available SPR II and SPR III operational data. The numerical methods employed herein may be modified and expanded in functionality to provide both accurate characterization of the behavior of fast burst reactors of any common geometry or isotropic fuel material in the design phase, as well as a computational tool for general coupled thermomechanical-neutronics behavior in the solid state for any reactor type.
Energy Technology Data Exchange (ETDEWEB)
Brueckel, Thomas; Heger, Gernot; Richter, Dieter; Roth, Georg; Zorn, Reiner (eds.)
2010-07-01
The following topics are dealt with: Neutron sources, symmetry of crystals, diffraction, nanostructures investigated by small-angle neutron scattering, the structure of macromolecules, spin dependent and magnetic scattering, structural analysis, neutron reflectometry, magnetic nanostructures, inelastic scattering, strongly correlated electrons, dynamics of macromolecules, applications of neutron scattering. (HSI)
Neutron Capture Nucleosynthesis
Kiss, Miklos
2016-01-01
Heavy elements (beyond iron) are formed in neutron capture nucleosynthesis processes. We have proposed a simple unified model to investigate the neutron capture nucleosynthesis in arbitrary neutron density environment. We have also investigated what neutron density is required to reproduce the measured abundance of nuclei assuming equilibrium processes. We found both of these that the medium neutron density has a particularly important role at neutron capture nucleosynthesis. About these results most of the nuclei can formed at medium neutron capture density environment e.g. in some kind of AGB stars. Besides these observations our model is capable to use educational purpose.
Nuclear reactor neutron shielding
Energy Technology Data Exchange (ETDEWEB)
Speaker, Daniel P; Neeley, Gary W; Inman, James B
2017-09-12
A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.
Design Analyses and Shielding of HFIR Cold Neutron Scattering Instruments
Energy Technology Data Exchange (ETDEWEB)
Gallmeier, F.X.; Selby, D.L.; Winn, B.; Stoica, D.; Jones, A.B.; Crow, L. [Neutron Sciences Directorate, Oak Ridge National Laboratory (United States)
2011-07-01
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)
Neutron counter based on beryllium activation
Energy Technology Data Exchange (ETDEWEB)
Bienkowska, B.; Prokopowicz, R.; Kaczmarczyk, J.; Paducha, M. [Institute of Plasma Physics and Laser Microfusion (IPPLM), Hery 23, 01-497 Warsaw (Poland); Scholz, M.; Igielski, A. [Institute of Nuclear Physics PAS (IFJPAN), Radzikowskiego 152, 31-342 Krakow (Poland); Karpinski, L. [Faculty of Electrical Engineering, Rzeszow University of Technology, Pola 2, 35-959 Rzeszow (Poland); Pytel, K. [National Centre for Nuclear Research (NCBJ), Soltana 7, 05-400 Otwock - Swierk (Poland)
2014-08-21
The fusion reaction occurring in DD plasma is followed by emission of 2.45 MeV neutrons, which carry out information about fusion reaction rate and plasma parameters and properties as well. Neutron activation of beryllium has been chosen for detection of DD fusion neutrons. The cross-section for reaction {sup 9}Be(n, α){sup 6}He has a useful threshold near 1 MeV, which means that undesirable multiple-scattered neutrons do not undergo that reaction and therefore are not recorded. The product of the reaction, {sup 6}He, decays with half-life T{sub 1/2} = 0.807 s emitting β{sup −} particles which are easy to detect. Large area gas sealed proportional detector has been chosen as a counter of β–particles leaving activated beryllium plate. The plate with optimized dimensions adjoins the proportional counter entrance window. Such set-up is also equipped with appropriate electronic components and forms beryllium neutron activation counter. The neutron flux density on beryllium plate can be determined from the number of counts. The proper calibration procedure needs to be performed, therefore, to establish such relation. The measurements with the use of known β–source have been done. In order to determine the detector response function such experiment have been modeled by means of MCNP5–the Monte Carlo transport code. It allowed proper application of the results of transport calculations of β{sup −} particles emitted from radioactive {sup 6}He and reaching proportional detector active volume. In order to test the counter system and measuring procedure a number of experiments have been performed on PF devices. The experimental conditions have been simulated by means of MCNP5. The correctness of simulation outcome have been proved by measurements with known radioactive neutron source. The results of the DD fusion neutron measurements have been compared with other neutron diagnostics.
Neutron counter based on beryllium activation
Bienkowska, B.; Prokopowicz, R.; Scholz, M.; Kaczmarczyk, J.; Igielski, A.; Karpinski, L.; Paducha, M.; Pytel, K.
2014-08-01
The fusion reaction occurring in DD plasma is followed by emission of 2.45 MeV neutrons, which carry out information about fusion reaction rate and plasma parameters and properties as well. Neutron activation of beryllium has been chosen for detection of DD fusion neutrons. The cross-section for reaction 9Be(n, α)6He has a useful threshold near 1 MeV, which means that undesirable multiple-scattered neutrons do not undergo that reaction and therefore are not recorded. The product of the reaction, 6He, decays with half-life T1/2 = 0.807 s emitting β- particles which are easy to detect. Large area gas sealed proportional detector has been chosen as a counter of β-particles leaving activated beryllium plate. The plate with optimized dimensions adjoins the proportional counter entrance window. Such set-up is also equipped with appropriate electronic components and forms beryllium neutron activation counter. The neutron flux density on beryllium plate can be determined from the number of counts. The proper calibration procedure needs to be performed, therefore, to establish such relation. The measurements with the use of known β-source have been done. In order to determine the detector response function such experiment have been modeled by means of MCNP5-the Monte Carlo transport code. It allowed proper application of the results of transport calculations of β- particles emitted from radioactive 6He and reaching proportional detector active volume. In order to test the counter system and measuring procedure a number of experiments have been performed on PF devices. The experimental conditions have been simulated by means of MCNP5. The correctness of simulation outcome have been proved by measurements with known radioactive neutron source. The results of the DD fusion neutron measurements have been compared with other neutron diagnostics.
Weapons Neutron Research Facility (WNR)
Federal Laboratory Consortium — The Weapons Neutron Research Facility (WNR) provides neutron and proton beams for basic, applied, and defense-related research. Neutron beams with energies ranging...
Energy Technology Data Exchange (ETDEWEB)
NONE
2010-07-15
This report is a kind of audit report on a research laboratory whose activity is organized in three departments: neutron transport and criticality (themes: numerical methods, maths and statistics related to the simulation of neutral particle propagation, nuclear data, uncertainty propagation and bias estimation, code qualification and associated experimental programs, neutron transport in reactors and fuel cycle, criticality accidents), radionuclide transfer in radioactive waste disposals (site identification strategy, hydro-mechanical phenomena affecting storage performance, physical-chemical evolution factors, storage modelling), and metrology and confinement of radioactive gases and aerosols. The authors discuss an assessment of the unit activities in terms of strengths and opportunities, aspects to be improved and recommendations, productions and publications. A more detailed assessment is presented for each department in terms of scientific quality, influence and attractiveness (awards, recruitment capacity, capacity to obtain financing and to tender, participation to international programs), strategy and governance, and project
Welch, J. A.; Kópházi, J.; Owens, A. R.; Eaton, M. D.
2017-10-01
In this paper a method is presented for the application of energy-dependent spatial meshes applied to the multigroup, second-order, even-parity form of the neutron transport equation using Isogeometric Analysis (IGA). The computation of the inter-group regenerative source terms is based on conservative interpolation by Galerkin projection. The use of Non-Uniform Rational B-splines (NURBS) from the original computer-aided design (CAD) model allows for efficient implementation and calculation of the spatial projection operations while avoiding the complications of matching different geometric approximations faced by traditional finite element methods (FEM). The rate-of-convergence was verified using the method of manufactured solutions (MMS) and found to preserve the theoretical rates when interpolating between spatial meshes of different refinements. The scheme's numerical efficiency was then studied using a series of two-energy group pincell test cases where a significant saving in the number of degrees-of-freedom can be found if the energy group with a complex variation in the solution is refined more than an energy group with a simpler solution function. Finally, the method was applied to a heterogeneous, seven-group reactor pincell where the spatial meshes for each energy group were adaptively selected for refinement. It was observed that by refining selected energy groups a reduction in the total number of degrees-of-freedom for the same total L2 error can be obtained.
Zeitelhack, K.; Schanzer, C.; Kastenmüller, A.; Röhrmoser, A.; Daniel, C.; Franke, J.; Gutsmiedl, E.; Kudryashov, V.; Maier, D.; Päthe, D.; Petry, W.; Schöffel, T.; Schreckenbach, K.; Urban, A.; Wildgruber, U.
2006-05-01
A sophisticated neutron guide system has been installed at the new Munich neutron source FRM-II to transport neutrons from the D 2 cold neutron source to several instruments, which are situated in a separate neutron guide hall. The guide system takes advantage of supermirror coatings and includes a worldwide unique "twisted" guide for a desired phase space transformation of the neutron beam. During the initial reactor commissioning in summer 2004, the integral and differential neutron flux as well as the distribution of beam divergence at the exit of two representative and the twisted neutron guide were measured using time-of-flight spectroscopy and gold-foil activation. The experimental results can be compared to extensive simulation calculations based on MCNP and McStas. The investigated guides fulfill the expectations of providing high neutron fluxes and reveal good quality with respect to the reflective coatings and the installation precision.
Energy Technology Data Exchange (ETDEWEB)
Zeitelhack, K. [ZWE FRM-II, TU Muenchen, D-85747 Garching (Germany)]. E-mail: karl.zeitelhack@frm2.tum.de; Schanzer, C. [Physik-Department E21, TU Muenchen, D-85747 Garching (Germany); Kastenmueller, A. [ZWE FRM-II, TU Muenchen, D-85747 Garching (Germany); Roehrmoser, A. [ZWE FRM-II, TU Muenchen, D-85747 Garching (Germany); Daniel, C. [Physik-Department E22, TU Muenchen, D-85747 Garching (Germany); Franke, J. [Max-Planck-Institut fuer Metallforschung, D-70569 Stuttgart (Germany); Gutsmiedl, E. [ZWE FRM-II, TU Muenchen, D-85747 Garching (Germany); Kudryashov, V. [GKSS Forschungszentrum GmbH, D-21502 Geesthacht (Germany); Maier, D. [ZWE FRM-II, TU Muenchen, D-85747 Garching (Germany); Paethe, D. [ZWE FRM-II, TU Muenchen, D-85747 Garching (Germany); Petry, W. [ZWE FRM-II, TU Muenchen, D-85747 Garching (Germany); Schoeffel, T. [ZWE FRM-II, TU Muenchen, D-85747 Garching (Germany); Schreckenbach, K. [ZWE FRM-II, TU Muenchen, D-85747 Garching (Germany); Urban, A. [ZWE FRM-II, TU Muenchen, D-85747 Garching (Germany); Wildgruber, U. [Max-Planck-Institut fuer Metallforschung, D-70569 Stuttgart (Germany)
2006-05-10
A sophisticated neutron guide system has been installed at the new Munich neutron source FRM-II to transport neutrons from the D{sub 2} cold neutron source to several instruments, which are situated in a separate neutron guide hall. The guide system takes advantage of supermirror coatings and includes a worldwide unique 'twisted' guide for a desired phase space transformation of the neutron beam. During the initial reactor commissioning in summer 2004, the integral and differential neutron flux as well as the distribution of beam divergence at the exit of two representative and the twisted neutron guide were measured using time-of-flight spectroscopy and gold-foil activation. The experimental results can be compared to extensive simulation calculations based on MCNP and McStas. The investigated guides fulfill the expectations of providing high neutron fluxes and reveal good quality with respect to the reflective coatings and the installation precision.
Commissioning of the upgraded ultracold neutron source at Los Alamos Neutron Science Center
Pattie, Robert; LANL-UCN Team Team
2016-09-01
The spallation-driven solid-deuterium ultracold neutron (UCN) source at Los Alamos Neutron Science Center (LANSCE) has provided a facility for precision measurements of fundamental symmetries via the decay observables from neutron beta decay for nearly a decade. In preparation for a new room temperature neutron electric dipole moment (nEDM) experiment and to increase the statistical sensitivity of all experiments using the source an effort to upgrade the existing source has been carried out during 2016. This upgrade includes installing a redesigned cold neutron moderator and with optimized UCN converter geometries, improved coupling and nickel-phosphorus coating of the UCN transport system through the biological shielding, optimization of beam timing structure, and increase of the proton beam current. We will present the result of the commissioning run of the new source.
Institute of Scientific and Technical Information of China (English)
张斌; 陈义学; 王伟金; 杨寿海; 吴军; 殷雯; 梁天骄; 贾学军
2011-01-01
The construction of China Spallation Neutron Source （CSNS） has been initiated in Dongguan, Guangdong, China. Thus a detailed radiation transport analysis of the shutter neutron beam stop is of vital importance. The analyses are performed using the coupled
Superheated drop neutron spectrometer
Das, M; Roy, B; Roy, S C; Das, Mala
2000-01-01
Superheated drops are known to detect neutrons through the nucleation caused by the recoil nuclei produced by the interactions of neutrons with the atoms constituting the superheated liquid molecule. A novel method of finding the neutron energy from the temperature dependence response of SDD has been developed. From the equivalence between the dependence of threshold energy for nucleation on temperature of SDD and the dependence of dE/dx of the recoil ions with the energy of the neutron, a new method of finding the neutron energy spectrum of a polychromatic as well as monochromatic neutron source has been developed.
Wang, Ching L.
1983-09-13
Apparatus for improved sensitivity and time resolution of a neutron measurement. The detector is provided with an electrode assembly having a neutron sensitive cathode which emits relatively low energy secondary electrons. The neutron sensitive cathode has a large surface area which provides increased sensitivity by intercepting a greater number of neutrons. The cathode is also curved to compensate for differences in transit time of the neutrons emanating from the point source. The slower speeds of the secondary electrons emitted from a certain portion of the cathode are matched to the transit times of the neutrons impinging thereupon.
Neutronic Analyses in Support of the HFIR Beamline Modifications and Lifetime Extension
Remec, I.; Blakeman, E. D.
2009-08-01
At the High Flux Isotope Reactor, in operation since 1966 at the Oak Ridge National Laboratory, a larger HB-2 beam tube was installed to enhance capabilities for neutron science research. Neutronic analyses, including dosimetry measurements, radiation transport simulations, and simultaneous neutron and gamma spectrum adjustment calculations, performed to assess the impact of modifications on the PV lifetime are presented.
Fusion Neutron Flux Monitor for ITER
Institute of Scientific and Technical Information of China (English)
YANG Jinwei; YANG Qingwei; XIAO Gongshan; ZHANG Wei; SONG Xianying; LI Xu
2008-01-01
Neutron flux monitor (NFM) as an important diagnostic sub-system in ITER (international thermonuclear experimental reactor) provides a global neutron source intensity, fusion power and neutron flux in real time. Three types of neutron flux monitor assemblies with different sensitivities and shielding materials have been designed. Through MCNP (Mante-Carlo neutral particle transport code) calculations, this extended system of NFM can detect the neutron flux in a range of 104 n/(cm2·s) to 1014 n/(cm2·s). It is capable of providing accurate neutron yield measurements for all operational modes encountered in the ITER experiments including the in-situ calibration. Combining both the counting mode and Campbelling (MSV; Mean Square Voltage) mode in the signal processing units, the requirement of the dynamic range (107) for these NFMs and time resolution (1 ms) can be met. Based on a uncertainty analysis, the estimated absolute measurement accuracies of the total fusion neutron yield can reach the required 10% level in both the early stage of the DD-phase and the full power DT operation mode. In the advanced DD-phase, the absolute measurement accuracy would be better than 20%.
Compositional terranes on Mercury: Information from fast neutrons
Lawrence, David J.; Peplowski, Patrick N.; Beck, Andrew W.; Feldman, William C.; Frank, Elizabeth A.; McCoy, Timothy J.; Nittler, Larry R.; Solomon, Sean C.
2017-01-01
We report measurements of the flux of fast neutrons at Mercury from 20ºS to the north pole. On the basis of neutron transport simulations and remotely sensed elemental compositions, cosmic-ray-induced fast neutrons are shown to provide a measure of average atomic mass, , a result consistent with earlier studies of the Moon and Vesta. The dynamic range of fast neutron flux at Mercury is 3%, which is smaller than the fast-neutron dynamic ranges of 30% and 6% at the Moon and Vesta, respectively. Fast-neutron data delineate compositional terranes on Mercury that are complementary to those identified with X-ray, gamma-ray, and slow-neutron data. Fast neutron measurements confirm the presence of a region with high , relative to the mean for the planet, that coincides with the previously identified high-Mg region and reveal the existence of at least two additional compositional terranes: a low- region within the northern smooth plains and a high- region near the equator centered near 90ºE longitude. Comparison of the fast-neutron map with elemental composition maps show that variations predicted from the combined element maps are not consistent with the measured variations in fast-neutron flux. This lack of consistency could be due to incomplete coverage for some elements or uncertainties in the interpretations of compositional and neutron data. Currently available data and analyses do not provide sufficient constraints to resolve these differences.
Review of Neutron Evaluation of 93Nb for Various Database
Institute of Scientific and Technical Information of China (English)
YU; Bao-sheng; CHEN; Guo-chang
2012-01-01
<正>Niobium is the important cladding structure material for fusion reactor, and also is one of the fission products produced by the nuclear fuel in reactors. So the 93Nb have important applications in neutron transport calculations.
Energy Technology Data Exchange (ETDEWEB)
Tuan, Hoang Sy Minh; Sun, Gwang Min [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2011-05-15
The HANARO (KAERI) research reactor has been developed a neutron guide system for cold neutron (CN) research facilities since July, 2003. The neutron guide system plays an important role in transporting cold neutrons from the CN source to the neutron facilities as CN-NDP, CN-PGAA, SANS, etc. The CN activation station is being installed in the HANARO cold-neutron research project. The CN-NDP and CN-PGAA were selected as two facilities using at this station. At the end position of CG1 and CG2B beam guides, the CN-NDP and CN-PGAA will be installed in the CN guide hall. In order to predict the neutron flux and intensity values at the CG1 and CG2B beam guides, the simulation results of neutron flux at the CG1 and CG2B beam guides are presented by using several Monte Carlo (MC) neutron ray-tracing simulation codes. The intercomparison of neutron flux values between McStas, VITESS and RESTRAX are performed for getting fairly correct results at two neutron beam guides
Energy Technology Data Exchange (ETDEWEB)
Beckert, C.
2007-12-19
Conventionally the data preparation of the neutron cross sections for reactor-core calculations pursues with 2D cell codes. Aim of this thesis was, to develop a 3D cell code, to study with this code 3D effects, and to evaluate the necessarity of a 3D data preparation of the neutron cross sections. For the calculation of the neutron transport the method of the first-collision probabilities, which are calculated with the ray-tracing method, was chosen. The mathematical algorithms were implemented in the 2D/3D cell code TransRay. For the geometry part of the program the geometry module of a Monte Carlo code was used.The ray tracing in 3D was parallelized because of the high computational time. The program TransRay was verified on 2D test problems. For a reference pressured-water reactor following 3D problems were studied: A partly immersed control rod and void (vacuum or steam) around a fuel rod as model of a steam void. All problems were for comparison calculated also with the programs HELIOS(2D) and MCNP(3D). The dependence of the multiplication factor and the averaged two-group cross section on the immersion depth of the control rod respectively of the height of the steam void were studied. The 3D-calculated two-group cross sections were compared with three conventional approximations: Linear interpolation, interpolation with flux weighting, and homogenization, At the 3D problem of the control rod it was shown that the interpolation with flux weighting is a good approximation. Therefore here a 3D data preparation is not necessary. At the test case of the single control rod, which is surrounded by the void, the three approximation for the two-group cross sections were proved as unsufficient. Therefore a 3D data preparation is necessary. The single fuel-rod cell with void can be considered as the limiting case of a reactor, in which a phase interface has been formed. [German] Standardmaessig erfolgt die Datenaufbereitung der Neutronenwirkungsquerschnitte fuer
Non-Fick ian law for the neutron density current; Atomos para el desarrollo de Mexico
Energy Technology Data Exchange (ETDEWEB)
Espinosa P, G.; Vazquez R, R. [UAM-Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, Mexico D.F. 09340 (Mexico); Morales S, J. [UNAM, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, Jiutepec, Morelos 62550 (Mexico)]. e-mail: gepe@xanum.uam.mx
2008-07-01
In this paper, a fractional wave equation for the average neutron motion in a nuclear reactor is considered. This representation covers the full spectrum of the average neutron transport behavior, i.e., Fick ian and non-Fick ian effects. The fractional diffusion model retains the main dynamic characteristics of the neutron motion. The relaxation time associated with a rapid variation in the neutron flux contains an adjustable parameter, which can be manipulated to obtain the best representation of the neutron transport phenomena. (Author)
Energy Technology Data Exchange (ETDEWEB)
Bacon, G.E. [Univ. of Sheffield (United Kingdom)
1994-12-31
The familiar extremes of crystalline material are single-crystals and random powders. In between these two extremes are polycrystalline aggregates, not randomly arranged but possessing some preferred orientation and this is the form taken by constructional materials, be they steel girders or the bones of a human or animal skeleton. The details of the preferred orientation determine the ability of the material to withstand stress in any direction. In the case of bone the crucial factor is the orientation of the c-axes of the mineral content - the crystals of the hexagonal hydroxyapatite - and this can readily be determined by neutron diffraction. In particular it can be measured over the volume of a piece of bone, utilizing distances ranging from 1mm to 10mm. The major practical problem is to avoid the intense incoherent scattering from the hydrogen in the accompanying collagen; this can best be achieved by heat-treatment and it is demonstrated that this does not affect the underlying apatite. These studies of bone give leading anatomical information on the life and activities of humans and animals - including, for example, the life history of the human femur, the locomotion of sheep, the fracture of the legs of racehorses and the life-styles of Neolithic tribes. We conclude that the material is placed economically in the bone to withstand the expected stresses of life and the environment. The experimental results are presented in terms of the magnitude of the 0002 apatite reflection. It so happens that for a random powder the 0002, 1121 reflections, which are neighboring lines in the powder pattern, are approximately equal in intensity. The latter reflection, being of manifold multiplicity, is scarcely affected by preferred orientation so that the numerical value of the 0002/1121 ratio serves quite accurately as a quantitative measure of the degree of orientation of the c-axes in any chosen direction for a sample of bone.
Directory of Open Access Journals (Sweden)
S.Z Kalantari
2015-01-01
Full Text Available One of the important neutron sources for Boron Neutron Capture Therapy (BNCT is a nuclear reactor. It needs a high flux of epithermal neutrons. The optimum conditions of the neutron spectra for BNCT are provided by the International Atomic Energy Agency (IAEA. In this paper, Miniature Neutron Source Reactor (MNSR as a neutron source for BNCT was investigated. For this purpose, we designed a Beam Shaping Assembly (BSA for the reactor and the neutron transport from the core of the reactor to the output windows of BSA was simulated by MCNPX code. To optimize the BSA performance, two sets of parameters should be evaluated, in-air and in-phantom parameters. For evaluating in-phantom parameters, a Snyder head phantom was used and biological dose rate and dose-depth curve were calculated in brain normal and tumor tissues. Our calculations showed that the neutron flux of the MNSR reactor can be used for BNCT, and the designed BSA in optimum conditions had a good therapeutic characteristic for BNCT.
Sztejnberg Gonçalves-Carralves, M L; Miller, M E
2015-12-01
Neutron generators based on inertial electrostatic confinement fusion were considered for the design of a neutron irradiation facility for explanted organ Boron Neutron Capture Therapy (BNCT) that could be installed in a health care center as well as in research areas. The chosen facility configuration is "irradiation chamber", a ~20×20×40 cm(3) cavity near or in the center of the facility geometry where samples to be irradiated can be placed. Neutron flux calculations were performed to study different manners for improving scattering processes and, consequently, optimize neutron flux in the irradiation position. Flux distributions were assessed through numerical simulations of several models implemented in MCNP5 particle transport code. Simulation results provided a wide spectrum of combinations of net fluxes and energy spectrum distributions. Among them one can find a group that can provide thermal neutron fluxes per unit of production rate in a range from 4.1·10(-4) cm(-2) to 1.6·10(-3) cm(-2) with epithermal-to-thermal ratios between 0.3% and 13% and fast-to-thermal ratios between 0.01% to 8%. Neutron generators could be built to provide more than 10(10) n s(-1) and, consequently, with an arrangement of several generators appropriate enough neutron fluxes could be obtained that would be useful for several BNCT-related irradiations and, eventually, for clinical practice.
Strategy for the absolute neutron emission measurement on ITER.
Sasao, M; Bertalot, L; Ishikawa, M; Popovichev, S
2010-10-01
Accuracy of 10% is demanded to the absolute fusion measurement on ITER. To achieve this accuracy, a functional combination of several types of neutron measurement subsystem, cross calibration among them, and in situ calibration are needed. Neutron transport calculation shows the suitable calibration source is a DT/DD neutron generator of source strength higher than 10(10) n/s (neutron/second) for DT and 10(8) n/s for DD. It will take eight weeks at the minimum with this source to calibrate flux monitors, profile monitors, and the activation system.
Neutron fluence in antiproton radiotherapy, measurements and simulations
DEFF Research Database (Denmark)
Bassler, Niels; Holzscheiter, Michael H.; Petersen, Jørgen B.B.
2010-01-01
A significant part of the secondary particle spectrum from antiproton annihilation consists of fast neutrons, which may contribute to a significant dose background found outside the primary beam. Using a polystyrene phantom as a moderator, we have performed absolute measurements of the thermalized...... part of the fast neutron spectrum using Lithium-6 and -7 Fluoride TLD pairs. The experimental results are found to be in good agreement with simulations using the Monte Carlo particle transport code FLUKA. The thermal neutron kerma resulting from the measured thermal neutron fluence is insignificant...
Workshop on neutron capture therapy
Energy Technology Data Exchange (ETDEWEB)
Fairchild, R.G.; Bond, V.P. (eds.)
1986-01-01
Potentially optimal conditions for Neutron Capture Therapy (NCT) may soon be in hand due to the anticipated development of band-pass filtered beams relatively free of fast neutron contaminations, and of broadly applicable biomolecules for boron transport such as porphyrins and monoclonal antibodies. Consequently, a number of groups in the US are now devoting their efforts to exploring NCT for clinical application. The purpose of this Workshop was to bring these groups together to exchange views on significant problems of mutual interest, and to assure a unified and effective approach to the solutions. Several areas of preclinical investigation were deemed to be necessary before it would be possible to initiate clinical studies. As neither the monomer nor the dimer of sulfhydryl boron hydride is unequivocally preferable at this time, studies on both compounds should be continued until one is proven superior.
Neutron Imaging Reveals Internal Plant Hydraulic Dynamics
Energy Technology Data Exchange (ETDEWEB)
Warren, Jeffrey [ORNL; Bilheux, Hassina Z [ORNL; Kang, Misun [ORNL; Voisin, Sophie [ORNL; Cheng, Chu-Lin [ORNL; Horita, Jusuke [ORNL; Perfect, Edmund [ORNL
2013-01-01
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.
Imaging with Scattered Neutrons
Ballhausen, H.; Abele, H.; Gaehler, R.; Trapp, M.; Van Overberghe, A.
2006-01-01
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-re...
Intense fusion neutron sources
Kuteev, B. V.; Goncharov, P. R.; Sergeev, V. Yu.; Khripunov, V. I.
2010-04-01
The review describes physical principles underlying efficient production of free neutrons, up-to-date possibilities and prospects of creating fission and fusion neutron sources with intensities of 1015-1021 neutrons/s, and schemes of production and application of neutrons in fusion-fission hybrid systems. The physical processes and parameters of high-temperature plasmas are considered at which optimal conditions for producing the largest number of fusion neutrons in systems with magnetic and inertial plasma confinement are achieved. The proposed plasma methods for neutron production are compared with other methods based on fusion reactions in nonplasma media, fission reactions, spallation, and muon catalysis. At present, intense neutron fluxes are mainly used in nanotechnology, biotechnology, material science, and military and fundamental research. In the near future (10-20 years), it will be possible to apply high-power neutron sources in fusion-fission hybrid systems for producing hydrogen, electric power, and technological heat, as well as for manufacturing synthetic nuclear fuel and closing the nuclear fuel cycle. Neutron sources with intensities approaching 1020 neutrons/s may radically change the structure of power industry and considerably influence the fundamental and applied science and innovation technologies. Along with utilizing the energy produced in fusion reactions, the achievement of such high neutron intensities may stimulate wide application of subcritical fast nuclear reactors controlled by neutron sources. Superpower neutron sources will allow one to solve many problems of neutron diagnostics, monitor nano-and biological objects, and carry out radiation testing and modification of volumetric properties of materials at the industrial level. Such sources will considerably (up to 100 times) improve the accuracy of neutron physics experiments and will provide a better understanding of the structure of matter, including that of the neutron itself.
Neutron absorbed dose in a pacemaker CMOS
Energy Technology Data Exchange (ETDEWEB)
Borja H, C. G.; Guzman G, K. A.; Valero L, C. Y.; Banuelos F, A.; Hernandez D, V. M.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Calle Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Paredes G, L., E-mail: candy_borja@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)
2011-11-15
The absorbed dose due to neutrons by a Complementary Metal Oxide Semiconductor (CMOS) has been estimated using Monte Carlo methods. Eventually a person with a pacemaker becomes a patient that must be treated by radiotherapy with a linear accelerator; the pacemaker has integrated circuits as CMOS that are sensitive to intense and pulsed radiation fields. When the Linac is working in Bremsstrahlung mode an undesirable neutron field is produced due to photoneutron reactions; these neutrons could damage the CMOS putting the patient at risk during the radiotherapy treatment. In order to estimate the neutron dose in the CMOS a Monte Carlo calculation was carried out where a full radiotherapy vault room was modeled with a W-made spherical shell in whose center was located the source term of photoneutrons produced by a Linac head operating in Bremsstrahlung mode at 18 MV. In the calculations a phantom made of tissue equivalent was modeled while a beam of photoneutrons was applied on the phantom prostatic region using a field of 10 x 10 cm{sup 2}. During simulation neutrons were isotropically transported from the Linac head to the phantom chest, here a 1 {theta} x 1 cm{sup 2} cylinder made of polystyrene was modeled as the CMOS, where the neutron spectrum and the absorbed dose were estimated. Main damages to CMOS are by protons produced during neutron collisions protective cover made of H-rich materials, here the neutron spectrum that reach the CMOS was calculated showing a small peak around 0.1 MeV and a larger peak in the thermal region, both connected through epithermal neutrons. (Author)
International Neutron Radiography Newsletter
DEFF Research Database (Denmark)
Domanus, Joseph Czeslaw
1986-01-01
At the First World Conference on Neutron Radiography i t was decided to continue the "Neutron Radiography Newsletter", published previously by J.P. Barton, as the "International Neutron Radiography Newsletter" (INRNL), with J.C. Doraanus as editor. The British Journal of Non-Destructive Testing...
Advanced neutron absorber materials
Branagan, Daniel J.; Smolik, Galen R.
2000-01-01
A neutron absorbing material and method utilizing rare earth elements such as gadolinium, europium and samarium to form metallic glasses and/or noble base nano/microcrystalline materials, the neutron absorbing material having a combination of superior neutron capture cross sections coupled with enhanced resistance to corrosion, oxidation and leaching.
Prototype Neutron Energy Spectrometer
Energy Technology Data Exchange (ETDEWEB)
Stephen Mitchell, Sanjoy Mukhopadhyay, Richard Maurer, Ronald Wolff
2010-06-16
The project goals are: (1) Use three to five pressurized helium tubes with varying polyethylene moderators to build a neutron energy spectrometer that is most sensitive to the incident neutron energy of interest. Neutron energies that are of particular interest are those from the fission neutrons (typically around 1-2 MeV); (2) Neutron Source Identification - Use the neutron energy 'selectivity' property as a tool to discriminate against other competing processes by which neutrons are generated (viz. Cosmic ray induced neutron production [ship effect], [a, n] reactions); (3) Determine the efficiency as a function of neutron energy (response function) of each of the detectors, and thereby obtain the composite neutron energy spectrum from the detector count rates; and (4) Far-field data characterization and effectively discerning shielded fission source. Summary of the presentation is: (1) A light weight simple form factor compact neutron energy spectrometer ready to be used in maritime missions has been built; (2) Under laboratory conditions, individual Single Neutron Source Identification is possible within 30 minutes. (3) Sources belonging to the same type of origin viz., (a, n), fission, cosmic cluster in the same place in the 2-D plot shown; and (4) Isotopes belonging to the same source origin like Cm-Be, Am-Be (a, n) or Pu-239, U-235 (fission) do have some overlap in the 2-D plot.
The Frankfurt neutron source FRANZ
Alzubaidi, Suha; Bartz, Ulrich; Basten, Markus; Bechtold, Alexander; Chau, Long Phi; Claessens, Christine; Dinter, Hannes; Droba, Martin; Fix, Christopher; Hähnel, Hendrik; Heilmann, Manuel; Hinrichs, Ole; Huneck, Simon; Klump, Batu; Lotz, Marcel; Mäder, Dominik; Meusel, Oliver; Noll, Daniel; Nowottnick, Tobias; Obermayer, Marcus; Payir, Onur; Petry, Nils; Podlech, Holger; Ratzinger, Ulrich; Schempp, Alwin; Schmidt, Stefan; Schneider, Philipp; Seibel, Anja; Schwarz, Malte; Schweizer, Waldemar; Volk, Klaus; Wagner, Christopher; Wiesner, Christoph
2016-05-01
A 2MeV proton beam will produce a quasi-Maxwellian neutron spectrum of around 30 keV by the 7Li(p, n)7Be reaction. The experiments are mainly focused on the measurement of differential neutron capture cross sections relevant for the astrophysical s-process in nuclear synthesis. Moreover, proton capture cross sections for the astrophysical p-process can be measured directly with the proton beam. For an efficient time of flight measurement of the neutron energies along the 0.7 m long drift from the Li-target to the sample, 1ns short, intense proton pulses are needed at the target. Additionally, to reach 107 n/cm2/s at the sample, a pulse repetition rate of 250 kHz is intended. After completion and successful running in, FRANZ will become a user facility with internal and external users. The 120 kV injector terminal and the 200mA proton source as well as the low-energy beam transport section and the FRANZ cave have been realized successfully. The 1.9 MV RF accelerator consists of a combined 4-Rod-RFQ/IH-DTL-resonator and is in the RF tuning and power testing phase. The 2 MeV transport and rebuncher section is ready for installation. In a first step FRANZ will offer experimental areas for neutron activation experiments and for proton beam experiments, as mentioned above. From the accelerator physics point of view, FRANZ will be an excellent facility for high current beam investigations and for beam wall interaction studies.
Review of current neutron detection systems for emergency response
Mukhopadhyay, Sanjoy; Maurer, Richard; Guss, Paul; Kruschwitz, Craig
2014-09-01
Neutron detectors are used in a myriad of applications—from safeguarding special nuclear materials (SNM) to determining lattice spacing in soft materials. The transformational changes taking place in neutron detection and imaging techniques in the last few years are largely being driven by the global shortage of helium-3 (3He). This article reviews the status of neutron sensors used specifically for SNM detection in radiological emergency response. These neutron detectors must be highly efficient, be rugged, have fast electronics to measure neutron multiplicity, and be capable of measuring direction of the neutron sources and possibly image them with high spatial resolution. Neutron detection is an indirect physical process: neutrons react with nuclei in materials to initiate the release of one or more charged particles that produce electric signals that can be processed by the detection system. Therefore, neutron detection requires conversion materials as active elements of the detection system; these materials may include boron-10 (10B), lithium-6 (6Li), and gadollinium-157 (157Gd), to name a few, but the number of materials available for neutron detection is limited. However, in recent years, pulse-shape-discriminating plastic scintillators, scintillators made of helium-4 (4He) under high pressure, pillar and trench semiconductor diodes, and exotic semiconductor neutron detectors made from uranium oxide and other materials have widely expanded the parameter space in neutron detection methodology. In this article we will pay special attention to semiconductor-based neutron sensors. Modern microfabricated nanotubes covered inside with neutron converter materials and with very high aspect ratios for better charge transport will be discussed.
Energy Technology Data Exchange (ETDEWEB)
Fhager, V
2000-01-01
In order to make correct predictions of the second moment of statistical nuclear variables, such as the number of fissions and the number of thermalized neutrons, the dependence of the energy distribution of the source particles on their number should be considered. It has been pointed out recently that neglecting this number dependence in accelerator driven systems might result in bad estimates of the second moment, and this paper contains qualitative and quantitative estimates of the size of these efforts. We walk towards the requested results in two steps. First, models of the number dependent energy distributions of the neutrons that are ejected in the spallation reactions are constructed, both by simple assumptions and by extracting energy distributions of spallation neutrons from a high-energy particle transport code. Then, the second moment of nuclear variables in a sub-critical reactor, into which spallation neutrons are injected, is calculated. The results from second moment calculations using number dependent energy distributions for the source neutrons are compared to those where only the average energy distribution is used. Two physical models are employed to simulate the neutron transport in the reactor. One is analytical, treating only slowing down of neutrons by elastic scattering in the core material. For this model, equations are written down and solved for the second moment of thermalized neutrons that include the distribution of energy of the spallation neutrons. The other model utilizes Monte Carlo methods for tracking the source neutrons as they travel inside the reactor material. Fast and thermal fission reactions are considered, as well as neutron capture and elastic scattering, and the second moment of the number of fissions, the number of neutrons that leaked out of the system, etc. are calculated. Both models use a cylindrical core with a homogenous mixture of core material. Our results indicate that the number dependence of the energy
Measurements of neutrons at JET by means of the activation methods
Energy Technology Data Exchange (ETDEWEB)
Prokopowicz, R., E-mail: prokopowicz@ifpilm.waw.p [EURATOM-IPPLM Association, 23, Hery St, Warsaw 01-497 (Poland); Bienkowska, B.; Drozdowicz, K.; Jednorog, S.; Kowalska-Strzeciwilk, E. [EURATOM-IPPLM Association, 23, Hery St, Warsaw 01-497 (Poland); Murari, A. [EURATOM-ENEA Fusion Association, Consorzio RFX, Padova I-35127 (Italy); Popovichev, S. [EURATOM-CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Pytel, K.; Scholz, M.; Szydlowski, A. [EURATOM-IPPLM Association, 23, Hery St, Warsaw 01-497 (Poland); Syme, B. [EURATOM-CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Tracz, G. [EURATOM-IPPLM Association, 23, Hery St, Warsaw 01-497 (Poland)
2011-05-01
The neutron diagnostics in tokamaks like Joint European Torus (JET) are essential in estimating fusion power. The neutron activation method, supported by neutron transport calculations, is particularly useful for the evaluation of the total neutron yield from a single plasma discharge. This paper presents the results of activation experiments and calculations carried out for JET plasmas, from the selection of the activation materials to their irradiations in the neutron field of JET discharges. Neutron transport calculations were performed, leading to activation coefficients for new materials. The results of the calculations were used to design new composite samples to obtain information on both the yield and the neutron spectrum. The neutron measurements using these new activation materials were performed during the last JET experimental campaigns. The results are compared with neutron transport calculations. Additionally, application of the cadmium difference method allows revelation of the part of thermal neutrons near the tokamak first wall. The advantages of new activation materials and benchmarking the activation method against neutron transport calculations are also discussed.
Neutron scattering. Experiment manuals
Energy Technology Data Exchange (ETDEWEB)
Brueckel, Thomas; Heger, Gernot; Richter, Dieter; Roth, Georg; Zorn, Reiner (eds.)
2010-07-01
The following topics are dealt with: The thermal triple axis spectrometer PUMA, the high-resolution powder diffractometer SPODI, the hot single-crystal diffractometer HEiDi for structure analysis with neutrons, the backscattering spectrometer SPHERES, neutron polarization analysis with tht time-of-flight spectrometer DNS, the neutron spin-echo spectrometer J-NSE, small-angle neutron scattering with the KWS-1 and KWS-2 diffractometers, the very-small-angle neutron scattering diffractrometer with focusing mirror KWS-3, the resonance spin-echo spectrometer RESEDA, the reflectometer TREFF, the time-of-flight spectrometer TOFTOF. (HSI)
Grazing Incidence Neutron Optics
Gubarev, Mikhail V. (Inventor); Ramsey, Brian D. (Inventor); Engelhaupt, Darell E. (Inventor)
2013-01-01
Neutron optics based on the two-reflection geometries are capable of controlling beams of long wavelength neutrons with low angular divergence. The preferred mirror fabrication technique is a replication process with electroform nickel replication process being preferable. In the preliminary demonstration test an electroform nickel optics gave the neutron current density gain at the focal spot of the mirror at least 8 for neutron wavelengths in the range from 6 to 20.ANG.. The replication techniques can be also be used to fabricate neutron beam controlling guides.
Mascarenhas, Nicholas; Marleau, Peter; Brennan, James S.; Krenz, Kevin D.
2010-06-22
An instrument that will directly image the fast fission neutrons from a special nuclear material source has been described. This instrument can improve the signal to background compared to non imaging neutron detection techniques by a factor given by ratio of the angular resolution window to 4.pi.. In addition to being a neutron imager, this instrument will also be an excellent neutron spectrometer, and will be able to differentiate between different types of neutron sources (e.g. fission, alpha-n, cosmic ray, and D-D or D-T fusion). Moreover, the instrument is able to pinpoint the source location.
Neutron shielding for a {sup 252} Cf source
Energy Technology Data Exchange (ETDEWEB)
Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M. [Unidades Academicas de Estudios Nucleares e Ingenieria Electrica, Universidad Autonoma de Zacatecas, C. Cipres 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Eduardo Gallego, Alfredo Lorente [Depto. de Ingenieria Nuclear, ETS Ingenieros Industriales, Universidad Politecnica de Madrid, C. Jose Gutierrez Abascal 2, 28006 Madrid (Spain)]. e-mail: fermineutron@yahoo.com
2006-07-01
To determine the neutron shielding features of water-extended polyester a Monte Carlo study was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through inelastic collisions and absorption reactions. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide production induced by neutron activation must be considered. In this investigation the Monte Carlo method was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a {sup 252}Cf isotopic neutron source. During calculations a detailed model for the {sup 252}Cf and the shield was utilized. To compare the shielding features of water extended polyester, the calculations were also made for the bare {sup 252}Cf in vacuum, air and the shield filled with water. For all cases the calculated neutron spectra was utilized to determine the ambient equivalent neutron dose at four sites around the shielding. In the case of water extended polyester and water shielding the calculations were extended to include the prompt gamma rays produced during neutron interactions, with this information the Kerma in air was calculated at the same locations where the ambient equivalent neutron dose was determined. (Author)
Neutron beam facilities at the Australian Replacement Research Reactor
Energy Technology Data Exchange (ETDEWEB)
Kennedy, Shane; Robinson, Robert; Hunter, Brett [Physics Division, ANSTO, Lucas Heights (Australia)
2001-03-01
Australia is building a research reactor to replace the HIFAR reactor at Lucas Heights by the end of 2005. Like HIFAR, the Replacement Research Reactor will be multipurpose with capabilities for both neutron beam research and radioisotope production. It will be a pool-type reactor with thermal neutron flux (unperturbed) of 4 x 10{sup 14} n/cm{sup 2}/sec and a liquid D{sub 2} cold neutron source. Cold and thermal neutron beams for neutron beam research will be provided at the reactor face and in a large neutron guide hall. Supermirror neutron guides will transport cold and thermal neutrons to the guide hall. The reactor and the associated infrastructure, with the exception of the neutron beam instruments, is to be built by INVAP S.E. under contract. The neutron beam instruments will be developed by ANSTO, in consultation with the Australian user community. This status report includes a review the planned scientific capabilities, a description of the facility and a summary of progress to date. (author)
Advances in neutron tomography
Indian Academy of Sciences (India)
W Treimer
2008-11-01
In the last decade neutron radiography (NR) and tomography (NCT) have experienced a number of improvements, due to the well-known properties of neutrons interacting with matter, i.e. the low attenuation by many materials, the strong attenuation by hydrogenous constituent in samples, the wavelength-dependent attenuation in the neighbourhood of Bragg edges and due to better 2D neutron detectors. So NR and NCT were improved by sophisticated techniques that are based on the attenuation of neutrons or on phase changes of the associated neutron waves if they pass through structured materials. Up to now the interaction of the neutron spin with magnetic fields in samples has not been applied to imaging techniques despite the fact that it was proposed many years ago. About ten years ago neutron depolarization as imaging signal for neutron radiography or tomography was demonstrated and in principle it works. Now one can present much improved test experiments using polarized neutrons for radiographic imaging. For this purpose the CONRAD instrument of the HMI was equipped with polarizing and analysing benders very similar to conventional scattering experiments using polarized neutrons. Magnetic fields in different coils and in samples (superconductors) at low temperatures could be visualized. In this lecture a summary about standard signals (attenuation) and the more `sophisticated' imaging signals as refraction, small angle scattering and polarized neutrons will be given.
Fermi, Enrico; Zinn, Walter H.
The argument of the present Patent is a radiation shield suitable for protection of personnel from both gamma rays and neutrons. Such a shield from dangerous radiations is achieved to the best by the combined action of a neutron slowing material (a moderator) and a neutron absorbing material. Hydrogen is particularly effective for this shield since it is a good absorber of slow neutrons and a good moderator of fast neutrons. The neutrons slowed down by hydrogen may, then, be absorbed by other materials such as boron, cadmium, gadolinium, samarium or steel. Steel is particularly convenient for the purpose, given its effectiveness in absorbing also the gamma rays from the reactor (both primary gamma rays and secondary ones produced by the moderation of neutrons). In particular, in the present Patent a shield is described, made of alternate layers of steel and Masonite (an hydrolized ligno-cellulose material). The object of the present Patent is not discussed in any other published paper.
Xu, Yanping; Randers-Pehrson, Gerhard; Marino, Stephen A; Garty, Guy; Harken, Andrew; Brenner, David J
2015-09-11
A novel neutron irradiation facility at the Radiological Research Accelerator Facility (RARAF) has been developed to mimic the neutron radiation from an Improvised Nuclear Device (IND) at relevant distances (e.g. 1.5 km) from the epicenter. The neutron spectrum of this IND-like neutron irradiator was designed according to estimations of the Hiroshima neutron spectrum at 1.5 km. It is significantly different from a standard reactor fission spectrum, because the spectrum changes as the neutrons are transported through air, and it is dominated by neutron energies from 100 keV up to 9 MeV. To verify such wide energy range neutron spectrum, detailed here is the development of a combined spectroscopy system. Both a liquid scintillator detector and a gas proportional counter were used for the recoil spectra measurements, with the individual response functions estimated from a series of Monte Carlo simulations. These normalized individual response functions were formed into a single response matrix for the unfolding process. Several accelerator-based quasi-monoenergetic neutron source spectra were measured and unfolded to test this spectroscopy system. These reference neutrons were produced from two reactions: T(p,n)(3)He and D(d,n)(3)He, generating neutron energies in the range between 0.2 and 8 MeV. The unfolded quasi-monoenergetic neutron spectra indicated that the detection system can provide good neutron spectroscopy results in this energy range. A broad-energy neutron spectrum from the (9)Be(d,n) reaction using a 5 MeV deuteron beam, measured at 60 degrees to the incident beam was measured and unfolded with the evaluated response matrix. The unfolded broad neutron spectrum is comparable with published time-of-flight results. Finally, the pair of detectors were used to measure the neutron spectrum generated at the RARAF IND-like neutron facility and a comparison is made to the neutron spectrum of Hiroshima.
Energy Technology Data Exchange (ETDEWEB)
Shibata, Yasunari; Iguchi, Tetsuo; Ogata, Tomohiro; Umemura, Norihiro; Asai, Keisuke; Kawarabayashi, Jun [Nagoya Univ., Nagoya, Aichi (Japan); Sasao, Mamiko [National Inst. for Fusion Science, Toki, Gifu (Japan)
2003-07-01
A compact Cockcroft-Walton type accelerator DD neutron source has been developed for in situ calibration experiments on neutron measurements at LHD. The equipment mainly consists of three parts; a deuterium (D) reservoir/ion source, a self-loaded deuterium target and a 100 kV high voltage power supply, all of which are contained in a compact cylindrical stainless steel (SUS) tube of 70 mm in diameter and 780 mm in length. About one hour steady operation was performed under the acceleration voltage of 80 keV and the ion beam current of {approx}60 {mu}A, corresponding to the DD neutron yield of around 10{sup 5} n/s. The neutron emission profile and energy spectrum were measured with an NE213 scintillator and a {sup 3}He gas proportional counter. Preliminary neutronic calculations with a Monte Carlo neutron transport code MCNP' were also executed for simulating the in situ calibration experiment for neutron detectors that will be installed on LHD. Through the experiments and the calculations, it is shown that the present DD neutron source is valid for in situ calibration on threshold type detectors used for neutron emission profile monitoring and neutron spectrometry at DD plasma experiments. (author)
The neutron instrument simulation package, NISP
Seeger, Philip A.; Daemen, Luke L.
2004-10-01
The Neutron Instrument Simulation Package (NISP) performs complete source-to-detector simulations of neutron instruments, including neutrons that do not follow the expected path. The original user interface (MC_Web) is a web-based application, http://strider.lansce.lanl.gov/NISP/Welcome.html. This report describes in detail the newer stand-alone Windows version, NISP_Win. Instruments are assembled from menu-selected elements, including neutron sources, collimation and transport elements, samples, analyzers, and detectors. Magnetic field regions may also be specified for the propagation of polarized neutrons including spin precession. Either interface writes a geometry file that is used as input to the Monte Carlo engine (MC_Run) in the user's computer. Both the interface and the engine rely on a subroutine library, MCLIB. The package is completely open source. New features include capillary optics, temperature dependence of Al and Be, revised source files for ISIS, and visualization of neutron trajectories at run time. Also, a single-crystal sample type has been successfully imported from McStas (with more generalized geometry), demonstrating the capability of including algorithms from other sources, and NISP_Win may render the instrument in a virtual reality file. Results are shown for two instruments under development.
The neutron instrument simulation package, NISP.
Energy Technology Data Exchange (ETDEWEB)
Seeger, P. A. (Philip A.); Daemen, L. L. (Luc L.)
2004-01-01
The Neutron Instrument Simulation Package (NISP) performs complete source-to-detector simulations of neutron instruments, including neutrons that do not follow the expected path. The original user interface (MC{_}Web) is a web-based application, http://strider.lansce.lanl.gov/NISP/Welcome.html. This report describes in detail the newer standalone Windows version, NISP{_}Win. Instruments are assembled from menu-selected elements, including neutron sources, collimation and transport elements, samples, analyzers, and detectors. Magnetic field regions may also be specified for the propagation of polarized neutrons including spin precession. Either interface writes a geometry file that is used as input to the Monte Carlo engine (MC{_}Run) in the user's computer. Both the interface and the engine rely on a subroutine library, MCLIB. The package is completely open source. New features include capillary optics, temperature dependence of Al and Be, revised source files for ISIS, and visualization of neutron trajectories at run time. Also, a single-crystal sample type has been successfully imported from McStas (with more generalized geometry), demonstrating the capability of including algorithms from other sources, and NISP{_}Win may render the instrument in a virtual reality file. Results are shown for two instruments under development.
Neutronic performances of the MEGAPIE target
Energy Technology Data Exchange (ETDEWEB)
Panebianco, S.; Bringer, O.; Chabod, S.; Dupont, E.; Letourneau, A. [CEA Saclay, Dept. d' Astrophysique de Physique des Particules, de Physique Nucleaire et de l' Instrumentation Associee (DSM/DAPNIA/SPhN), 91- Gif sur Yvette (France); Beauvais, P.; Lotrus, P.; Molinie, F.; Toussaint, J.Ch. [CEA Saclay, Dept. d' Astrophysique de Physique des Particules, de Physique Nucleaire et de l' Instrumentation Associee (DSM/DAPNIA), 91- Gif sur Yvette (France); Chartier, F. [CEA Saclay, Dept. de Physico-Chimie (DEN/DPC/SECR), 91 - Gif sur Yvette (France); Oriol, L. [CEA Cadarache, Dept. d' Etudes des Reacteurs (DEN/DER/SPEX), 13 - Saint Paul lez Durance (France)
2008-07-01
The MEGAPIE project is a key experiment on the road to Accelerator Driven Systems and it provides the scientific community with unique data on the behavior of a liquid lead-bismuth spallation target under realistic and long term irradiation conditions. The neutronic of such target is of course of prime importance when considering its final destination as an intense neutron source. This is the motivation to characterize the inside neutron flux of the target in operation. A complex detector, made of 8 'micro' fission-chambers, has been built and installed in the core of the target, few tens of centimeters from the proton/Pb-Bi interaction zone. This detector is designed to measure the absolute neutron flux inside the target, to give its spatial distribution and to correlate its temporal variations with the beam intensity. Moreover, integral information on the neutron energy distribution as a function of the position along the beam axis could be extracted, giving integral constraints on the neutron production models implemented in transport codes such as MCNPX. (authors)
Allen, D A; Beynon, T D; Green, S
1999-01-01
This paper is concerned with the proposed Birmingham accelerator-based epithermal neutron beam for boron neutron capture therapy (BNCT). In particular, the option of producing a therapy beam at an orthogonal direction to the incoming protons is considered. Monte Carlo radiation transport simulations, both with and without a head phantom, have shown that an orthogonal beam geometry is not only acceptable but is indeed beneficial, in terms of a lower mean neutron energy and an enhanced therapeutic ratio for the same useful neutron fluence in the therapy beam. Typical treatment times for various beam options have been calculated, and range from 20 to 48 min with a 5 mA beam of 2.8 MeV protons, if the maximum photon-equivalent dose delivered to healthy tissue is to be 12.6 Gy Eq. The effects of proton beam diameter upon the therapy beam parameters have also been considered.
Design of the Mechanical Parts for the Neutron Guide System at HANARO
Shin, J. W.; Cho, Y. G.; Cho, S. J.; Ryu, J. S.
2008-03-01
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.
Upgrades to the ultracold neutron source at the Los Alamos Neutron Science Center
Pattie, Robert; LANL-nEDM Collaboration
2015-10-01
The spallation-driven solid deutrium-based ultracold neutron (UCN) source at the Los Alamos Neutron Science Center (LANSCE) has provided a facility for precision measurements of fundamental symmetries via the decay observables from neutron beta decay for nearly a decade. In preparation for a new room temperature neutron electric dipole moment (nEDM) experiment and to increase the statistical sensitivity of all experiments using the source an effort to increase the UCN output is underway. The ultimate goal is to provide a density of 100 UCN/cc or greater in the nEDM storage cell. This upgrade includes redesign of the cold neutron moderator and UCN converter geometries, improved coupling and coating of the UCN transport system through the biological shielding, optimization of beam timing structure, and increase of the proton beam current. We will present the results of the MCNP and UCN transport simulations that led to the new design, which will be installed spring 2016, and UCN guide tests performed at LANSCE and the Institut Laue-Langevin to study the UCN transport properties of a new nickel-based guide coating.
Status of spallation neutron source
Energy Technology Data Exchange (ETDEWEB)
Oyama, Yukio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1997-03-01
Existing and planned facilities using proton accelerator driven spallation neutron source are reviewed. These include new project of neutron science proposed from Japan Atomic Energy Research Institute. The present status of facility requirement and accelerator technology leads us to new era of neutron science such as neutron scattering research and nuclear transmutation study using very intense neutron source. (author)
Energy Technology Data Exchange (ETDEWEB)
Jammes, Ch
1997-11-28
The aim of this work is to create, validate theoretically and experimentally a calculation route for a thermal irradiation reactor. This is the research reactor of the University of Strasbourg, which presents all of characteristics of this reactor-type: compact and heterogeneous core, slab-type fuel with a high 235-uranium enrichment. This calculation route is based on the first use of the following two modern transport methods: the TDT method and the Monte Carlo method. The former, programmed within the APOLLO2 code, is a two dimensional collision probabilities method. The later, used by the TRIPOLI4 code, is a stochastic method. Both can be applied to complex geometries. After a few theoretical reminders about transport codes, a set of integral experiments is described which have been realized within the research reactor of the University of Strasbourg. One of them has been performed for this study. At the beginning of the theoretical part, significant errors are apparent due to the use of calculation route based on homogenization, condensation and the diffusion approximation. An extensive comparison between the discrete ordinates method and the TDT method carries out that the use of the TDT method is relevant for the studied reactor. The treatment of axial leakage with this method is the only disadvantage. Therefore, the use of the code TRIPOLI4 is recommended for a more accurate study of leakage within a reflector. By means of the experimental data, the ability of our calculation route is confirmed for essential neutronics questions such as the critical mass determination, the power distribution and the fuel management. (author)
Neutron sources and applications
Energy Technology Data Exchange (ETDEWEB)
Price, D.L. [ed.] [Argonne National Lab., IL (United States); Rush, J.J. [ed.] [National Inst. of Standards and Technology, Gaithersburg, MD (United States)
1994-01-01
Review of Neutron Sources and Applications was held at Oak Brook, Illinois, during September 8--10, 1992. This review involved some 70 national and international experts in different areas of neutron research, sources, and applications. Separate working groups were asked to (1) review the current status of advanced research reactors and spallation sources; and (2) provide an update on scientific, technological, and medical applications, including neutron scattering research in a number of disciplines, isotope production, materials irradiation, and other important uses of neutron sources such as materials analysis and fundamental neutron physics. This report summarizes the findings and conclusions of the different working groups involved in the review, and contains some of the best current expertise on neutron sources and applications.
Prototype Stilbene Neutron Collar
Energy Technology Data Exchange (ETDEWEB)
Prasad, M. K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Shumaker, D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Snyderman, N. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Verbeke, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wong, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)
2016-10-26
A neutron collar using stilbene organic scintillator cells for fast neutron counting is described for the assay of fresh low enriched uranium (LEU) fuel assemblies. The prototype stilbene collar has a form factor similar to standard He-3 based collars and uses an AmLi interrogation neutron source. This report describes the simulation of list mode neutron correlation data on various fuel assemblies including some with neutron absorbers (burnable Gd poisons). Calibration curves (doubles vs ^{235}U linear mass density) are presented for both thermal and fast (with Cd lining) modes of operation. It is shown that the stilbene collar meets or exceeds the current capabilities of He-3 based neutron collars. A self-consistent assay methodology, uniquely suited to the stilbene collar, using triples is described which complements traditional assay based on doubles calibration curves.
Energy Technology Data Exchange (ETDEWEB)
Terry, James Russell [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2016-05-05
Emission of neutrons in a given event is modeled by the DioramaEmitterNeutron object, a subclass of the abstract DioramaEmitterModule object. The GenerateEmission method of this object is the entry point for generation of a neutron population for a given event. Shown in table 1, this method requires a number of parameters to be defined in the event definition.
Development of cold neutron depth profiling system at HANARO
Energy Technology Data Exchange (ETDEWEB)
Park, B.G. [Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong-gu, Daejeon 305-355 (Korea, Republic of); Sun, G.M., E-mail: gmsun@kaeri.re.kr [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong-gu, Daejeon 305-355 (Korea, Republic of); Choi, H.D. [Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of)
2014-07-01
A neutron depth profiling (NDP) system has been designed and developed at HANARO, a 30 MW research reactor at the Korea Atomic Energy Research Institute (KAERI). The KAERI-NDP system utilizes cold neutrons that are transported along the CG1 neutron guide from the cold neutron source and it consists of a neutron beam collimator, a target chamber, a beam stopper, and charged particle detectors along with NIM-standard modules for charged particle pulse-height analysis. A 60 cm in diameter stainless steel target chamber was designed to control the positions of the sample and detector. The energy distribution of the cold neutron beam at the end of the neutron guide was calculated by using the Monte Carlo simulation code McStas, and a neutron flux of 1.8×10{sup 8} n/cm{sup 2} s was determined by using the gold foil activation method at the sample position. The performance of the charged particle detection of the KAERI-NDP system was tested by using Standard Reference Materials. The energy loss spectra of alpha particles and Li ions emitted from {sup 10}B, which was irradiated by cold neutrons, were measured. The measured peak concentration and the areal density of {sup 10}B in the Standard Reference Material are consistent with the reference values within 1% and 3.4%, respectively.
Development of cold neutron depth profiling system at HANARO
Park, B. G.; Sun, G. M.; Choi, H. D.
2014-07-01
A neutron depth profiling (NDP) system has been designed and developed at HANARO, a 30 MW research reactor at the Korea Atomic Energy Research Institute (KAERI). The KAERI-NDP system utilizes cold neutrons that are transported along the CG1 neutron guide from the cold neutron source and it consists of a neutron beam collimator, a target chamber, a beam stopper, and charged particle detectors along with NIM-standard modules for charged particle pulse-height analysis. A 60 cm in diameter stainless steel target chamber was designed to control the positions of the sample and detector. The energy distribution of the cold neutron beam at the end of the neutron guide was calculated by using the Monte Carlo simulation code McStas, and a neutron flux of 1.8×108 n/cm2 s was determined by using the gold foil activation method at the sample position. The performance of the charged particle detection of the KAERI-NDP system was tested by using Standard Reference Materials. The energy loss spectra of alpha particles and Li ions emitted from 10B, which was irradiated by cold neutrons, were measured. The measured peak concentration and the areal density of 10B in the Standard Reference Material are consistent with the reference values within 1% and 3.4%, respectively.
Long distance propagation of a polarized neutron beam in zero magnetic field
Energy Technology Data Exchange (ETDEWEB)
Schmidt, U.; Bitter, T.; El-Muzeini, P. (Heidelberg Univ. (Germany). Physikalisches Inst.); Dubbers, D. (Technische Univ. Muenchen, Garching (Germany). Fakultaet fuer Physik E21); Schaerpf, O. (Inst. Laue Langevin, 38 - Grenoble (France))
1992-09-01
A beam of fully polarized cold neutrons was transported through a zero magnetic field region of 70 m length without loss of polarization. The purpose of this exercise was twofold: Firstly, to demonstrate that the new zero-field neutron spin-echo method will work also for very long neutron flight paths; secondly, to prove in the most direct way that the neutron free-flight region of the ILL neutron-antineutron oscillation experiment was indeed sufficiently field-free ('quasifree condition') by using the neutrons themselves as a magnetometer. To this purpose the residual magnetic field integrals in the long 'zero-field' region were measured with a conventional neutron spin-echo method. The overall spin precession angle of the neutrons during their flight through the long zero-field region was found to be less than 2[sup 0]. (orig.).
Neutron scattering. Experiment manuals
Energy Technology Data Exchange (ETDEWEB)
Brueckel, Thomas; Heger, Gernot; Richter, Dieter; Roth, Georg; Zorn, Reiner (eds.)
2014-07-01
The following topics are dealt with: The thermal triple-axis spectrometer PUMA, the high-resolution powder diffractometer SPODI, the hot-single-crystal diffractometer HEiDi, the three-axis spectrometer PANDA, the backscattering spectrometer SPHERES, the DNS neutron-polarization analysis, the neutron spin-echo spectrometer J-NSE, small-angle neutron scattering at KWS-1 and KWS-2, a very-small-angle neutron scattering diffractometer with focusing mirror, the reflectometer TREFF, the time-of-flight spectrometer TOFTOF. (HSI)
Neutron-emission measurements at a white neutron source
Energy Technology Data Exchange (ETDEWEB)
Haight, Robert C [Los Alamos National Laboratory
2010-01-01
Data on the spectrum of neutrons emittcd from neutron-induced reactions are important in basic nuclear physics and in applications. Our program studies neutron emission from inelastic scattering as well as fission neutron spectra. A ''white'' neutron source (continuous in energy) allows measurements over a wide range of neutron energies all in one experiment. We use the tast neutron source at the Los Alamos Neutron Science Center for incident neutron energies from 0.5 MeV to 200 MeV These experiments are based on double time-of-flight techniques to determine the energies of the incident and emitted neutrons. For the fission neutron measurements, parallel-plate ionization or avalanche detectors identify fission in actinide samples and give the required fast timing pulse. For inelastic scattering, gamma-ray detectors provide the timing and energy spectroscopy. A large neutron-detector array detects the emitted neutrons. Time-of-flight techniques are used to measure the energies of both the incident and emitted neutrons. Design considerations for the array include neutron-gamma discrimination, neutron energy resolution, angular coverage, segmentation, detector efficiency calibration and data acquisition. We have made preliminary measurements of the fission neutron spectra from {sup 235}U, {sup 238}U, {sup 237}Np and {sup 239}Pu. Neutron emission spectra from inelastic scattering on iron and nickel have also been investigated. The results obtained will be compared with evaluated data.
Monte Carlo simulation of neutron scattering instruments
Energy Technology Data Exchange (ETDEWEB)
Seeger, P.A.; Daemen, L.L.; Hjelm, R.P. Jr.
1998-12-01
A code package consisting of the Monte Carlo Library MCLIB, the executing code MC{_}RUN, the web application MC{_}Web, and various ancillary codes is proposed as an open standard for simulation of neutron scattering instruments. The architecture of the package includes structures to define surfaces, regions, and optical elements contained in regions. A particle is defined by its vector position and velocity, its time of flight, its mass and charge, and a polarization vector. The MC{_}RUN code handles neutron transport and bookkeeping, while the action on the neutron within any region is computed using algorithms that may be deterministic, probabilistic, or a combination. Complete versatility is possible because the existing library may be supplemented by any procedures a user is able to code. Some examples are shown.
Ion Acceleration in Solar Flares Determined by Solar Neutron Observations
Watanabe, K.; Solar Neutron Observation Group
2013-05-01
Large amounts of particles can be accelerated to relativistic energy in association with solar flares and/or accompanying phenomena (e.g., CME-driven shocks), and they sometimes reach very near the Earth and penetrate the Earth's atmosphere. These particles are observed by ground-based detectors (e.g., neutron monitors) as Ground Level Enhancements (GLEs). Some of the GLEs originate from high energy solar neutrons which are produced in association with solar flares. These neutrons are also observed by ground-based neutron monitors and solar neutron telescopes. Recently, some of the solar neutron detectors have also been operating in space. By observing these solar neutrons, we can obtain information about ion acceleration in solar flares. Such neutrons were observed in association with some X-class flares in solar cycle 23, and sometimes they were observed by two different types of detectors. For example, on 2005 September 7, large solar neutron signals were observed by the neutron monitor at Mt. Chacaltaya in Bolivia and Mexico City, and by the solar neutron telescopes at Chacaltaya and Mt. Sierra Negra in Mexico in association with an X17.0 flare. The neutron signal continued for more than 20 minutes with high statistical significance. Intense gamma-ray emission was also registered by INTEGRAL, and by RHESSI during the decay phase. We analyzed these data using the solar-flare magnetic-loop transport and interaction model of Hua et al. (2002), and found that the model could successfully fit the data with intermediate values of loop magnetic convergence and pitch angle scattering parameters. These results indicate that solar neutrons were produced at the same time as the gamma-ray line emission and that ions were continuously accelerated at the emission site. In this paper, we introduce some of the solar neutron observations in solar cycle 23, and discuss the tendencies of the physical parameters of solar neutron GLEs, and the energy spectrum and population of the
Zimmer, Oliver
2016-03-01
A new neutron-cooling mechanism is proposed with potential benefits for novel intense sources of very cold neutrons with wavelengths >2 nm, and for enhancing the production of ultracold neutrons. It employs inelastic magnetic scattering in weakly absorbing, cold paramagnetic systems. Kinetic energy is removed from the neutron stepwise in constant decrements determined by the Zeeman energy of paramagnetic atoms or ions in an external magnetic field, or by zero-field level splittings in magnetic molecules. The stationary neutron transport equation is analyzed for an infinite, homogeneous medium with Maxwellian neutron sources, using inelastic scattering cross sections derived in an appendix. Nonmagnetic inelastic scattering processes are neglected. The solution therefore still underestimates very cold neutron densities that should be achievable in a real medium. Molecular oxygen with its triplet ground state appears particularly promising, notably as a host in fully deuterated O2-clathrate hydrate. Other possibilities are dry O2-4He van der Waals clusters and O2 intercalated in fcc-C60. For conversion of cold to ultracold neutrons, where an incident neutron imparts only a single energy quantum to the medium, the paramagnetic scattering in the clathrate system is found to be stronger, by more than an order of magnitude, than the single-phonon emission in superfluid helium, when evaluated for an incident neutron spectrum with the optimum temperature for the respective medium. Moreover, the multistep paramagnetic cooling cascade leads to further strong enhancements of very cold neutron densities, e.g., by a factor 14 (57) for an initial neutron temperature of 30 K (100 K ), for the moderator held at about 1.3 K . Due to a favorable Bragg cutoff of the O2 clathrate, the cascade-cooling can take effect in a moderator with linear extensions smaller than a meter.
Energy Technology Data Exchange (ETDEWEB)
Vazquez R, R.; Vazquez R, A.; Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)], e-mail: rvr@xanum.uam.mx
2009-10-15
Originally developed for heterogeneous means, the linear extended neutron diffusion theory is applied to the limit case of monoenergetic neutron diffusion in a semi-infinite homogeneous mean with a neutron source, located in the coordinate origin situated in the frontier of dispersive material. The monoenergetic neutron diffusion is studied taking into account the spatial deviations in the neutron flux to the interfacial current caused by the neutron source, as well as the influence of the spatial deviations in the absorption rate. The developed pattern is an unidimensional model for an energy group obtained of application of volumetric average diffusion equation in the moderator. The obtained results are compared against the classic diffusion theory and qualitatively against the neutron transport theory. (Author)
FY16 Status Report on NEAMS Neutronics Activities
Energy Technology Data Exchange (ETDEWEB)
Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Shemon, E. R. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Smith, M. A. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Jung, Y. S. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division
2016-09-30
The goal of the NEAMS neutronics effort is to develop a neutronics toolkit for use on sodium-cooled fast reactors (SFRs) which can be extended to other reactor types. The neutronics toolkit includes the high-fidelity deterministic neutron transport code PROTEUS and many supporting tools such as a cross section generation code MC^{2}-3, a cross section library generation code, alternative cross section generation tools, mesh generation and conversion utilities, and an automated regression test tool. The FY16 effort for NEAMS neutronics focused on supporting the release of the SHARP toolkit and existing and new users, continuing to develop PROTEUS functions necessary for performance improvement as well as the SHARP release, verifying PROTEUS against available existing benchmark problems, and developing new benchmark problems as needed. The FY16 research effort was focused on further updates of PROTEUS-SN and PROTEUS-MOCEX and cross section generation capabilities as needed.
First Neutron Spectrometry Measurement at the HL-2A Tokamak
Xi, Yuan; Xufei, Xie; Zhongjing, Chen; Xingyu, Peng; Tieshuan, Fan; Jinxiang, Chen; Xiangqing, Li; Guoliang, Yuan; Jinwei, Yang; Qingwei, Yang
2013-01-01
A compact neutron spectrometer based on the liquid scintillator is presented for the neutron energy spectrum measurement at the HL-2A tokamak. The spectrometer has been well characterized and a fast digital pulse shape discrimination software has been developed using the charge comparison method. A digitizer data acquisition system with the maximum frequency of 1 MHz can work under the high count rate environment at HL-2A. Specific radiation shielding and magnetic shielding for the spectrometerhas been designed for the neutron spectrum measurement at the HL-2A Tokamak. For the analysis of the pulse height spectrum, dedicated numerical simulation utilizing NUBEAM combining with GENESIS has been made to obtain the neutron energy spectrum, following which the transportation process from the plasma to the detector has been evaluated with Monte Carlo calculations. The distorted neutron energy spectrum has been folded with response matrix of the liquid scintillation spectrometer, and good consistency has been found...
Design of an integrating type neutron dose monitor.
Yamanishi, Hirokuni
2011-07-01
It is intended that deuterium-deuterium reaction experiments will be performed for the next phase of the large helical device (LHD) at National Institute for Fusion Science (NIFS), Toki, Japan. To protect workers against radiation, the characteristics of the radiation field at the LHD workplace should be evaluated. The neutron fluence at the workplace was calculated by means of the radiation transportation code. Since the neutron energy distribution at the workplace has a wide energy range, from thermal to fast neutrons, a neutron dose monitor had to be especially designed. The author designed an integrating type neutron dose monitor for this purpose. Since this monitor has good responses for dose evaluation in every energy range, it should be able to evaluate the dose at the LHD workplace accurately.
Characteristics of the WNR: a pulsed spallation neutron source
Energy Technology Data Exchange (ETDEWEB)
Russell, G.J.; Lisowski, P.W.; Howe, S.D.; King, N.S.P.; Meier, M.M.
1982-01-01
The Weapons Neutron Research facility (WNR) is a pulsed spallation neutron source in operation at the Los Alamos National Laboratory. The WNR uses part of the 800-MeV proton beam from the Clinton P. Anderson Meson Physics Facility accelerator. By choosing different target and moderator configurations and varying the proton pulse structure, the WNR can provide a white neutron source spanning the energy range from a few MeV to 800 MeV. The neutron spectrum from a bare target has been measured and is compared with predictions using an Intranuclear Cascade model coupled to a Monte Carlo transport code. Calculations and measurements of the neutronics of WNR target-moderator assemblies are presented.
Energy Technology Data Exchange (ETDEWEB)
Ku, L.P.; Hendel, H.W.; Liew, S.L.; Strachan, J.D.
1990-02-01
Accurate determinations of fusion neutron yields on the TFTR require that the neutron detectors be absolutely calibrated in-situ, using neutron sources of known strengths. For such calibrations, numerical simulations of neutron transport can be powerful tools in the design of experiments and the study of measurement results. On the TFTR, numerical calibration experiments' have been frequently used to complement actual detector calibrations. We present calculational approaches and transport models used in these numerical simulations, and summarize the results from simulating the calibration of {sup 235}U fission detectors carried out in December 1988. 12 refs., 9 figs., 6 tabs.
Neutron scattering at Australia's replacement research reactor
Robinson, R. A.; Kennedy, S. J.
2002-01-01
On August 25 1999, the Australian government gave final approval to build a research reactor to replace the existing HIFAR reactor at Lucas Heights. The replacement reactor, which will commence operation in 2005, will be multipurpose in function, with capabilities for both neutron-beam research and radioisotope production. Regarding beams, cold and thermal neutron sources are to be installed and the intent is to use supermirror guides, with coatings with critical angles up to 3 times that of natural Ni, to transport cold and thermal neutron beams into a large modern guide hall. The reactor and all the associated infrastructure, with the exception of the neutron beam instruments, is to be built by INVAP, SE and subcontractors in a turnkey contract. The goal is to have at least eight leading-edge neutron-beam instruments ready in 2005, and they will be developed by ANSTO and other contracted organisations, in consultation with the Australian user community and interested overseas parties. A review of the planned scientific capabilities, a description of the facility and a status report on the activities so far is given.
Neutron Angular Scatter Effects in 3DHZETRN: Quasi-Elastic
Wilson, John W.; Werneth, Charles M.; Slaba, Tony C.; Badavi, Francis F.; Reddell, Brandon D.; Bahadori, Amir A.
2017-01-01
The current 3DHZETRN code has a detailed three dimensional (3D) treatment of neutron transport based on a forward/isotropic assumption and has been compared to Monte Carlo (MC) simulation codes in various geometries. In most cases, it has been found that 3DHZETRN agrees with the MC codes to the extent they agree with each other. However, a recent study of neutron leakage from finite geometries revealed that further improvements to the 3DHZETRN formalism are needed. In the present report, angular scattering corrections to the neutron fluence are provided in an attempt to improve fluence estimates from a uniform sphere. It is found that further developments in the nuclear production models are required to fully evaluate the impact of transport model updates. A model for the quasi-elastic neutron production spectra is therefore developed and implemented into 3DHZETRN.
Neutron resonance spectroscopy
Energy Technology Data Exchange (ETDEWEB)
Gunsing, F
2005-06-15
The present document has been written in order to obtain the diploma 'Habilitation a Diriger des Recherches'. Since this diploma is indispensable to supervise thesis students, I had the intention to write a document that can be useful for someone starting in the field of neutron resonance spectroscopy. Although the here described topics are already described elsewhere, and often in more detail, it seemed useful to have most of the relevant information in a single document. A general introduction places the topic of neutron-nucleus interaction in a nuclear physics context. The large variations of several orders of magnitude in neutron-induced reaction cross sections are explained in terms of nuclear level excitations. The random character of the resonances make nuclear model calculation predictions impossible. Then several fields in physics where neutron-induced reactions are important and to which I have contributed in some way or another, are mentioned in a first synthetic chapter. They concern topics like parity nonconservation in certain neutron resonances, stellar nucleosynthesis by neutron capture, and data for nuclear energy applications. The latter item is especially important for the transmutation of nuclear waste and for alternative fuel cycles. Nuclear data libraries are also briefly mentioned. A second chapter details the R-matrix theory. This formalism is the foundation of the description of the neutron-nucleus interaction and is present in all fields of neutron resonance spectroscopy. (author)
Development of Neutron Spectrometer
Energy Technology Data Exchange (ETDEWEB)
Lee, Chang Hee; Lee, J. S.; Seong, B. S. (and others)
2007-06-15
Neutron spectrometers which are used in the basic researches such as physics, chemistry and materials science and applied in the industry were developed at the horizontal beam port of HANARO reactor. In addition, the development of core components for neutron scattering and the upgrade of existing facilities are also performed. The vertical neutron reflectometer was fabricated and installed at ST3 beam port. The performance test of the reflectometer was completed and the reflectometer was opened to users. The several core parts and options were added in the polarized neutron spectrometer. The horizontal neutron reflectometer from Brookhaven National Laboratory was moved to HANARO and installed, and the performance of the reflectometer was examined. The HIPD was developed and the performance test was completed. The base shielding for TAS was fabricated. The soller collimator, Cu mosaic monochromator, Si BPC monochromator and position sensitive detector were developed and applied in the neutron spectrometer as part of core component development activities. In addition, the sputtering machine for mirror device are fabricated and the neutron mirror is made using the sputtering machine. The FCD was upgraded and the performance of the FCD are improved over the factor of 10. The integration and upgrade of the neutron detection system were also performed.
Wambach, Jochen
2013-01-01
In this presentation I discuss two aspects of the neutron-matter equation of state. One relates to the symmetry energy of nuclear matter and empirical constraints on its slope parameter at saturation density. The second deals with spatially inhomogeneous chiral phases of deconfined quark matter in the inner core of a neutron star.
Energy Technology Data Exchange (ETDEWEB)
Frame, Katherine Chiyoko [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2017-06-28
Neutron multiplicity measurements are widely used for nondestructive assay (NDA) of special nuclear material (SNM). When combined with isotopic composition information, neutron multiplicity analysis can be used to estimate the spontaneous fission rate and leakage multiplication of SNM. When combined with isotopic information, the total mass of fissile material can also be determined. This presentation provides an overview of this technique.
Energy Technology Data Exchange (ETDEWEB)
Yanch, Jacquelyn C. (Cambridge, MA); Shefer, Ruth E. (Newton, MA); Klinkowstein, Robert E. (Winchester, MA)
1999-01-01
In one embodiment there is provided an application of the .sup.10 B(n,.alpha.).sup.7 Li nuclear reaction or other neutron capture reactions for the treatment of rheumatoid arthritis. This application, called Boron Neutron Capture Synovectomy (BNCS), requires substantially altered demands on neutron beam design than for instance treatment of deep seated tumors. Considerations for neutron beam design for the treatment of arthritic joints via BNCS are provided for, and comparisons with the design requirements for Boron Neutron Capture Therapy (BNCT) of tumors are made. In addition, exemplary moderator/reflector assemblies are provided which produce intense, high-quality neutron beams based on (p,n) accelerator-based reactions. In another embodiment there is provided the use of deuteron-based charged particle reactions to be used as sources for epithermal or thermal neutron beams for neutron capture therapies. Many d,n reactions (e.g. using deuterium, tritium or beryllium targets) are very prolific at relatively low deuteron energies.
Energy Technology Data Exchange (ETDEWEB)
Yanch, J.C.; Shefer, R.E.; Klinkowstein, R.E.
1999-11-02
In one embodiment there is provided an application of the {sup 10}B(n,{alpha}){sup 7}Li nuclear reaction or other neutron capture reactions for the treatment of rheumatoid arthritis. This application, called Boron Neutron Capture Synovectomy (BNCS), requires substantially altered demands on neutron beam design than for instance treatment of deep seated tumors. Considerations for neutron beam design for the treatment of arthritic joints via BNCS are provided for, and comparisons with the design requirements for Boron Neutron Capture Therapy (BNCT) of tumors are made. In addition, exemplary moderator/reflector assemblies are provided which produce intense, high-quality neutron beams based on (p,n) accelerator-based reactions. In another embodiment there is provided the use of deuteron-based charged particle reactions to be used as sources for epithermal or thermal neutron beams for neutron capture therapies. Many d,n reactions (e.g. using deuterium, tritium or beryllium targets) are very prolific at relatively low deuteron energies.
Simakov, S P; Heinzel, V; Moellendorff, U V
2002-01-01
The report describes the new results of the development work performed at Forschungszentrum Karlsruhe on the neutronics of the International Fusion Materials Irradiation Facility (IFMIF). An important step forward has been done in the simulation of neutron production of the deuteron-lithium source using the Li(d,xn) reaction cross sections from evaluated data files. The developed Monte Carlo routine and d-Li reaction data newly evaluated at INPE Obninsk have been verified against available experimental data on the differential neutron yield from deuteron-bombarded thick lithium targets. With the modified neutron source three-dimensional distributions of neutron and photon fluxes, displacement and gas production rates and nuclear heating inside the high flux test module (HFTM) were calculated. In order to estimate the uncertainty resulting from the evaluated data, two independent libraries, recently released by INPE and LANL, have been used in the transport calculations. The proposal to use a reflector around ...
Becker, Werner
2009-01-01
Neutron stars are the most compact astronomical objects in the universe which are accessible by direct observation. Studying neutron stars means studying physics in regimes unattainable in any terrestrial laboratory. Understanding their observed complex phenomena requires a wide range of scientific disciplines, including the nuclear and condensed matter physics of very dense matter in neutron star interiors, plasma physics and quantum electrodynamics of magnetospheres, and the relativistic magneto-hydrodynamics of electron-positron pulsar winds interacting with some ambient medium. Not to mention the test bed neutron stars provide for general relativity theories, and their importance as potential sources of gravitational waves. It is this variety of disciplines which, among others, makes neutron star research so fascinating, not only for those who have been working in the field for many years but also for students and young scientists. The aim of this book is to serve as a reference work which not only review...
THERMAL NEUTRON BACKSCATTER IMAGING.
Energy Technology Data Exchange (ETDEWEB)
VANIER,P.; FORMAN,L.; HUNTER,S.; HARRIS,E.; SMITH,G.
2004-10-16
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.
Pocked surface neutron detector
Energy Technology Data Exchange (ETDEWEB)
McGregor, Douglas (Whitmore Lake, MI); Klann, Raymond (Bolingbrook, IL)
2003-04-08
The detection efficiency, or sensitivity, of a neutron detector material such as of Si, SiC, amorphous Si, GaAs, or diamond is substantially increased by forming one or more cavities, or holes, in its surface. A neutron reactive material such as of elemental, or any compound of, .sup.10 B, .sup.6 Li, .sup.6 LiF, U, or Gd is deposited on the surface of the detector material so as to be disposed within the cavities therein. The portions of the neutron reactive material extending into the detector material substantially increase the probability of an energetic neutron reaction product in the form of a charged particle being directed into and detected by the neutron detector material.
Neutron stars - General review
Cameron, A. G. W.; Canuto, V.
1974-01-01
A review is presented of those properties of neutron stars upon which there is general agreement and of those areas which currently remain in doubt. Developments in theoretical physics of neutron star interiors are summarized with particular attention devoted to hyperon interactions and the structure of interior layers. Determination of energy states and the composition of matter is described for successive layers, beginning with the surface and proceeding through the central region into the core. Problems encountered in determining the behavior of matter in the ultra-high density regime are discussed, and the effects of the magnetic field of a neutron star are evaluated along with the behavior of atomic structures in the field. The evolution of a neutron star is outlined with discussion centering on carbon detonation, cooling, vibrational damping, rotation, and pulsar glitches. The role of neutron stars in cosmic-ray propagation is considered.
Conception d'un nouveau système de distribution de neutrons pour source à haut flux
Boffy, Romain
2016-01-01
The building of new experimental neutron beam facilities as well as the renewal programmes under development at some of the already existing installations have pinpointed the urgent need to develop neutron guide technology in order to make such neutron transport devices more efficient and durable. In fact, a number of mechanical failures of neutron guides have been reported by several research centres. It is therefore important to understand the behaviour of the glass substrates on top of whi...
SINGLE CRYSTAL NEUTRON DIFFRACTION.
Energy Technology Data Exchange (ETDEWEB)
KOETZLE,T.F.
2001-03-13
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.
Advanced Neutron Source (ANS) Project Progress report, FY 1991
Energy Technology Data Exchange (ETDEWEB)
Campbell, J.H. (ed.) (Oak Ridge National Lab., TN (United States)); Selby, D.L.; Harrington, R.M. (Oak Ridge National Lab., TN (United States)); Thompson, P.B. (Martin Marietta Energy Systems, Inc., (United States). Engineering Division)
1992-01-01
This report discusses the following about the Advanced Neutron Source: Project Management; Research and Development; Fuel Development; Corrosion Loop Tests and Analyses; Thermal-Hydraulic Loop Tests; Reactor Control and Shutdown 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; Design; and Safety.
Advanced Neutron Source (ANS) Project Progress report, FY 1991
Energy Technology Data Exchange (ETDEWEB)
Campbell, J.H. [ed.] [Oak Ridge National Lab., TN (United States); Selby, D.L.; Harrington, R.M. [Oak Ridge National Lab., TN (United States); Thompson, P.B. [Martin Marietta Energy Systems, Inc., (United States). Engineering Division
1992-01-01
This report discusses the following about the Advanced Neutron Source: Project Management; Research and Development; Fuel Development; Corrosion Loop Tests and Analyses; Thermal-Hydraulic Loop Tests; Reactor Control and Shutdown 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; Design; and Safety.
Neutron detection by scintillation of noble-gas excimers
McComb, Jacob Collin
Neutron detection is a technique essential to homeland security, nuclear reactor instrumentation, neutron diffraction science, oil-well logging, particle physics and radiation safety. The current shortage of helium-3, the neutron absorber used in most gas-filled proportional counters, has created a strong incentive to develop alternate methods of neutron detection. Excimer-based neutron detection (END) provides an alternative with many attractive properties. Like proportional counters, END relies on the conversion of a neutron into energetic charged particles, through an exothermic capture reaction with a neutron absorbing nucleus (10B, 6Li, 3He). As charged particles from these reactions lose energy in a surrounding gas, they cause electron excitation and ionization. Whereas most gas-filled detectors collect ionized charge to form a signal, END depends on the formation of diatomic noble-gas excimers (Ar*2, Kr*2,Xe* 2) . Upon decaying, excimers emit far-ultraviolet (FUV) photons, which may be collected by a photomultiplier tube or other photon detector. This phenomenon provides a means of neutron detection with a number of advantages over traditional methods. This thesis investigates excimer scintillation yield from the heavy noble gases following the boron-neutron capture reaction in 10B thin-film targets. Additionally, the thesis examines noble-gas excimer lifetimes with relationship to gas type and gas pressure. Experimental data were collected both at the National Institute of Standards and Technology (NIST) Center for Neutron Research, and on a newly developed neutron beamline at the Maryland University Training Reactor. The components of the experiment were calibrated at NIST and the University of Maryland, using FUV synchrotron radiation, neutron imaging, and foil activation techniques, among others. Computer modeling was employed to simulate charged-particle transport and excimer photon emission within the experimental apparatus. The observed excimer
Optical polarizing neutron devices designed for pulsed neutron sources
Energy Technology Data Exchange (ETDEWEB)
Takeda, M.; Kurahashi, K.; Endoh, Y. [Tohoku Univ, Sendai (Japan); Itoh, S. [National Lab. for High Energy Physics, Tsukuba (Japan)
1997-09-01
We have designed two polarizing neutron devices for pulsed cold neutrons. The devices have been tested at the pulsed neutron source at the Booster Synchrotron Utilization Facility of the National Laboratory for High Energy Physics. These two devices proved to have a practical use for experiments to investigate condensed matter physics using pulsed cold polarized neutrons.
Development of the RRR cold neutron beam facility
Energy Technology Data Exchange (ETDEWEB)
Lovotti, Osvaldo; Masriera, Nestor; Lecot, Carlos; Hergenreder, Daniel [INVAP S.E., Bariloche, Rio Negro (Argentina)
2002-07-01
This paper describes some general design issues on the neutron beam facilities (cold neutron source and neutron beam transport system) of the Replacement Research Reactor (RRR) for the Australian Nuclear Science and Technology Organisation (ANSTO). The description covers different aspect of the design: the requirements that lead to an innovative design, the overall design itself, the definition of a technical approach in order to develop the necessary design solutions, and finally the organizational framework by which international expertise from five different institutions is integrated. From the technical viewpoint, the RRR-CNS is a liquid Deuterium (LD2) moderator, sub-cooled to ensure maximum moderation efficiency, flowing within a closed natural circulation thermosyphon loop. The thermosyphon is surrounded by a zirconium alloy CNS vacuum containment that provides thermal insulation and a multiple barriers scheme to prevent Deuterium from mixing with water or air. Consistent with international practice, this vessel is designed to withstand any hypothetical energy reaction should Deuterium and air mix in its interior. The 'cold' neutrons are then taken by the NBTS and transported by the neutron guide system into the reactor beam hall and neutron guide hall, where neutron scattering instruments are located. From the management viewpoint, the adopted distributed scheme is successful to manage the complex interfacing between highly specialized technologies, allowing a smooth integration within the project. (author)
Collective Modes in the Superfluid Inner Crust of Neutron Stars
Urban, Michael
2015-01-01
The neutron-star inner crust is assumed to be superfluid at relevant temperatures. The contribution of neutron quasiparticles to thermodynamic and transport properties of the crust is therefore strongly suppressed by the pairing gap. Nevertheless, the neutron gas still has low-energy excitations, namely long-wavelength collective modes. We summarize different approaches to describe the collective modes in the crystalline phases of the inner crust and present an improved model for the description of the collective modes in the pasta phases within superfluid hydrodynamics.
Measurement of photoneutron spectrum at Pohang Neutron Facility
Kim, G N; Lee, Y S; Skoy, V; Cho, M H; Ko, I S; Namkung, W; Lee, D W; Kim, H D; Ko, S K; Park, S H; Kim, D S; Ro, T I; Min, Y G
2002-01-01
The Pohang Neutron Facility, an electron linear accelerator (linac) based pulsed neutron facility, was constructed for nuclear data production in Korea. It consists of an electron linac, a water-cooled Ta target with a water moderator, and a time-of-flight path with an 11 m length. The neutron energy spectra are measured for different water levels inside the moderator and compared with calculations by the Monte Carlo N-Particle transport code. The optimum size of the water moderator is determined based on these results.
Tanaka, H.; Sakurai, Y.; Suzuki, M.; Masunaga, S.; Kinashi, Y.; Kashino, G.; Liu, Y.; Mitsumoto, T.; Yajima, S.; Tsutsui, H.; Maruhashi, A.; Ono, K.
2009-06-01
At Kyoto University Research Reactor Institute (KURRI), 275 clinical trials of boron neutron capture therapy (BNCT) have been performed as of March 2006, and the effectiveness of BNCT has been revealed. In order to further develop BNCT, it is desirable to supply accelerator-based epithermal-neutron sources that can be installed near the hospital. We proposed the method of filtering and moderating fast neutrons, which are emitted from the reaction between a beryllium target and 30-MeV protons accelerated by a cyclotron accelerator, using an optimum moderator system composed of iron, lead, aluminum and calcium fluoride. At present, an epithermal-neutron source is under construction from June 2008. This system consists of a cyclotron accelerator, beam transport system, neutron-yielding target, filter, moderator and irradiation bed. In this article, an overview of this system and the properties of the treatment neutron beam optimized by the MCNPX Monte Carlo neutron transport code are presented. The distribution of biological effect weighted dose in a head phantom compared with that of Kyoto University Research Reactor (KUR) is shown. It is confirmed that for the accelerator, the biological effect weighted dose for a deeply situated tumor in the phantom is 18% larger than that for KUR, when the limit dose of the normal brain is 10 Gy-eq. The therapeutic time of the cyclotron-based neutron sources are nearly one-quarter of that of KUR. The cyclotron-based epithermal-neutron source is a promising alternative to reactor-based neutron sources for treatments by BNCT.
Fail-safe neutron shutter used for thermal neutron radiography
Energy Technology Data Exchange (ETDEWEB)
Sachs, R.D.; Morris, R.A.
1976-11-01
A fail-safe, reliable, easy-to-use neutron shutter was designed, built, and put into operation at the Omega West Reactor, Los Alamos Scientific Laboratory. The neutron shutter will be used primarily to perform thermal neutron radiography, but is also available for a highly collimated source of thermal neutrons (neutron flux = 3.876 x 10/sup 6/ (neutrons)/(cm/sup 2/.s)). Neutron collimator sizes of either 10.16 by 10.16 cm or 10.16 by 30.48 cm are available.
Neutron scattering effects on fusion ion temperature measurements.
Energy Technology Data Exchange (ETDEWEB)
Ziegler, Lee (Bechtel/Nevada, Las Vegas, NV); Starner, Jason R.; Cooper, Gary Wayne; Ruiz, Carlos L.; Franklin, James Kenneth (Ktech Corporation, Albuquerque, NM); Casey, Daniel T.
2006-06-01
To support the nuclear fusion program at Sandia National Laboratories (SNL), a consistent and verifiable method to determine fusion ion temperatures needs to be developed. Since the fusion temperature directly affects the width in the spread of neutron energies produced, a measurement of the neutron energy width can yield the fusion temperature. Traditionally, the spread in neutron energies is measured by using time-of-flight to convert a spread in neutron energies at the source to a spread in time at detector. One potential obstacle to using this technique at the Z facility at SNL is the need to shield the neutron detectors from the intense bremsstrahlung produced. The shielding consists of eight inches of lead and the concern is that neutrons will scatter in the lead, artificially broaden the neutron pulse width and lead to an erroneous measurement. To address this issue, experiments were performed at the University of Rochester's Laboratory for Laser Energetics, which demonstrated that a reliable ion temperature measurement can be achieved behind eight inches of lead shielding. To further expand upon this finding, Monte Carlo N-Particle eXtended (MCNPX) was used to simulate the experimental geometric conditions and perform the neutron transport. MCNPX was able to confidently estimate results observed at the University of Rochester.
Energy Technology Data Exchange (ETDEWEB)
Do, Changwoo, E-mail: doc1@ornl.gov [Biology and Soft Matter Division, Neutron Sciences Directorate, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Heller, William T.; Stanley, Christopher [Biology and Soft Matter Division, Neutron Sciences Directorate, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Gallmeier, Franz X. [Instrument and Source Design Division, Neutron Sciences Directorate, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Doucet, Mathieu [Neutron Data Analysis and Visualization Division, Neutron Sciences Directorate, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Smith, Gregory S. [Biology and Soft Matter Division, Neutron Sciences Directorate, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)
2014-02-11
It is generally assumed by most of the small-angle neutron scattering (SANS) user community that a neutron's energy is unchanged during SANS measurements. Here, the scattering from water, specifically light water, was measured on the EQ-SANS instrument, a time-of-flight (TOF) SANS instrument located at the Spallation Neutron Source of Oak Ridge National Laboratory. A significant inelastic process was observed in the TOF spectra of neutrons scattered from water. Analysis of the TOF spectra from the sample showed that the scattered neutrons have energies consistent with room-temperature thermal energies (∼20 meV) regardless of the incident neutron's energy. With the aid of Monte Carlo particle transport simulations, we conclude that the thermalization process within the sample results in faster neutrons that arrive at the detector earlier than expected based on the incident neutron energies. This thermalization process impacts the measured SANS intensities in a manner that will ultimately be sample- and temperature-dependent, necessitating careful processing of the raw data into the SANS cross-section.
Development of In-pile Plug Assembly and Primary Shutter for Cold Neutron Guide System
Energy Technology Data Exchange (ETDEWEB)
Shin, Jin Won; Cho, Yeong Garp; Ryu, Jeong Soo; Lee, Jung Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2008-12-15
The HANARO, a 30 MW multi-purpose research 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. The functions of the in-pile plug assembly are to shield the reactor environment from a nuclear radiation and to support the neutron guides and maintain them precisely oriented. The primary shutter is a mechanical device to be installed just after the in-pile plug assembly, which stops neutron flux on demand. This report describes the mechanical design, fabrication, and installation procedure of the in-pile plug assembly and the primary shutter for the neutron guide system at HANARO. A special tool and procedure for a replacement of in-pile plug and guide cassette is also presented with the interface condition in the reactor hall.
Energy Technology Data Exchange (ETDEWEB)
Barocchi, F. [Florence Univ. (Italy). Ist. di Fisica
1996-12-31
Together with X-rays, thermal neutrons are the ideal probe to study the microscopic structure of condensed matter, however the precision attainable usually with neutrons for the measurement of atomic position correlation functions in liquids is, at least, one order of magnitude better than for X-rays. In order to measure properly the microscopic dynamics a wide range of momentum transfer with corresponding energy transfer must be available in the range of liquid state excitations. This again is only attainable, with good resolution, with neutrons. (author) 7 figs., 3 refs.
Imae, Toyoko; Furusaka, Michihiro; Torikai, Naoya
2011-01-01
Neutron and synchrotron facilities, which are beyond the scale of the laboratory, and supported on a national level in countries throughout the world. These tools for probing micro- and nano-structure research and on fast dynamics research of atomic location in materials have been key in the development of new polymer-based materials. Different from several existing professional books on neutron science, this book focuses on theory, instrumentation, an applications. The book is divided into five parts: Part 1 describes the underlying theory of neutron scattering. Part 2 desc
Nuclear characteristics of epoxy resin as a space environment neutron shielding
Energy Technology Data Exchange (ETDEWEB)
Adeli, Ruhollah [Nuclear Science and Technology Research Institute, Yazd (Iran, Islamic Republic of). Central Iran Research Complex; Shirmardi, Seyed Pezhman [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of). Radiation Application Research School; Mazinani, Saideh [Amirkabir Nanotechnology Research Institute, Tehran (Iran, Islamic Republic of); Ahmadi, Seyed Javad [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of). Nuclear Fuel Cycle Research School
2017-03-15
In recent years many investigations have been done for choosing applicable light neutron shielding in space environmental applications. In this study, we have considered the neutron radiation-protective characteristics of neat epoxy resin, a thermoplastic polymer material and have compared it with various candidate materials in neutron radiation protection such as Al 6061 alloy and Polyethylene. The aim of this investigation is the effect of type of moderator for fast neutron, notwithstanding neutron absorbers fillers. The nuclear interactions and the effective dose at shields have been studied with the Monte Carlo N-Particle transport code (MCNP), using variance reductions to reduce the relative error. Among the candidates, polymer matrix showed a better performance in attenuating fast neutrons and caused a lower neutron and secondary photon effective dose.
Prompt fission neutron investigation in 235U(nth,f) reaction
Zeynalov, Shakir; Sedyshev, Pavel; Shvetsov, Valery; Sidorova, Olga
2017-09-01
The prompt neutron emission in thermal neutron induced fission of 235U has been investigated applying digital signal electronics. The goal was to compare the results of this digital data acquisition and digital signal processing analysis to the results of the pioneering work of Apalin et al. Using a twin Frisch grid ionization chamber for the fission fragment detection and a NE213 equivalent neutron detector in total about 106 neutron coincidences have been registered. The fission fragment kinetic energy, mass and angular distribution has been investigated along with prompt neutron time of flight and pulse shape using a six channel synchronous waveform digitizer with sampling frequency of 250 MHz and 12 bit resolution. The signals have been analyzed using digital pulse processing algorithms, developed by authors. The thermal neutron beam was transported from the IBR-2 reactor to the target with bent mirror neutron guide.
Recent neutronics developments for reactor safety studies with SIMMER code at KIT
Rineiski, A.; Marchetti, M.; Andriolo, L.; Gabrielli, F.
2017-01-01
The SIMMER family of codes is applied for safety studies of sodium fast reactors and reactors of other types. Both neutronics and fluid-dynamics parts of SIMMER are under development. In the paper new neutronics capabilities are presented. In particular developments for neutron transport solvers and a new technique for taking into account thermal expansion effects are described. These new capabilities facilitate 3D simulations and improve accuracy of modelling for the initiation transient phase during a hypothetical severe accident.
Developing an interface between MCNP and McStas for simulation of neutron moderators
Klinkby, Esben Bryndt; Lauritzen, Bent; Nonbøl, Erik; Willendrup, Peter Kjær
2012-01-01
Simulations of target-moderator-reflector system at spallation sources are conventionally carried out using MCNP/X whereas simulations of neutron transport and instrument performance are carried out by neutron ray tracing codes such as McStas. The coupling between the two simulations suites typically consists of providing analytical fits from MCNP/X neutron spectra to McStas. This method is generally successful, but as will be discussed in the this paper, there are limitations and a more dire...
3DHZETRN: Neutron leakage in finite objects.
Wilson, John W; Slaba, Tony C; Badavi, Francis F; Reddell, Brandon D; Bahadori, Amir A
2015-11-01
The 3DHZETRN formalism was recently developed as an extension to HZETRN with an emphasis on 3D corrections for neutrons and light ions. Comparisons to Monte Carlo (MC) simulations were used to verify the 3DHZETRN methodology in slab and spherical geometry, and it was shown that 3DHZETRN agrees with MC codes to the degree that various MC codes agree among themselves. One limitation of such comparisons is that all of the codes (3DHZETRN and three MC codes) utilize different nuclear models/databases; additionally, using a common nuclear model is impractical due to the complexity of the software. It is therefore difficult to ascertain if observed discrepancies are caused by transport code approximations or nuclear model differences. In particular, an important simplification in the 3DHZETRN formalism assumes that neutron production cross sections can be represented as the sum of forward and isotropic components, where the forward component is subsequently solved within the straight-ahead approximation. In the present report, previous transport model results in specific geometries are combined with additional results in related geometries to study neutron leakage using the Webber 1956 solar particle event as a source boundary condition. A ratio is defined to quantify the fractional neutron leakage at a point in a finite object relative to a semi-infinite slab geometry. Using the leakage ratio removes some of the dependence on the magnitude of the neutron production and clarifies the effects of angular scattering and absorption with regard to differences between the models. Discussion is given regarding observed differences between the MC codes and conclusions drawn about the need for further code development. Although the current version of 3DHZETRN is reasonably accurate compared to MC simulations, this study shows that improved leakage estimates can be obtained by replacing the isotropic/straight-ahead approximation with more detailed descriptions. Published by
3DHZETRN: Neutron leakage in finite objects
Wilson, John W.; Slaba, Tony C.; Badavi, Francis F.; Reddell, Brandon D.; Bahadori, Amir A.
2015-11-01
The 3DHZETRN formalism was recently developed as an extension to HZETRN with an emphasis on 3D corrections for neutrons and light ions. Comparisons to Monte Carlo (MC) simulations were used to verify the 3DHZETRN methodology in slab and spherical geometry, and it was shown that 3DHZETRN agrees with MC codes to the degree that various MC codes agree among themselves. One limitation of such comparisons is that all of the codes (3DHZETRN and three MC codes) utilize different nuclear models/databases; additionally, using a common nuclear model is impractical due to the complexity of the software. It is therefore difficult to ascertain if observed discrepancies are caused by transport code approximations or nuclear model differences. In particular, an important simplification in the 3DHZETRN formalism assumes that neutron production cross sections can be represented as the sum of forward and isotropic components, where the forward component is subsequently solved within the straight-ahead approximation. In the present report, previous transport model results in specific geometries are combined with additional results in related geometries to study neutron leakage using the Webber 1956 solar particle event as a source boundary condition. A ratio is defined to quantify the fractional neutron leakage at a point in a finite object relative to a semi-infinite slab geometry. Using the leakage ratio removes some of the dependence on the magnitude of the neutron production and clarifies the effects of angular scattering and absorption with regard to differences between the models. Discussion is given regarding observed differences between the MC codes and conclusions drawn about the need for further code development. Although the current version of 3DHZETRN is reasonably accurate compared to MC simulations, this study shows that improved leakage estimates can be obtained by replacing the isotropic/straight-ahead approximation with more detailed descriptions.
Neutron flux reduction programs for reactor pressure vessel
Energy Technology Data Exchange (ETDEWEB)
Yoo, C.S. [Korea Atomic Energy Research Inst. KAERI, 150 Deogjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Kim, B.C. [Korea Reactor Integrity Surveillance Technology KRIST, 150 Deogjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)
2011-07-01
The objective of this work is to implement various fast neutron flux reduction programs on the belt-line region of the reactor pressure vessel to reduce the increasing rate of reference temperature for pressurized thermal shock (RT PTS) for Korea Nuclear Unit 1. A pressurized thermal shock (PTS) event is an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. A PTS concern arises if one of these transients acts in the belt-line region of a reactor vessel where a reduced fracture resistance exists because of neutron irradiation. Generally, the RT PTS value is continuously increasing according to the fast neutron irradiation during the reactor operation, and it can reach the screening criterion prior to the expiration of the operating license. To reduce the increasing rate of RT PTS, various neutron flux reduction programs can be implemented, which are focused on license renewal. In this paper, neutron flux reduction programs, such as low leakage loading pattern strategy, loading of neutron absorber rods, and dummy fuel assembly loading are considered for Korea Nuclear Unit 1, of which the RT PTS value of the leading material (circumferential weld) is going to reach the screening criterion in the near future. To evaluate the effects of the neutron flux reduction programs, plant and cycle specific forward neutron transport calculations for the various neutron flux reduction programs were carried out. For the analysis, all transport calculations were carried out by using the DORT 3.1 discrete ordinate code and BUGLE-96 cross-section library. (authors)
Neutron signal transfer analysis
Pleinert, H; Lehmann, E
1999-01-01
A new method called neutron signal transfer analysis has been developed for quantitative determination of hydrogenous distributions from neutron radiographic measurements. The technique is based on a model which describes the detector signal obtained in the measurement as a result of the action of three different mechanisms expressed by signal transfer functions. The explicit forms of the signal transfer functions are determined by Monte Carlo computer simulations and contain only the distribution as a variable. Therefore an unknown distribution can be determined from the detector signal by recursive iteration. This technique provides a simple and efficient tool for analysis of this type while also taking into account complex effects due to the energy dependency of neutron interaction and single and multiple scattering. Therefore this method provides an efficient tool for precise quantitative analysis using neutron radiography, as for example quantitative determination of moisture distributions in porous buil...
Tsuruta, Sachiko
2016-07-01
We start with a brief introduction to the historical background in the early pioneering days when the first neutron star thermal evolution calculations predicted the presence of neutron stars hot enough to be observable. We then report on the first detection of neutron star temperatures by ROSAT X-ray satellite, which vindicated the earlier prediction of hot neutron stars. We proceed to present subsequent developments, both in theory and observation, up to today. We then discuss the current status and the future prospect, which will offer useful insight to the understanding of basic properties of ultra-high density matter beyond the nuclear density, such as the possible presence of such exotic particles as pion condensates.
Decoherence Free Neutron Interferometry
Pushin, Dmitry A; Cory, David G
2016-01-01
Perfect single-crystal neutron interferometers are adversely sensitive to environmental disturbances, particularly mechanical vibrations. The sensitivity to vibrations results from the slow velocity of thermal neutrons and the long measurement time that are encountered in a typical experiment. Consequently, to achieve a good interference solutions for reducing vibration other than those normally used in optical experiments must be explored. Here we introduce a geometry for a neutron interferometer that is less sensitive to low-frequency vibrations. This design may be compared with both dynamical decoupling methods and decoherence-free subspaces that are described in quantum information processing. By removing the need for bulky vibration isolation setups, this design will make it easier to adopt neutron interferometry to a wide range of applications and increase its sensitivity.
Neutron Stars Recent Developments
Heiselberg, H
1999-01-01
Recent developments in neutron star theory and observation are discussed. Based on modern nucleon-nucleon potentials more reliable equations of state for dense nuclear matter have been constructed. Furthermore, phase transitions such as pion, kaon and hyperon condensation, superfluidity and quark matter can occur in cores of neutron stars. Specifically, the nuclear to quark matter phase transition and its mixed phases with intriguing structures is treated. Rotating neutron stars with and without phase transitions are discussed and compared to observed masses, radii and glitches. The observations of possible heavy $\\sim 2M_\\odot$ neutron stars in X-ray binaries and QPO's require relatively stiff equation of states and restrict strong phase transitions to occur at very high nuclear densities only.
Energy Technology Data Exchange (ETDEWEB)
Lewis, W.B
1966-07-01
The presentation discusses both the economic and research contexts that would be served by producing neutrons in gram quantities at high intensities by electrical means without uranium-235. The revenue from producing radioisotopes is attractive. The array of techniques introduced by the multipurpose 65 megawatt Intense Neutron Generator project includes liquid metal cooling, superconducting magnets for beam bending and focussing, super-conductors for low-loss high-power radiofrequency systems, efficient devices for producing radiofrequency power, plasma physics developments for producing and accelerating hydrogen, ions at high intensity that are still far out from established practice, a multimegawatt high voltage D.C. generating machine that could have several applications. The research fields served relate principally to materials science through neutron-phonon and other quantum interactions as well as through neutron diffraction. Nuclear physics is served through {mu}-, {pi}- and K-meson production. Isotope production enters many fields of applied research. (author)
Introduction to neutron scattering
Energy Technology Data Exchange (ETDEWEB)
Fischer, W.E. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)
1996-11-01
We give here an introduction to the theoretical principles of neutron scattering. The relationship between scattering- and correlation-functions is particularly emphasized. Within the framework of linear response theory (justified by the weakness of the basic interaction) the relation between fluctuation and dissipation is discussed. This general framework explains the particular power of neutron scattering as an experimental method. (author) 4 figs., 4 refs.
DEFF Research Database (Denmark)
Gudel, H. U.; Furrer, A.; Kjems, Jørgen
1986-01-01
Insulating compounds containing dimers of transition metal and rare earth ions have been studied by inelastic neutron scattering (INS). Energy splittings can be directly determined, and the corresponding parameters are easily extracted from the experimental data. The intensities of dimer excitati......Insulating compounds containing dimers of transition metal and rare earth ions have been studied by inelastic neutron scattering (INS). Energy splittings can be directly determined, and the corresponding parameters are easily extracted from the experimental data. The intensities of dimer...
Pulsed spallation Neutron Sources
Energy Technology Data Exchange (ETDEWEB)
Carpenter, J.M. [Argonne National Lab., IL (United States)
1994-12-31
This paper reviews the early history of pulsed spallation neutron source development at Argonne and provides an overview of existing sources world wide. A number of proposals for machines more powerful than currently exist are under development, which are briefly described. The author reviews the status of the Intense Pulsed Neutron Source, its instrumentation, and its user program, and provides a few examples of applications in fundamental condensed matter physics, materials science and technology.
NSCool: Neutron star cooling code
Page, Dany
2016-09-01
NSCool is a 1D (i.e., spherically symmetric) neutron star cooling code written in Fortran 77. The package also contains a series of EOSs (equation of state) to build stars, a series of pre-built stars, and a TOV (Tolman- Oppenheimer-Volkoff) integrator to build stars from an EOS. It can also handle “strange stars” that have a huge density discontinuity between the quark matter and the covering thin baryonic crust. NSCool solves the heat transport and energy balance equations in whole GR, resulting in a time sequence of temperature profiles (and, in particular, a Teff - age curve). Several heating processes are included, and more can easily be incorporated. In particular it can evolve a star undergoing accretion with the resulting deep crustal heating, under a steady or time-variable accretion rate. NSCool is robust, very fast, and highly modular, making it easy to add new subroutines for new processes.
Neutron scattering in Australia
Energy Technology Data Exchange (ETDEWEB)
Knott, R.B. [Australian Nuclear Science and Technology Organisation, Menai (Australia)
1994-12-31
Neutron scattering techniques have been part of the Australian scientific research community for the past three decades. The High Flux Australian Reactor (HIFAR) is a multi-use facility of modest performance that provides the only neutron source in the country suitable for neutron scattering. The limitations of HIFAR have been recognized and recently a Government initiated inquiry sought to evaluate the future needs of a neutron source. In essence, the inquiry suggested that a delay of several years would enable a number of key issues to be resolved, and therefore a more appropriate decision made. In the meantime, use of the present source is being optimized, and where necessary research is being undertaken at major overseas neutron facilities either on a formal or informal basis. Australia has, at present, a formal agreement with the Rutherford Appleton Laboratory (UK) for access to the spallation source ISIS. Various aspects of neutron scattering have been implemented on HIFAR, including investigations of the structure of biological relevant molecules. One aspect of these investigations will be presented. Preliminary results from a study of the interaction of the immunosuppressant drug, cyclosporin-A, with reconstituted membranes suggest that the hydrophobic drug interdigitated with lipid chains.
Energy Technology Data Exchange (ETDEWEB)
Bingham, Philip R [ORNL; Santos-Villalobos, Hector J [ORNL
2011-01-01
Coded aperture techniques have been applied to neutron radiography to address limitations in neutron flux and resolution of neutron detectors in a system labeled coded source imaging (CSI). By coding the neutron source, a magnified imaging system is designed with small spot size aperture holes (10 and 100 m) for improved resolution beyond the detector limits and with many holes in the aperture (50% open) to account for flux losses due to the small pinhole size. An introduction to neutron radiography and coded aperture imaging is presented. A system design is developed for a CSI system with a development of equations for limitations on the system based on the coded image requirements and the neutron source characteristics of size and divergence. Simulation has been applied to the design using McStas to provide qualitative measures of performance with simulations of pinhole array objects followed by a quantitative measure through simulation of a tilted edge and calculation of the modulation transfer function (MTF) from the line spread function. MTF results for both 100um and 10um aperture hole diameters show resolutions matching the hole diameters.
Bingham, Philip; Santos-Villalobos, Hector; Tobin, Ken
2011-03-01
Coded aperture techniques have been applied to neutron radiography to address limitations in neutron flux and resolution of neutron detectors in a system labeled coded source imaging (CSI). By coding the neutron source, a magnified imaging system is designed with small spot size aperture holes (10 and 100μm) for improved resolution beyond the detector limits and with many holes in the aperture (50% open) to account for flux losses due to the small pinhole size. An introduction to neutron radiography and coded aperture imaging is presented. A system design is developed for a CSI system with a development of equations for limitations on the system based on the coded image requirements and the neutron source characteristics of size and divergence. Simulation has been applied to the design using McStas to provide qualitative measures of performance with simulations of pinhole array objects followed by a quantitative measure through simulation of a tilted edge and calculation of the modulation transfer function (MTF) from the line spread function. MTF results for both 100μm and 10μm aperture hole diameters show resolutions matching the hole diameters.
Neutron Nucleic Acid Crystallography.
Chatake, Toshiyuki
2016-01-01
The hydration shells surrounding nucleic acids and hydrogen-bonding networks involving water molecules and nucleic acids are essential interactions for the structural stability and function of nucleic acids. Water molecules in the hydration shells influence various conformations of DNA and RNA by specific hydrogen-bonding networks, which often contribute to the chemical reactivity and molecular recognition of nucleic acids. However, X-ray crystallography could not provide a complete description of structural information with respect to hydrogen bonds. Indeed, X-ray crystallography is a powerful tool for determining the locations of water molecules, i.e., the location of the oxygen atom of H2O; however, it is very difficult to determine the orientation of the water molecules, i.e., the orientation of the two hydrogen atoms of H2O, because X-ray scattering from the hydrogen atom is very small.Neutron crystallography is a specialized tool for determining the positions of hydrogen atoms. Neutrons are not diffracted by electrons, but are diffracted by atomic nuclei; accordingly, neutron scattering lengths of hydrogen and its isotopes are comparable to those of non-hydrogen atoms. Therefore, neutron crystallography can determine both of the locations and orientations of water molecules. This chapter describes the current status of neutron nucleic acid crystallographic research as well as the basic principles of neutron diffraction experiments performed on nucleic acid crystals: materials, crystallization, diffraction experiments, and structure determination.
Neutronic analysis of the 1D and 1E banks reflux detection system
Energy Technology Data Exchange (ETDEWEB)
Blanchard, A.
1999-12-21
Two H Canyon neutron monitoring systems for early detection of postulated abnormal reflux conditions in the Second Uranium Cycle 1E and 1D Mixer-Settle Banks have been designed and built. Monte Carlo neutron transport simulations using the general purpose, general geometry, n-particle MCNP code have been performed to model expected response of the monitoring systems to varying conditions.The confirmatory studies documented herein conclude that the 1E and 1D neutron monitoring systems are able to achieve adequate neutron count rates for various neutron source and detector configurations, thereby eliminating excessive integration count time. Neutron count rate sensitivity studies are also performed. Conversely, the transport studies concluded that the neutron count rates are statistically insensitive to nitric acid content in the aqueous region and to the transition region length. These studies conclude that the 1E and 1D neutron monitoring systems are able to predict the postulated reflux conditions for all examined perturbations in the neutron source and detector configurations. In the cases examined, the relative change in the neutron count rates due to postulated transitions from normal {sup 235}U concentration levels to reflux levels remain satisfactory detectable.
Energy Technology Data Exchange (ETDEWEB)
ZALIZNYAK,I.A.; LEE,S.H.
2004-07-30
Much of our understanding of the atomic-scale magnetic structure and the dynamical properties of solids and liquids was gained from neutron-scattering studies. Elastic and inelastic neutron spectroscopy provided physicists with an unprecedented, detailed access to spin structures, magnetic-excitation spectra, soft-modes and critical dynamics at magnetic-phase transitions, which is unrivaled by other experimental techniques. Because the neutron has no electric charge, it is an ideal weakly interacting and highly penetrating probe of matter's inner structure and dynamics. Unlike techniques using photon electric fields or charged particles (e.g., electrons, muons) that significantly modify the local electronic environment, neutron spectroscopy allows determination of a material's intrinsic, unperturbed physical properties. The method is not sensitive to extraneous charges, electric fields, and the imperfection of surface layers. Because the neutron is a highly penetrating and non-destructive probe, neutron spectroscopy can probe the microscopic properties of bulk materials (not just their surface layers) and study samples embedded in complex environments, such as cryostats, magnets, and pressure cells, which are essential for understanding the physical origins of magnetic phenomena. Neutron scattering is arguably the most powerful and versatile experimental tool for studying the microscopic properties of the magnetic materials. The magnitude of the cross-section of the neutron magnetic scattering is similar to the cross-section of nuclear scattering by short-range nuclear forces, and is large enough to provide measurable scattering by the ordered magnetic structures and electron spin fluctuations. In the half-a-century or so that has passed since neutron beams with sufficient intensity for scattering applications became available with the advent of the nuclear reactors, they have became indispensable tools for studying a variety of important areas of modern
Energy Technology Data Exchange (ETDEWEB)
ZALIZNYAK,I.A.; LEE,S.H.
2004-07-30
Much of our understanding of the atomic-scale magnetic structure and the dynamical properties of solids and liquids was gained from neutron-scattering studies. Elastic and inelastic neutron spectroscopy provided physicists with an unprecedented, detailed access to spin structures, magnetic-excitation spectra, soft-modes and critical dynamics at magnetic-phase transitions, which is unrivaled by other experimental techniques. Because the neutron has no electric charge, it is an ideal weakly interacting and highly penetrating probe of matter's inner structure and dynamics. Unlike techniques using photon electric fields or charged particles (e.g., electrons, muons) that significantly modify the local electronic environment, neutron spectroscopy allows determination of a material's intrinsic, unperturbed physical properties. The method is not sensitive to extraneous charges, electric fields, and the imperfection of surface layers. Because the neutron is a highly penetrating and non-destructive probe, neutron spectroscopy can probe the microscopic properties of bulk materials (not just their surface layers) and study samples embedded in complex environments, such as cryostats, magnets, and pressure cells, which are essential for understanding the physical origins of magnetic phenomena. Neutron scattering is arguably the most powerful and versatile experimental tool for studying the microscopic properties of the magnetic materials. The magnitude of the cross-section of the neutron magnetic scattering is similar to the cross-section of nuclear scattering by short-range nuclear forces, and is large enough to provide measurable scattering by the ordered magnetic structures and electron spin fluctuations. In the half-a-century or so that has passed since neutron beams with sufficient intensity for scattering applications became available with the advent of the nuclear reactors, they have became indispensable tools for studying a variety of important areas of modern
Fundamental neutron physics at LANSCE
Energy Technology Data Exchange (ETDEWEB)
Greene, G.
1995-10-01
Modern neutron sources and science share a common origin in mid-20th-century scientific investigations concerned with the study of the fundamental interactions between elementary particles. Since the time of that common origin, neutron science and the study of elementary particles have evolved into quite disparate disciplines. The neutron became recognized as a powerful tool for studying condensed matter with modern neutron sources being primarily used (and justified) as tools for neutron scattering and materials science research. The study of elementary particles has, of course, led to the development of rather different tools and is now dominated by activities performed at extremely high energies. Notwithstanding this trend, the study of fundamental interactions using neutrons has continued and remains a vigorous activity at many contemporary neutron sources. This research, like neutron scattering research, has benefited enormously by the development of modern high-flux neutron facilities. Future sources, particularly high-power spallation sources, offer exciting possibilities for continuing this research.
Deficiency in Monte Carlo simulations of coupled neutron-gamma-ray fields
Maleka, Peane P.; Maucec, Marko; de Meijer, Robert J.
2011-01-01
The deficiency in Monte Carlo simulations of coupled neutron-gamma-ray field was investigated by benchmarking two simulation codes with experimental data. Simulations showed better correspondence with the experimental data for gamma-ray transport only. In simulations, the neutron interactions with m
Rivard, Mark Joseph
Examination of neutron dosimetry for 252Cf has been conducted using calculative and experimental means. Monte Carlo N-Particle (MCNP) transport code was used in a distributed computing environment as a parallel virtual machine (PVM) to determine the absorbed neutron dose and neutron energy spectrum from 252Cf in a variety of clinically relevant materials. Herein, a Maxwellian spectrum was used to model the 252Cf neutron emissions within these materials. 252Cf mixed-field dosimetry of Applicator Tube (AT) type sources was measured using 1.0 and 0.05 cm3 tissue-equivalent ion chambers and a miniature GM counter. A dosimetry protocol was formulated similar that of ICRU 45. The 252Cf AT neutron dosimetry was determined in the cylindrical coordinate system formalism recommended by the AAPM Task Group 43. These results demonstrated the overwhelming dependence of dosimetry on the source geometry factor as there was no significant neutron attenuation within the source or encapsulation. Gold foils and TLDs were used to measure the thermal flux in the vicinity of 252Cf AT sources to compare with the results calculated using MCNP. As the fast neutron energy spectrum did not markedly changed at increasing distances from the AT source, neutron dosimetry results obtained with paired ion chambers using fixed sensitivity factors agreed well with MCNP results and those in the literature. Calculations of moderated 252Cf neutron energy spectrum with various loadings of 10B and 157Gd were performed, in addition to analysis of neutron capture therapy dosimetry with these isotopes. Radiological concerns such as personnel exposure and shielding of 252Cf emissions were examined. Feasibility of a high specific-activity 252Cf HDR source was investigated through radiochemical and metallurgical studies using stand-ins such as Tb, Gd and 249Cf. Issues such as capsule burst strength due to helium production for a variety of proposed HDR sources were addressed. A recommended 252Cf source
Designing accelerator-based epithermal neutron beams for boron neutron capture therapy.
Bleuel, D L; Donahue, R J; Ludewigt, B A; Vujic, J
1998-09-01
The 7Li(p,n)7Be reaction has been investigated as an accelerator-driven neutron source for proton energies between 2.1 and 2.6 MeV. Epithermal neutron beams shaped by three moderator materials, Al/AlF3, 7LiF, and D2O, have been analyzed and their usefulness for boron neutron capture therapy (BNCT) treatments evaluated. Radiation transport through the moderator assembly has been simulated with the Monte Carlo N-particle code (MCNP). Fluence and dose distributions in a head phantom were calculated using BNCT treatment planning software. Depth-dose distributions and treatment times were studied as a function of proton beam energy and moderator thickness. It was found that an accelerator-based neutron source with Al/AlF3 or 7LiF as moderator material can produce depth-dose distributions superior to those calculated for a previously published neutron beam design for the Brookhaven Medical Research Reactor, achieving up to approximately 50% higher doses near the midline of the brain. For a single beam treatment, a proton beam current of 20 mA, and a 7LiF moderator, the treatment time was estimated to be about 40 min. The tumor dose deposited at a depth of 8 cm was calculated to be about 21 Gy-Eq.
Radiography with polarised neutrons
Energy Technology Data Exchange (ETDEWEB)
Schulz, Michael L.
2010-08-20
In this thesis I present a new technique for the spatially resolved investigation of the magnetic properties of bulk samples. Standard one dimensional neutron depolarisation analysis is combined with neutron radiography to a method we call Neutron Depolarisation Imaging (NDI). The experimental setup which was installed at the neutron radiography beam line ANTARES at FRM II consists of a double crystal monochromator, neutron polariser, spin flipper, polarisation analyser and a position sensitive CCD detector. A comprehensive discussion of the requirements for these components is given and the limitations of the method are shown. The maximum spatial resolution which can be achieved with a neutron radiography setup is determined by the collimation of the neutron beam and the distance between sample and detector. Different types of polarisers have been tested and their advantages and disadvantages are discussed. A double crystal monochromator and a new type of polariser employing polarising neutron supermirrors based on the principle of an optical periscope were developed and tested during this work. Furthermore, NDI measurements on various samples of the weakly ferromagnetic materials Pd{sub 1-x}Ni{sub x} and Ni{sub 3}Al are presented. Neutron depolarisation radiography and tomography measurements were conducted with a spatial resolution as high as 0.3 mm on Pd{sub 1-x}Ni{sub x} and Ni{sub 3}Al samples. The feasibility of NDI experiments under hydrostatic pressures up to 10 kbar was shown on a sample of Ni{sub 3}Al using a modified Cu:Be clamp cell. A decrease of the ordering temperature by 2 K under hydrostatic pressure was determined from the NDI measurements and shows the potential of the method for further high pressure experiments. Additionally a method was developed which in principle allows to obtain the intrinsic dependence of the ordering temperature T{sub C} on the ordered moment Ms from NDI measurements on inhomogeneous samples containing regions with
Design of a boron neutron capture enhanced fast neutron therapy assembly
Energy Technology Data Exchange (ETDEWEB)
Wang, Zhonglu [Georgia Inst. of Technology, Atlanta, GA (United States)
2006-12-01
The use of boron neutron capture to boost tumor dose in fast neutron therapy has been investigated at several fast neutron therapy centers worldwide. This treatment is termed boron neutron capture enhanced fast neutron therapy (BNCEFNT). It is a combination of boron neutron capture therapy (BNCT) and fast neutron therapy (FNT). It is believed that BNCEFNT may be useful in the treatment of some radioresistant brain tumors, such as glioblastoma multiform (GBM). A boron neutron capture enhanced fast neutron therapy assembly has been designed for the Fermilab Neutron Therapy Facility (NTF). This assembly uses a tungsten filter and collimator near the patient's head, with a graphite reflector surrounding the head to significantly increase the dose due to boron neutron capture reactions. The assembly was designed using Monte Carlo radiation transport code MCNP version 5 for a standard 20x20 cm^{2} treatment beam. The calculated boron dose enhancement at 5.7-cm depth in a water-filled head phantom in the assembly with a 5x5 cm^{2} collimation was 21.9% per 100-ppm ^{10}B for a 5.0-cm tungsten filter and 29.8% for a 8.5-cm tungsten filter. The corresponding dose rate for the 5.0-cm and 8.5-cm thick filters were 0.221 and 0.127 Gy/min, respectively; about 48.5% and 27.9% of the dose rate of the standard 10x10 cm^{2} fast neutron treatment beam. To validate the design calculations, a simplified BNCEFNT assembly was built using four lead bricks to form a 5x5 cm^{2} collimator. Five 1.0-cm thick 20x20 cm^{2} tungsten plates were used to obtain different filter thicknesses and graphite bricks/blocks were used to form a reflector. Measurements of the dose enhancement of the simplified assembly in a water-filled head phantom were performed using a pair of tissue-equivalent ion chambers. One of the ion chambers is loaded with 1000-ppm natural boron (184-ppm ^{10}B) to measure dose due to boron neutron capture. The
Neutronic predesign tool for fusion power reactors system assessment
Energy Technology Data Exchange (ETDEWEB)
Jaboulay, J.-C., E-mail: jean-charles.jaboulay@cea.fr [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Li Puma, A. [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Martínez Arroyo, J. [ETSEIB, Internship in CEA (Spain)
2013-10-15
SYCOMORE, a fusion reactor system code based on a modular approach, is under development at CEA. In this framework, this paper describes a methodology developed to build the neutronic module of SYCOMORE. This neutronic module aims to evaluate main neutronic parameters characterising a fusion reactor (tokamak): tritium breeding ratio, multiplication factor, nuclear heating as a function of the reactor main geometrical parameters (major radius, elongation, etc.), of the radial build, Li enrichment, blanket and shield thickness, etc. It is based on calculations carried out with APOLLO2 and TRIPOLI-4 CEA transport code on simplified 1D and 2D neutronic models. These models are validated versus a more detailed 3D Monte-Carlo model (using TRIPOLI-4). To ease the integration of this neutronic module in SYCOMORE and provide results instantly, a surrogate model that replicates the 1D and 2D neutronic model results was used. Among the different surrogate models types (polynomial interpolation, responses functions, interpolating by Kriging, artificial neural network, etc.) the neural networks were selected for their efficiency and flexibility. The methodology described in this paper to build SYCOMORE neutronic module is devoted to HCLL blanket, but it could be applied to any breeder blanket concept provided that appropriate validation could be carried out.
A collimated neutron detector for RFP plasmas in MST
Capecchi, W. J.; Anderson, J. K.; Bonofiglo, P. J.; Kim, J.; Sears, S.
2016-11-01
The neutron emissivity profile in the Madison Symmetric Torus is being reconstructed through the use of a collimated neutron detector. A scintillator-photomultiplier tube (PMT) system is employed to detect the fusion neutrons with the plasma viewing volume defined by a 55 cm deep, 5 cm diameter aperture. Effective detection of neutrons from the viewing volume is achieved through neutron moderation using 1300 lbs of high density polyethylene shielding, which modeling predicts attenuates the penetrating flux by a factor of 104 or more. A broad spectrum of gamma radiation is also present due to the unconfined fusion proton bombardment of the thick aluminum vacuum vessel. A 15 cm cylindrical liquid scintillator of 3.8 cm diameter is used to further increase directional sensitivity. A fast (5 ns rise time) preamplifier and digitization at 500 MHz prevent pulse pile-up even at high count rates (˜104/s). The entire neutron camera system is situated on an adjustable inclining base which provides the differing plasma viewing volumes necessary for reconstruction of the neutron emissivity profile. This profile, directly related to the fast-ion population, allows for an investigation of the critical fast-ion pressure gradient required to destabilize a neutral beam driven Alfvénic mode which has been shown to transport fast ions.
Production of Medical isotope Technecium-99 from DT Fusion neutrons
Boguski, John; Gentile, Charles; Ascione, George
2011-10-01
High energy neutrons produced in DT fusion reactors have a secondary application for use in the synthesis of valuable man-made isotopes utilized in industry today. One such isotope is metastable Technecium-99 (Tc99m), a low energy gamma emitter used in ~ 85% of all medical imaging diagnostics. Tc99m is created through beta decay of Molybdenum-99 (Mo99), which itself has only a 66 hour half-life and must be created from a neutron capture by the widely available and stable isotope Molydenum-98. Current worldwide production of Tc99m occurs in just five locations and relies on obtaining the fission byproduct Mo99 from highly enriched Uranium reactors. A Tc99m generator using DT fusion neutrons, however, could potentially be operated at individual hospitals and medical facilities without the use of any fissile material. The neutron interaction of the DT neutrons with Molybdenum in a potential device geometry was modeled using Monte Carlo neutron transport code MCNP. Trial experiments were also performed to test the viability of using DT neutrons to create ample quantities of Tc99m. Modeling and test results will follow.
Neutron-Induced Failures in Semiconductor Devices
Energy Technology Data Exchange (ETDEWEB)
Wender, Stephen Arthur [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2016-04-06
This slide presentation explores single event effect, environmental neutron flux, system response, the Los Alamos Neutron Science Center (LANSCE) neutron testing facility, examples of SEE measurements, and recent interest in thermal neutrons.
Neutron drip transition in accreting and nonaccreting neutron star crusts
Chamel, N; Zdunik, J L; Haensel, P
2015-01-01
The neutron-drip transition in the dense matter constituting the interior of neutron stars generally refers to the appearance of unbound neutrons as the matter density reaches some threshold density $\\rho_\\textrm{drip}$. This transition has been mainly studied under the cold catalyzed matter hypothesis. However, this assumption is unrealistic for accreting neutron stars. After examining the physical processes that are thought to be allowed in both accreting and nonaccreting neutron stars, suitable conditions for the onset of neutron drip are derived and general analytical expressions for the neutron drip density and pressure are obtained. Moreover, we show that the neutron-drip transition occurs at lower density and pressure than those predicted within the mean-nucleus approximation. This transition is studied numerically for various initial composition of the ashes from X-ray bursts and superbursts using microscopic nuclear mass models.
Tailoring phase-space in neutron beam extraction
Energy Technology Data Exchange (ETDEWEB)
Weichselbaumer, S. [Heinz Maier-Leibnitz Zentrum und Physik-Department E21, Technische Universität München, Lichtenbergstr. 1, D-85748 Garching (Germany); Brandl, G. [Heinz Maier-Leibnitz Zentrum und Physik-Department E21, Technische Universität München, Lichtenbergstr. 1, D-85748 Garching (Germany); Physik-Department E21, Technische Universität München, James-Franck-Str. 1, D-85748 Garching (Germany); Georgii, R., E-mail: Robert.Georgii@frm2.tum.de [Heinz Maier-Leibnitz Zentrum und Physik-Department E21, Technische Universität München, Lichtenbergstr. 1, D-85748 Garching (Germany); Physik-Department E21, Technische Universität München, James-Franck-Str. 1, D-85748 Garching (Germany); Stahn, J. [Laboratory for Neutron Scattering, Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland); Panzner, T. [Material Science and Simulations, Neutrons and Muons, Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland); Böni, P. [Physik-Department E21, Technische Universität München, James-Franck-Str. 1, D-85748 Garching (Germany)
2015-09-01
In view of the trend towards smaller samples and experiments under extreme conditions it is important to deliver small and homogeneous neutron beams to the sample area. For this purpose, elliptic and/or Montel mirrors are ideally suited as the phase space of the neutrons can be defined far away from the sample. Therefore, only the useful neutrons will arrive at the sample position leading to a very low background. We demonstrate the ease of designing neutron transport systems using simple numeric tools, which are verified using Monte-Carlo simulations that allow taking into account effects of gravity and finite beam size. It is shown that a significant part of the brilliance can be transferred from the moderator to the sample. Our results may have a serious impact on the design of instruments at spallation sources such as the European Spallation Source (ESS) in Lund, Sweden.
Neutron scattering at Australia's replacement research reactor
Energy Technology Data Exchange (ETDEWEB)
Kennedy, S.J.; Robinson, R.A.; Hunter, B.A. [Physics Division, ANSTO, Lucas Heights, NSW (Australia)
2001-03-01
On August 25{sup th} 1999, the Australian government gave final approval to build a research reactor to replace the existing HIFAR reactor at Lucas Heights. The replacement reactor, which will commence operation in 2005, will be multipurpose with capabilities for both neutron beam research and radioisotope production. Cold, and thermal neutron sources are to be installed and supermirror guides will transport cold and thermal neutron beams into a large modern guide hall. The reactor and associated infrastructure is to be built by INVAP, SE and subcontractors under contract. The neutron beam instruments will be developed by ANSTO in consultation with the Australian user community and interested overseas parties. We review the planned scientific capabilities, give a description of the facility and a status report on the activities so far. (author)
Current status of high energy nucleon-meson transport code
Energy Technology Data Exchange (ETDEWEB)
Takada, Hiroshi; Sasa, Toshinobu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1998-03-01
Current status of design code of accelerator (NMTC/JAERI code), outline of physical model and evaluation of accuracy of code were reported. To evaluate the nuclear performance of accelerator and strong spallation neutron origin, the nuclear reaction between high energy proton and target nuclide and behaviors of various produced particles are necessary. The nuclear design of spallation neutron system used a calculation code system connected the high energy nucleon{center_dot}meson transport code and the neutron{center_dot}photon transport code. NMTC/JAERI is described by the particle evaporation process under consideration of competition reaction of intranuclear cascade and fission process. Particle transport calculation was carried out for proton, neutron, {pi}- and {mu}-meson. To verify and improve accuracy of high energy nucleon-meson transport code, data of spallation and spallation neutron fragment by the integral experiment were collected. (S.Y.)
Energy Technology Data Exchange (ETDEWEB)
Baltazar R, A. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica, Programa de Doctorado en Ingenieria y Tecnologia Aplicada, 98068 Zacatecas, Zac. (Mexico); Medina C, D.; Soto B, T. G. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Programa de Doctorado en Ciencias Basicas, 98068 Zacatecas, Zac. (Mexico); Vega C, H. R., E-mail: raigosa.antonio@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico)
2016-10-15
Neutrons occur naturally, regardless of whether they are obtained as a by-product of other reactions or intentionally, mainly as a by-product of the interaction of cosmic rays with the nuclei of the atmosphere, and in anthropogenic or artificial form with neutron generators, nuclear reactors, radioisotope sources, etc. Due to their high radiobiological efficiency is important measure them in order to estimate the effective dose in occupationally exposed personnel and the public in general. This dose depends on the amount of neutrons and their energy; in order to reduce neutron energy, light materials based on H, D, C, Be are used which moderate and thermalize them. The objective of this work was to determine the behavior of monoenergetic sources of neutrons in their transport within polyethylene of different thicknesses. The study was carried out using Monte Carlo methods with the code MCNP5, where 23 monoenergetic sources of I E(-9) were used at 20 MeV by influencing the neutrons on various polyethylene surfaces whose thickness was varied from 5.08 to 30.48 cm and the total neutron flux was estimated, as well as its spectrum when crossing the various thicknesses used in the study. (Author)
Shin, Yunchang; Lavelle, Christopher M; Baxter, David V; Tong, Xin; Yan, Haiyang; Leuschner, Mark
2007-01-01
Neutron energy spectrum measurements from a solid methane moderator were performed at the Low Energy Neutron Source (LENS) at Indiana University Cyclotron Facility (IUCF) to verify our neutron scattering model of solid methane. The time-of-flight method was used to measure the energy spectrum of the moderator in the energy range of 0.1$meV\\sim$ 1$eV$. Neutrons were counted with a high efficiency $^{3}{He}$ detector. The solid methane moderator was operated in phase II temperature and the energy spectra were measured at the temperatures of 20K and 4K. We have also tested our newly-developed scattering kernels for phase II solid methane by calculating the neutron spectral intensity expected from the methane moderator at the LENS neutron source using MCNP (Monte Carlo N-particle Transport Code). Within the expected accuracy of our approximate approach, our model predicts both the neutron spectral intensity and the optimal thickness of the moderator at both temperatures. The predictions are compared to the measur...
Hybrid superconducting neutron detectors
Energy Technology Data Exchange (ETDEWEB)
Merlo, V.; Lucci, M.; Ottaviani, I. [Dipartimento di Fisica, Università Tor Vergata, Via della Ricerca Scientifica, I-00133 Roma (Italy); Salvato, M.; Cirillo, M. [Dipartimento di Fisica, Università Tor Vergata, Via della Ricerca Scientifica, I-00133 Roma (Italy); CNR SPIN Salerno, Università di Salerno, Via Giovanni Paolo II, n.132, 84084 Fisciano (Italy); Scherillo, A. [Science and Technology Facility Council, ISIS Facility Chilton, Didcot, Oxfordshire OX11 0QX (United Kingdom); Celentano, G. [ENEA Frascati Research Centre, Via. E. Fermi 45, 00044 Frascati (Italy); Pietropaolo, A., E-mail: antonino.pietropaolo@enea.it [ENEA Frascati Research Centre, Via. E. Fermi 45, 00044 Frascati (Italy); Mediterranean Institute of Fundamental Physics, Via Appia Nuova 31, 00040 Marino, Roma (Italy)
2015-03-16
A neutron detection concept is presented that is based on superconductive niobium (Nb) strips coated by a boron (B) layer. The working principle of the detector relies on the nuclear reaction, {sup 10}B + n → α + {sup 7}Li, with α and Li ions generating a hot spot on the current-biased Nb strip which in turn induces a superconducting-normal state transition. The latter is recognized as a voltage signal which is the evidence of the incident neutron. The above described detection principle has been experimentally assessed and verified by irradiating the samples with a pulsed neutron beam at the ISIS spallation neutron source (UK). It is found that the boron coated superconducting strips, kept at a temperature T = 8 K and current-biased below the critical current I{sub c}, are driven into the normal state upon thermal neutron irradiation. As a result of the transition, voltage pulses in excess of 40 mV are measured while the bias current can be properly modulated to bring the strip back to the superconducting state, thus resetting the detector. Measurements on the counting rate of the device are presented and the basic physical features of the detector are discussed.
Hybrid Superconducting Neutron Detectors
Merlo, V; Cirillo, M; Lucci, M; Ottaviani, I; Scherillo, A; Celentano, G; Pietropaolo, A
2014-01-01
A new neutron detection concept is presented that is based on superconductive niobium (Nb) strips coated by a boron (B) layer. The working principle of the detector relies on the nuclear reaction 10B+n $\\rightarrow$ $\\alpha$+ 7Li , with $\\alpha$ and Li ions generating a hot spot on the current-biased Nb strip which in turn induces a superconducting-normal state transition. The latter is recognized as a voltage signal which is the evidence of the incident neutron. The above described detection principle has been experimentally assessed and verified by irradiating the samples with a pulsed neutron beam at the ISIS spallation neutron source (UK). It is found that the boron coated superconducting strips, kept at a temperature T = 8 K and current-biased below the critical current Ic, are driven into the normal state upon thermal neutron irradiation. As a result of the transition, voltage pulses in excess of 40 mV are measured while the bias current can be properly modulated to bring the strip back to the supercond...
Energy Technology Data Exchange (ETDEWEB)
Van Esch, Patrick; Crisanti, Marta; Mutti, Paolo [Institut Laue Langevin, Grenoble (France)
2015-07-01
A research project is presented in which we aim at counting individual neutrons with CCD-like cameras. We explore theoretically a technique that allows us to use imaging detectors as counting detectors at lower counting rates, and transits smoothly to continuous imaging at higher counting rates. As such, the hope is to combine the good background rejection properties of standard neutron counting detectors with the absence of dead time of integrating neutron imaging cameras as well as their very good spatial resolution. Compared to Xray detection, the essence of thermal neutron detection is the nuclear conversion reaction. The released energies involved are of the order of a few MeV, while X-ray detection releases energies of the order of the photon energy, which is in the 10 KeV range. Thanks to advances in camera technology which have resulted in increased quantum efficiency, lower noise, as well as increased frame rate up to 100 fps for CMOS-type cameras, this more than 100-fold higher available detection energy implies that the individual neutron detection light signal can be significantly above the noise level, as such allowing for discrimination and individual counting, which is hard to achieve with X-rays. The time scale of CMOS-type cameras doesn't allow one to consider time-of-flight measurements, but kinetic experiments in the 10 ms range are possible. The theory is next confronted to the first experimental results. (authors)
Nesvizhevsky, Valery V.; Voronin, Alexei Yu.; Cubitt, Robert; Protasov, Konstantin V.
2010-02-01
The `whispering gallery' effect has been known since ancient times for sound waves in air, later in water and more recently for a broad range of electromagnetic waves: radio, optics, Roentgen and so on. It consists of wave localization near a curved reflecting surface and is expected for waves of various natures, for instance, for atoms and neutrons. For matter waves, it would include a new feature: a massive particle would be settled in quantum states, with parameters depending on its mass. Here, we present for the first time the quantum whispering-gallery effect for cold neutrons. This phenomenon provides an example of an exactly solvable problem analogous to the `quantum bouncer'; it is complementary to the recently discovered gravitationally bound quantum states of neutrons . These two phenomena provide a direct demonstration of the weak equivalence principle for a massive particle in a pure quantum state. Deeply bound whispering-gallery states are long-living and weakly sensitive to surface potential; highly excited states are short-living and very sensitive to the wall potential shape. Therefore, they are a promising tool for studying fundamental neutron-matter interactions, quantum neutron optics and surface physics effects.
McCARD for Neutronics Design and Analysis of Research Reactor Cores
Shim, Hyung Jin; Park, Ho Jin; Kwon, Soonwoo; Seo, Geon Ho; Hyo Kim, Chang
2014-06-01
McCARD is a Monte Carlo (MC) neutron-photon transport simulation code developed exclusively for the neutronics design and analysis of nuclear reactor cores. McCARD is equipped with the hierarchical modeling and scripting functions, the CAD-based geometry processing module, the adjoint-weighted kinetics parameter and source multiplication factor estimation modules as well as the burnup analysis capability for the neutronics design and analysis of both research and power reactor cores. This paper highlights applicability of McCARD for the research reactor core neutronics analysis, as demonstrated for Kyoto University Critical Assembly, HANARO, and YALINA.
Virtual neutron scattering experiments
DEFF Research Database (Denmark)
Overgaard, Julie Hougaard; Bruun, Jesper; May, Michael
2017-01-01
We describe how virtual experiments can be utilized in a learning design that prepares students for hands-on experiments at large-scale facilities. We illustrate the design by showing how virtual experiments are used at the Niels Bohr Institute in a master level course on neutron scattering....... In the last week of the course, students travel to a large-scale neutron scattering facility to perform real neutron scattering experiments. Through student interviews and survey answers, we argue, that the virtual training prepares the students to engage more fruitfully with experiments by letting them focus...... on physics and data rather than the overwhelming instrumentation. We argue that this is because they can transfer their virtual experimental experience to the real-life situation. However, we also find that learning is still situated in the sense that only knowledge of particular experiments is transferred...
Virtual neutron scattering experiments
DEFF Research Database (Denmark)
Overgaard, Julie Hougaard; Bruun, Jesper; May, Michael
2016-01-01
We describe how virtual experiments can be utilized in a learning design that prepares students for hands-on experiments at large-scale facilities. We illustrate the design by showing how virtual experiments are used at the Niels Bohr Institute in a master level course on neutron scattering....... In the last week of the course, students travel to a large-scale neutron scattering facility to perform real neutron scattering experiments. Through student interviews and survey answers, we argue, that the virtual training prepares the students to engage more fruitfully with experiments by letting them focus...... on physics and data rather than the overwhelming instrumentation. We argue that this is because they can transfer their virtual experimental experience to the real-life situation. However, we also find that learning is still situated in the sense that only knowledge of particular experiments is transferred...
Direction sensitive neutron detector
Energy Technology Data Exchange (ETDEWEB)
Ahlen, Steven; Fisher, Peter; Dujmic, Denis; Wellenstein, Hermann F.; Inglis, Andrew
2017-01-31
A neutron detector includes a pressure vessel, an electrically conductive field cage assembly within the pressure vessel and an imaging subsystem. A pressurized gas mixture of CF.sub.4, .sup.3He and .sup.4He at respective partial pressures is used. The field cage establishes a relatively large drift region of low field strength, in which ionization electrons generated by neutron-He interactions are directed toward a substantially smaller amplification region of substantially higher field strength in which the ionization electrons undergo avalanche multiplication resulting in scintillation of the CF.sub.4 along scintillation tracks. The imaging system generates two-dimensional images of the scintillation patterns and employs track-finding to identify tracks and deduce the rate and direction of incident neutrons. One or more photo-multiplier tubes record the time-profile of the scintillation tracks permitting the determination of the third coordinate.
Energy Technology Data Exchange (ETDEWEB)
Hayashi, Masayuki
1998-12-04
The neutron absorbing alloy of the present invention comprises Ti or an alloy thereof as a mother material, to which from 2 to 40% by weight of Hf and Gd within a range of from 4 to 50% by weight in total are added respectively. Ti is excellent in specific strength, corrosion resistance and workability, and produces no noxious intermetallic compound with Hf and Gd. In addition, since the alloy can incorporate a great quantity of Hf and Gd, a neutron absorbing material having excellent neutron absorbing performance than usual and excellent in specific strength, corrosion resistance and workability can be manufactured conveniently and economically not by a special manufacturing method. (T.M.)
Carbon neutron star atmospheres
Suleimanov, V F; Pavlov, G G; Werner, K
2013-01-01
The accuracy of measuring the basic parameters of neutron stars is limited in particular by uncertainties in chemical composition of their atmospheres. For example, atmospheres of thermally - emitting neutron stars in supernova remnants might have exotic chemical compositions, and for one of them, the neutron star in CasA, a pure carbon atmosphere has recently been suggested by Ho & Heinke (2009). To test such a composition for other similar sources, a publicly available detailed grid of carbon model atmosphere spectra is needed. We have computed such a grid using the standard LTE approximation and assuming that the magnetic field does not exceed 10^8 G. The opacities and pressure ionization effects are calculated using the Opacity Project approach. We describe the properties of our models and investigate the impact of the adopted assumptions and approximations on the emergent spectra.
Uniformly rotating neutron stars
Boshkayev, Kuantay
2016-01-01
In this chapter we review the recent results on the equilibrium configurations of static and uniformly rotating neutron stars within the Hartle formalism. We start from the Einstein-Maxwell-Thomas-Fermi equations formulated and extended by Belvedere et al. (2012, 2014). We demonstrate how to conduct numerical integration of these equations for different central densities ${\\it \\rho}_c$ and angular velocities $\\Omega$ and compute the static $M^{stat}$ and rotating $M^{rot}$ masses, polar $R_p$ and equatorial $R_{\\rm eq}$ radii, eccentricity $\\epsilon$, moment of inertia $I$, angular momentum $J$, as well as the quadrupole moment $Q$ of the rotating configurations. In order to fulfill the stability criteria of rotating neutron stars we take into considerations the Keplerian mass-shedding limit and the axisymmetric secular instability. Furthermore, we construct the novel mass-radius relations, calculate the maximum mass and minimum rotation periods (maximum frequencies) of neutron stars. Eventually, we compare a...
Bonner sphere neutron spectrometry at spent fuel casks
Rimpler, A
2002-01-01
For transport and interim storage of spent fuel elements from power reactors and vitrified highly active waste (HAW) from reprocessing, various types of casks are used. The radiation exposure of the personnel during transportation and storage of these casks is caused by mixed photon-neutron fields and, frequently, the neutron dose is predominant. In operational radiation protection, survey meters and even personal dosemeters with imperfect energy dependence of the dose-equivalent response are used, i.e. the fluence response of the devices does not match the fluence-to-dose equivalent conversion function. In order to achieve more accurate dosimetric information and to investigate the performance of dosemeters, spectrometric investigations of the neutron fields are necessary. Therefore, fluence spectra and dose rates were measured by means of a simple portable Bonner multisphere spectrometer (BSS). The paper describes briefly the experimental set-up and evaluation procedure. Measured spectra for different locat...
Corrosion resistant neutron absorbing coatings
Choi, Jor-Shan [El Cerrito, CA; Farmer, Joseph C [Tracy, CA; Lee, Chuck K [Hayward, CA; Walker, Jeffrey [Gaithersburg, MD; Russell, Paige [Las Vegas, NV; Kirkwood, Jon [Saint Leonard, MD; Yang, Nancy [Lafayette, CA; Champagne, Victor [Oxford, PA
2012-05-29
A method of forming a corrosion resistant neutron absorbing coating comprising the steps of spray or deposition or sputtering or welding processing to form a composite material made of a spray or deposition or sputtering or welding material, and a neutron absorbing material. Also a corrosion resistant neutron absorbing coating comprising a composite material made of a spray or deposition or sputtering or welding material, and a neutron absorbing material.
Atmospheres around Neutron Stars
Fryer, Chris L.; Benz, Willy
1994-12-01
Interest in the behavior of atmospheres around neutron stars has grown astronomically in the past few years. Some of this interest arrived in the wake of the explosion of Supernova 1987A and its elusive remnant; spawning renewed interest in a method to insure material ``fall-back'' onto the adolescent neutron star in an effort to transform it into a silent black hole. However, the bulk of the activity with atmospheres around neutron stars is concentrated in stellar models with neutron star, rather than white dwarf, cores; otherwise known as Thorne-Zytkow objects. First a mere seed in the imagination of theorists, Thorne-Zytkow objects have grown into an observational reality with an ever-increasing list of formation scenarios and observational prospects. Unfortunately, the analytic work of Chevalier on supernova fall-back implies that, except for a few cases, the stellar simulations of Thorne-Zytkow objects are missing an important aspect of physics: neutrinos. Neutrino cooling removes the pressure support of these atmospheres, allowing accretion beyond the canonical Eddington rate for these objects. We present here the results of detailed hydrodynamical simulations in one and two dimensions with the additional physical effects of neutrinos, advanced equations of state, and relativity over a range of parameters for our atmosphere including entropy and chemical composition as well as a range in the neutron star size. In agreement with Chevalier, we find, under the current list of formation scenarios, that the creature envisioned by Thorne and Zytkow will not survive the enormous appetite of a neutron star. However, neutrino heating (a physical effect not considered in Chevalier's analysis) can play an important role in creating instabilities in some formation schemes, leading to an expulsion of matter rather than rapid accretion. By placing scrutiny upon the formation methods, we can determine the observational prospects for each.
Pixelated neutron image plates
Schlapp, M.; Conrad, H.; von Seggern, H.
2004-09-01
Neutron image plates (NIPs) have found widespread application as neutron detectors for single-crystal and powder diffraction, small-angle scattering and tomography. After neutron exposure, the image plate can be read out by scanning with a laser. Commercially available NIPs consist of a powder mixture of BaFBr : Eu2+ and Gd2O3 dispersed in a polymer matrix and supported by a flexible polymer sheet. Since BaFBr : Eu2+ is an excellent x-ray storage phosphor, these NIPs are particularly sensitive to ggr-radiation, which is always present as a background radiation in neutron experiments. In this work we present results on NIPs consisting of KCl : Eu2+ and LiF that were fabricated into ceramic image plates in which the alkali halides act as a self-supporting matrix without the necessity for using a polymeric binder. An advantage of this type of NIP is the significantly reduced ggr-sensitivity. However, the much lower neutron absorption cross section of LiF compared with Gd2O3 demands a thicker image plate for obtaining comparable neutron absorption. The greater thickness of the NIP inevitably leads to a loss in spatial resolution of the image plate. However, this reduction in resolution can be restricted by a novel image plate concept in which a ceramic structure with square cells (referred to as a 'honeycomb') is embedded in the NIP, resulting in a pixelated image plate. In such a NIP the read-out light is confined to the particular illuminated pixel, decoupling the spatial resolution from the optical properties of the image plate material and morphology. In this work, a comparison of experimentally determined and simulated spatial resolutions of pixelated and unstructured image plates for a fixed read-out laser intensity is presented, as well as simulations of the properties of these NIPs at higher laser powers.
Energy Technology Data Exchange (ETDEWEB)
COURANT,E.D.
1998-04-27
There does not appear to be any obvious way to accelerate neutrons, polarized or otherwise, to high energies by themselves. To investigate the behavior of polarized neutrons the authors therefore have to obtain them by accelerating them as components of heavier nuclei, and then sorting out the contribution of the neutrons in the analysis of the reactions produced by the heavy ion beams. The best neutron carriers for this purpose are probably {sup 3}He nuclei and deuterons. A polarized deuteron is primarily a combination of a proton and a neutron with their spins pointing in the same direction; in the {sup 3}He nucleus the spins of the two protons are opposite and the net spin (and magnetic moment) is almost the same as that of a free neutron. Polarized ions other than protons may be accelerated, stored and collided in a ring such as RHIC provided the techniques proposed for polarized proton operation can be adapted (or replaced by other strategies) for these ions. To accelerate polarized particles in a ring, one must make provisions for overcoming the depolarizing resonances that occur at certain energies. These resonances arise when the spin tune (ratio of spin precession frequency to orbit frequency) resonates with a component present in the horizontal field. The horizontal field oscillates with the vertical motion of the particles (due to vertical focusing); its frequency spectrum is dominated by the vertical oscillation frequency and its modulation by the periodic structure of the accelerator ring. In addition, the magnet imperfections that distort the closed orbit vertically contain all integral Fourier harmonics of the orbit frequency.
Lewis, Jason M.
Active neutron methods use a neutron source to interrogate fissionable material. In this work a 4He gas scintillation fast neutron detection system is used to measure neutrons created by the interrogation. Three new applications of this method are developed: spent nuclear fuel assay, fission rate measurement, and special nuclear material detection. Three active neutron methods are included in this thesis. First a non-destructive plutonium assay technique called Multispectral Active Neutron Interrogation Analysis is developed. It is based on interrogating fuel with neutrons at several different energies. The induced fission rates at each interrogation energy are compared with results from a neutron transport model of the irradiation geometry in a system of equations to iteratively solve the inverse problem for isotopic composition. The model is shown to converge on the correct composition for a material with 3 different fissionable components, a representative neutron absorber, and any neutron transparent material such as oxygen in a variety of geometries. Next an experimental fission rate measurement technique is developed using 4He gas scintillation fast neutron detector. Several unique features of this detector allow it to detect and provide energy information on fast neutrons with excellent gamma discrimination efficiency. The detector can measure induced fission rate by energetically differentiating between interrogation neutrons and higher energy fission neutrons. The detector response to a mono-energetic deuterium-deuterium fusion neutron generator and a 252Cf source are compared to examine the difference in detected energy range. Finally we demonstrate a special nuclear material detection technique by detecting an unambiguous fission neutron signal produced in natural uranium during active neutron interrogation using a deuterium-deuterium neutron generator and a high pressure 4He gas fast neutron scintillation detector. Energy histograms resulting from this
Krist, Th; Kennedy, S. J.; Hicks, T. J.; Mezei, F.
A new type of a neutron polarizing bender was developed in co-operation with BENSC and ANSTO. It is based upon bent thin silicon wafers coated on one side with SiFeCo polarizing supermirrors and on the other side with Gd. Initial tests at BENSC in a 300 Oe magnetic field yielded a transmission of spin-up neutrons of about 55% over an angle range of 0.75° and flipping ratios > 30. Subsequent tests at ANSTO at 1200 Oe yielded a transmission of 48% with a flipping ratio > 45.
Snell, A.H.
1957-12-01
This patent relates to a reactor and process for carrying out a controlled fast neutron chain reaction. A cubical reactive mass, weighing at least 920 metric tons, of uranium metal containing predominantly U/sup 238/ and having a U/sup 235/ content of at least 7.63% is assembled and the maximum neutron reproduction ratio is limited to not substantially over 1.01 by insertion and removal of a varying amount of boron, the reactive mass being substantially freed of moderator.
Helium 3 neutron precision polarimetry
Menard, Christopher
2009-10-01
Measuring neutron polarization to a high degree of precision is critical for the next generation of neutron decay correlation experiments. Polarized neutrons are also used in experiments to probe the hadronic weak interaction which contributes a small portion (˜10-7) of the force between nucleons. Using a beam of cold neutrons at Los Alamos Neutron Science Center (LANSCE), we polarized neutrons and measured their absolute polarization to ˜0.1%. Neutrons were polarized by passing them through a ^3He spin filter, relying on the maximally spin dependent 3He neutron absorption cross section. The neutron polarization can be determined by measuring the wavelength-dependent neutron transmission through the ^3He cell. An independent measurement of the neutron polarization was also obtained by passing the polarized beam through an RF spin flipper and a second polarized ^3He cell, used as an analyzer. To measure the efficiency of the spin flipper, the same measurements were made after reversing the ^3He polarization in the polarizer by using NMR techniques (adiabatic fast passage). We will show the consistency of these two measurements and the resulting precision of neutron polarimetry using these techniques.
Neutron storage time measurement for the neutron EDM experiment
Griffith, W. Clark; Ito, Takeyasu; Ramsey, John; Makela, Mark; Clayton, Steven; Hennings-Yeomans, Raul; Saidur Rahaman, M.; Currie, Scott; Womack, Todd; Sondheim, Walter; Cooper, Martin
2010-11-01
A new experiment to search for the neutron electric dipole moment (nEDM) is under development for installation at the Spallation Neutron Source (SNS) at Oakridge National Laboratory. The experiment will use ultra-cold neutrons (UCN) stored in superfluid helium, along with ^3He atoms acting as a neutron spin analyzer and comagnetometer. One crucial factor affecting the ultimate sensitivity of the experiment is the neutron storage time that can be obtained in the acrylic measurement cell. The acrylic cell walls will be coated with deuterated polystyrene (dPS), which is expected to give a wall loss factor of ˜room temperature and below 20 K.
Neutron recognition in LAND detector for large neutron multiplicity
Pawłowski, P; Leifels, Y; Trautmann, W; Adrich, P; Aumann, T; Bacri, C O; Barczyk, T; Bassini, R; Bianchin, S; Boiano, C; Boretzky, K; Boudard, A; Chbihi, A; Cibor, J; Czech, B; De Napoli, M; Ducret, J -E; Emling, H; Frankland, J D; Gorbinet, T; Hellström, M; Henzlova, D; Hlavac, S; Immè, J; Iori, I; Johansson, H; Kezzar, K; Kupny, S; Lafriakh, A; Fèvre, A Le; Gentil, E Le; Leray, S; Łukasik, J; Lühning, J; Lynch, W G; Lynen, U; Majka, Z; Mocko, M; Müller, W F J; Mykulyak, A; Orth, H; Otte, A N; Palit, R; Panebianco, S; Pullia, A; Raciti, G; Rapisarda, E; Rossi, D; Salsac, M -D; Sann, H; Schwarz, C; Simon, H; Sfienti, C; Sümmerer, K; Tsang, M B; Verde, G; Veselsky, M; Volant, C; Wallace, M; Weick, H; Wiechula, J; Wieloch, A; Zwiegliński, B
2012-01-01
The performance of the LAND neutron detector is studied. Using an event-mixing technique based on one-neutron data obtained in the S107 experiment at the GSI laboratory, we test the efficiency of various analytic tools used to determine the multiplicity and kinematic properties of detected neutrons. A new algorithm developed recently for recognizing neutron showers from spectator decays in the ALADIN experiment S254 is described in detail. Its performance is assessed in comparison with other methods. The properties of the observed neutron events are used to estimate the detection efficiency of LAND in this experiment.
Neutron beam imaging at neutron spectrometers at Dhruva
Desai, Shraddha S.; Rao, Mala N.
2012-06-01
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 106-107 n/cm2/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.
Verbeke, Jerome Maurice
This thesis consists of three main parts. The first part is related to boron neutron capture therapy (BNCT), the second part to boron neutron capture synovectomy (BNCS), and the third part to the neutron generator development. A monoenergetic neutron beam simulation study is carried out to determine the most suitable neutron energy for treatment of shallow and deep-seated brain tumors in the context of BNCT. Two figures-of-merit---the absorbed skin dose and the absorbed tumor dose at a given depth in the brain---are used to measure the neutron beam quality. Based on the results of this study, moderators, reflectors and delimiters are designed and optimized to moderate the high-energy neutrons from the fusion reactions D-D and D-T down to a suitable energy spectrum. Two different computational models have been used to study the dose distribution in the brain. A monoenergetic neutron beam simulation study is carried out to determine the optimal neutron energy for treatment of rheumatoid arthritis. Two figures-of-merit are used to measure the neutron beam quality, the ratio of the synovium absorbed dose to the skin absorbed dose, and the ratio of the synovium absorbed dose to the bone absorbed dose. It was found that thermal neutron beams are optimal for treatment. Computation of the dose distribution in the knee requires the simulation of particle transport from the neutron source to the knee phantom through the moderator. The third part describes the development of high-intensity D-D and D-T neutron generators. Thick target neutron yield computations have been performed to estimate the neutron yield of titanium and scandium targets. With an average deuteron beam current of 1 A and an energy of 120 keV, a neutron production of about 1014 n/s can be estimated for a tritiated target. In mixed deuteron/triton beam operation, a beam current of 2 A at 150 keV is required for the same neutron output. Despite this lower neutron production, this mode of operation is
Santarius, John; Navarro, Marcos; Michalak, Matthew; Fancher, Aaron; Kulcinski, Gerald; Bonomo, Richard
2016-10-01
A newly initiated research project will be described that investigates methods for detecting shielded special nuclear materials by combining multi-dimensional neutron sources, forward/adjoint calculations modeling neutron and gamma transport, and sparse data analysis of detector signals. The key tasks for this project are: (1) developing a radiation transport capability for use in optimizing adaptive-geometry, inertial-electrostatic confinement (IEC) neutron source/detector configurations for neutron pulses distributed in space and/or phased in time; (2) creating distributed-geometry, gas-target, IEC fusion neutron sources; (3) applying sparse data and noise reduction algorithms, such as principal component analysis (PCA) and wavelet transform analysis, to enhance detection fidelity; and (4) educating graduate and undergraduate students. Funded by DHS DNDO Project 2015-DN-077-ARI095.
Effect of Different Structural Materials on Neutronic Performance of a Hybrid Reactor
Übeyli, Mustafa; Tel, Eyyüp
2003-06-01
Selection of structural material for a fusion-fission (hybrid) reactor is very important by taking into account of neutronic performance of the blanket. Refractory metals and alloys have much higher operating temperatures and neutron wall load (NWL) capabilities than low activation materials (ferritic/martensitic steels, vanadium alloys and SiC/SiC composites) and austenitic stainless steels. In this study, effect of primary candidate refractory alloys, namely, W-5Re, T111, TZM and Nb-1Zr on neutronic performance of the hybrid reactor was investigated. Neutron transport calculations were conducted with the help of SCALE 4.3 System by solving the Boltzmann transport equation with code XSDRNPM. Among the investigated structural materials, tantalum had the worst performance due to the fact that it has higher neutron absorption cross section than others. And W-5Re and TZM having similar results showed the best performance.
Application of neutron activation tracer sediment technique on environmental science
Institute of Scientific and Technical Information of China (English)
YinYi; ZhongWei－Ni; 等
1997-01-01
Field and laboratory inverstigations were carried out to study the transport and dispersion law of polluted sediments near wastewater outlet using neutron activation tracer technique.The direction of transport and dispersion of polluted sediments,dispersion amount in different directions,sedimentary region of polluted sediment and evaluation of polluted risk are given.This provided a new test method for the study of environmental science and added a new forecasted content for the evaluation of environmental influence.
Some Implications of Neutron Mirror Neutron Oscillation
Mohapatra, Rabindra N; Nussinov, S
2005-01-01
We comment on a recently discussed possibility of oscillations between neutrons and degenerate mirror neutrons in the context of mirror models for particles and forces. It has been noted by Bento and Berezhiani that if these oscillations occurred at a rate of $\\tau^{-1}_{NN'}\\sim sec^{-1}$, it would help explain putative super GKZ cosmic ray events provided the temperature of the mirror radiation is $\\sim 0.3-0.4$ times that of familiar cosmic microwave background radiation. We discuss how such oscillation time scales can be realized in mirror models and find that the simplest nonsupersymmetric model for this idea requires the existence of a low mass (30-3000 GeV) color triplet scalar or vector boson. A supersymmetric model, where this constraint can be avoided is severely constrained by the requirement of maintaining a cooler mirror sector. We also find that the reheat temperature after inflation in generic models that give fast $n-n'$ oscillation be less than about 100 GeV in order to maintain the required ...
Measurements of the neutron activation cross sections for Bi and Co at 386 MeV.
Yashima, H; Sekimoto, S; Ninomiya, K; Kasamatsu, Y; Shima, T; Takahashi, N; Shinohara, A; Matsumura, H; Satoh, D; Iwamoto, Y; Hagiwara, M; Nishiizumi, K; Caffee, M W; Shibata, S
2014-10-01
Neutron activation cross sections for Bi and Co at 386 MeV were measured by activation method. A quasi-monoenergetic neutron beam was produced using the (7)Li(p,n) reaction. The energy spectrum of these neutrons has a high-energy peak (386 MeV) and a low-energy tail. Two neutron beams, 0° and 25° from the proton beam axis, were used for sample irradiation, enabling a correction for the contribution of the low-energy neutrons. The neutron-induced activation cross sections were estimated by subtracting the reaction rates of irradiated samples for 25° irradiation from those of 0° irradiation. The measured cross sections were compared with the findings of other studies, evaluated in relation to nuclear data files and the calculated data by Particle and Heavy Ion Transport code System code.
Neutron Scattering Investigations of Correlated Electron Systems and Neutron Instrumentation
DEFF Research Database (Denmark)
Holm, Sonja Lindahl
are a unique probe for studying the atomic and molecular structure and dynamics of materials. Even though neutrons are very expensive to produce, the advantages neutrons provide overshadow the price. As neutrons interact weakly with materials compared to many other probes, e.g. electrons or photons......, it is possible to make a neutron scattering experiment through sample environment equipment like cryostats or pressure cells. Another advantage of neutron experiments is that the wavelength and energy of the neutron match the inter-atomic distances and basic excitations of solid materials. The scattering cross...... is not taken into account in previous reports on the field effect of magnetic scattering, since usually only L 0 is probed. A paper draft submitted for publication describing the results of elastic and inelastic neutron scattering experiments performed on the oxygen-doped La2CuO4+y HTSC is appended (Tc 40 K...
Neutronic studies of the coupled moderators for spallation neutron sources
Institute of Scientific and Technical Information of China (English)
Yin Wen; Liang Jiu-Qing
2005-01-01
We investigate the neutronic performance of coupled moderators to be implemented in spallation neutron sources by Monte-Carlo simulation and give the slow neutron spectra for the cold and thermal moderators. CH4 moderator can provide slow neutrons with highly desirable characteristics and will be used in low-power spallation neutron soureces. The slow neutron intensity extracted from different angles has been calculated. The capability of moderation of liquid H2 is lower than H2O and liquid CH4 due to lower atomic number density of hydrogen but we can compensate for this disadvantage by using a premoderator. The H2O premoderator of 2cm thickness can reduce the heat deposition in the cold moderator by about 33% without spoiling the neutron pulse.
General Electric PETtrace cyclotron as a neutron source for boron neutron capture therapy
Bosko, Andrey
This research investigates the use of a PETtrace cyclotron produced by General Electric (GE) as a neutron source for boron neutron capture therapy (BNCT). The GE PETtrace was chosen for this investigation because this type of cyclotron is popular among nuclear pharmacies and clinics in many countries; it is compact and reliable; it produces protons with energies high enough to produce neutrons with appropriate energy and fluence rate for BNCT and it does not require significant changes in design to provide neutrons. In particular, the standard PETtrace 18O target is considered. The cyclotron efficiency may be significantly increased if unused neutrons produced during radioisotopes production could be utilized for other medical modalities such as BNCT at the same time. The resulting dose from the radiation emitted from the target is evaluated using the Monte Carlo radiation transport code MCNP at several depths in a brain phantom for different scattering geometries. Four different moderating materials of various thicknesses were considered: light water, carbon, heavy water, arid Fluental(TM). The fluence rate tally was used to calculate photon and neutron dose, by applying fluence rate-to-dose conversion factors. Fifteen different geometries were considered and a 30-cm thick heavy water moderator was chosen as the most suitable for BNCT with the GE PETtrace cyclotron. According to the Brookhaven Medical Research Reactor (BMRR) protocol, the maximum dose to the normal brain is set to 12.5 RBEGy, which for the conditions of using a heavy water moderator, assuming a 60 muA beam current, would be reached with a treatment time of 258 min. Results showed that using a PETtrace cyclotron in this configuration provides a therapeutic ratio of about 2.4 for depths up to 4 cm inside a brain phantom. Further increase of beam current proposed by GE should significantly improve the beam quality or the treatment time and allow treating tumors at greater depths.
A review on neutron reflectometry
Energy Technology Data Exchange (ETDEWEB)
Lee, Jeong Soo; Lee, Chang Hee; Shim, Hae Seop; Seong, Baek Seok
1999-03-01
This report contains principle and characteristic of neutron reflectometry. Therefore, in case of operating neutron reflectometer at HANARO in future, it will be a reference to the user who wishes to use the instrument effectively. Also, the current situation of neutron reflectometer operating in the world was examined. The detail of neutron reflectometer such as GANS(MURR), ADAM(ILL), POSY II(ANL), ROG(IRI) was described. The recent research situation on neutron reflectometry was also examined and it helps us to determine research field. (author)
Neutron proton crystallography station (PCS)
Energy Technology Data Exchange (ETDEWEB)
Fisher, Zoe [Los Alamos National Laboratory; Kovalevsky, Andrey [Los Alamos National Laboratory; Johnson, Hannah [Los Alamos National Laboratory; Mustyakimov, Marat [Los Alamos National Laboratory
2009-01-01
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.
Energy Technology Data Exchange (ETDEWEB)
Trotter, D.E. Gonzalez [Department of Physics, Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States)], E-mail: crowell@tunl.duke.edu; Meneses, F. Salinas [Department of Physics, Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States); Tornow, W. [Department of Physics, Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States)], E-mail: tornow@tunl.duke.edu; Crowell, A.S.; Howell, C.R. [Department of Physics, Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States); Schmidt, D. [Physikalisch-Technische Bundesanstalt, D-38116, Braunschweig (Germany); Walter, R.L. [Department of Physics, Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States)
2009-02-11
The methods employed and the results obtained from measurements and calculations of the detection efficiency for the neutron detectors used at Triangle Universities Nuclear Laboratory (TUNL) in the simultaneous determination of the {sup 1}S{sub 0} neutron-neutron and neutron-proton scattering lengths a{sub nn} and a{sub np}, respectively, are described. Typical values for the detector efficiency were 0.3. Very good agreement between the different experimental methods and between data and calculation has been obtained in the neutron energy range below E{sub n}=13MeV.
Directory of Open Access Journals (Sweden)
Grigorian H.
2010-10-01
Full Text Available We introduce the theoretical basis for modeling the cooling evolution of compact stars starting from Boltzmann equations in curved space-time. We open a discussion on observational veriﬁcation of diﬀerent neutron star models by consistent statistics. Particular interest has the question of existence of quark matter deep inside of compact object, which has to have a speciﬁc inﬂuence on the cooling history of the star. Besides of consideration of several constraints and features of cooling evolution, which are susceptible of being critical for internal structure of hot compact stars we have introduced a method of extraction of the mass distribution of the neutron stars from temperature and age data. The resulting mass distribution has been compared with the one suggested by supernove simulations. This method can be considered as an additional checking tool for the consistency of theoretical modeling of neutron stars. We conclude that the cooling data allowed existence of neutron stars with quark cores even with one-ﬂavor quark matter.
2001-01-01
CERN has been operating an Individual Dosimetry Service for neutrons for about 35 years. The service was based on nuclear emulsions in the form of film packages which were developed and scanned in the Service. In 1999, the supplier of theses packages informed CERN that they will discontinue production of this material. TIS-RP decided to look for an external service provider for individual neutron dosimetry. After an extensive market survey and an invitation for tender, a supplier that met the stringent technical requirements set up by CERN's host states for personal dosimeters was identified. The new dosimeter is based on a track-etching technique. Neutrons have the capability of damaging plastic material. The microscopic damage centres are revealed by etching them in a strong acid. The resulting etch pits can be automatically counted and their density is proportional to dose equivalent from neutrons. On the technical side, the new dosimeter provides an improved independence of its response from energy and th...
Neutron protein crystallography
Energy Technology Data Exchange (ETDEWEB)
Niimura, Nobuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1998-10-01
X-ray diffraction of single crystal has enriched the knowledge of various biological molecules such as proteins, DNA, t-RNA, viruses, etc. It is difficult to make structural analysis of hydrogen atoms in a protein using X-ray crystallography, whereas neutron diffraction seems usable to directly determine the location of those hydrogen atoms. Here, neutron diffraction method was applied to structural analysis of hen egg-white lysozyme. Since the crystal size of a protein to analyze is generally small (5 mm{sup 3} at most), the neutron beam at the sample position in monochromator system was set to less than 5 x 5 mm{sup 2} and beam divergence to 0.4 degree or less. Neutron imaging plate with {sup 6}Li or Gd mixed with photostimulated luminescence material was used and about 2500 Bragg reflections were recorded in one crystal setting. A total of 38278 reflections for 2.0 A resolution were collected in less than 10 days. Thus, stereo views of Trp-111 omit map around the indol ring of Trp-111 was presented and the three-dimensional arrangement of 696H and 264D atoms in the lysozyme molecules was determined using the omit map. (M.N.)
Neutron capture reactions at DANCE
Bredeweg, T. A.
2008-05-01
The Detector for Advanced Neutron Capture Experiments (DANCE) is a 4π BaF2 array consisting of 160 active detector elements. The primary purpose of the array is to perform neutron capture cross section measurements on small (>~100 μg) and/or radioactive (DANCE we have performed neutron capture cross section measurements on a wide array of medium to heavy mass nuclides. Measurements to date include neutron capture cross sections on 241,243Am, neutron capture and neutron-induced fission cross sections and capture-to-fission ratio (α = σγ/σf) for 235U using a new fission-tagging detector as well as neutron capture cross sections for several astrophysics branch-point nuclei. Results from several of these measurements will be presented along with a discussion of additional physics information that can be extracted from the DANCE data.
Nanostructure Neutron Converter Layer Development
Park, Cheol (Inventor); Sauti, Godfrey (Inventor); Kang, Jin Ho (Inventor); Lowther, Sharon E. (Inventor); Thibeault, Sheila A. (Inventor); Bryant, Robert G. (Inventor)
2016-01-01
Methods for making a neutron converter layer are provided. The various embodiment methods enable the formation of a single layer neutron converter material. The single layer neutron converter material formed according to the various embodiments may have a high neutron absorption cross section, tailored resistivity providing a good electric field penetration with submicron particles, and a high secondary electron emission coefficient. In an embodiment method a neutron converter layer may be formed by sequential supercritical fluid metallization of a porous nanostructure aerogel or polyimide film. In another embodiment method a neutron converter layer may be formed by simultaneous supercritical fluid metallization of a porous nanostructure aerogel or polyimide film. In a further embodiment method a neutron converter layer may be formed by in-situ metalized aerogel nanostructure development.
The neutrons for science facility at SPIRAL-2
Ledoux, X.; Aïche, M.; Avrigeanu, M.; Avrigeanu, V.; Balanzat, E.; Ban-d'Etat, B.; Ban, G.; Bauge, E.; Bélier, G.; Bém, P.; Borcea, C.; Caillaud, T.; Chatillon, A.; Czajkowski, S.; Dessagne, P.; Doré, D.; Fischer, U.; Frégeau, M. O.; Grinyer, J.; Guillous, S.; Gunsing, F.; Gustavsson, C.; Henning, G.; Jacquot, B.; Jansson, K.; Jurado, B.; Kerveno, M.; Klix, A.; Landoas, O.; Lecolley, F. R.; Lecouey, J. L.; Majerle, M.; Marie, N.; Materna, T.; Mrázek, J.; Negoita, F.; Novák, J.; Oberstedt, S.; Oberstedt, A.; Panebianco, S.; Perrot, L.; Plompen, A. J. M.; Pomp, S.; Prokofiev, A. V.; Ramillon, J. M.; Farget, F.; Ridikas, D.; Rossé, B.; Sérot, O.; Simakov, S. P.; Šimečková, E.; Štefánik, M.; Sublet, J. C.; Taïeb, J.; Tarrío, D.; Tassan-Got, L.; Thfoin, I.; Varignon, C.
2017-09-01
Numerous domains, in fundamental research as well as in applications, require the study of reactions induced by neutrons with energies from few MeV up to few tens of MeV. Reliable measurements also are necessary to improve the evaluated databases used by nuclear transport codes. This energy range covers a large number of topics like transmutation of nuclear waste, design of future fission and fusion reactors, nuclear medicine or test and development of new detectors. A new facility called Neutrons For Science (NFS) is being built for this purpose on the GANIL site at Caen (France). NFS is composed of a pulsed neutron beam for time-of-flight facility as well as irradiation stations for cross-section measurements. Neutrons will be produced by the interaction of deuteron and proton beams, delivered by the SPIRAL-2 linear accelerator, with thick or thin converters made of beryllium or lithium. Continuous and quasi-mono-energetic spectra will be available at NFS up to 40 MeV. In this fast energy region, the neutron flux is expected to be up to 2 orders of magnitude higher than at other existing time-of-flight facilities. In addition, irradiation stations for neutron-, proton- and deuteron-induced reactions will allow performing cross-section measurements by the activation technique. After a description of the facility and its characteristics, the experiments to be performed in the short and medium term will be presented.
Land-surface studies with a directional neutron detector.
Energy Technology Data Exchange (ETDEWEB)
Desilets, Darin (Sandia National Laboratories, Albuquerque, NM); Brennan, James S.; Mascarenhas, Nicholas; Marleau, Peter
2009-09-01
Direct measurements of cosmic-ray neutron intensity were recorded with a neutron scatter camera developed at SNL. The instrument used in this work is a prototype originally designed for nuclear non-proliferation work, but in this project it was used to characterize the response of ambient neutrons in the 0.5-10 MeV range to water located on or above the land surface. Ambient neutron intensity near the land surface responds strongly to the presence of water, suggesting the possibility of an indirect method for monitoring soil water content, snow water equivalent depth, or canopy intercepted water. For environmental measurements the major advantage of measuring neutrons with the scatter camera is the limited (60{sup o}) field of view that can be obtained, which allows observations to be conducted at a previously unattainable spatial scales. This work is intended to provide new measurements of directional fluxes which can be used in the design of new instruments for passively and noninvasively observing land-surface water. Through measurements and neutron transport modeling we have demonstrated that such a technique is feasible.
The neutrons for science facility at SPIRAL-2
Directory of Open Access Journals (Sweden)
Ledoux X.
2017-01-01
Full Text Available Numerous domains, in fundamental research as well as in applications, require the study of reactions induced by neutrons with energies from few MeV up to few tens of MeV. Reliable measurements also are necessary to improve the evaluated databases used by nuclear transport codes. This energy range covers a large number of topics like transmutation of nuclear waste, design of future fission and fusion reactors, nuclear medicine or test and development of new detectors. A new facility called Neutrons For Science (NFS is being built for this purpose on the GANIL site at Caen (France. NFS is composed of a pulsed neutron beam for time-of-flight facility as well as irradiation stations for cross-section measurements. Neutrons will be produced by the interaction of deuteron and proton beams, delivered by the SPIRAL-2 linear accelerator, with thick or thin converters made of beryllium or lithium. Continuous and quasi-mono-energetic spectra will be available at NFS up to 40 MeV. In this fast energy region, the neutron flux is expected to be up to 2 orders of magnitude higher than at other existing time-of-flight facilities. In addition, irradiation stations for neutron-, proton- and deuteron-induced reactions will allow performing cross-section measurements by the activation technique. After a description of the facility and its characteristics, the experiments to be performed in the short and medium term will be presented.
The muon-induced neutron indirect detection EXperiment, MINIDEX
Abt, I.; Caldwell, A.; Carissimo, C.; Gooch, C.; Kneißl, R.; Langford, J.; Liu, X.; Majorovits, B.; Palermo, M.; Schulz, O.; Vanhoefer, L.
2017-04-01
A new experiment to quantitatively measure neutrons induced by cosmic-ray muons in selected high-Z materials is introduced. The design of the Muon-Induced Neutron Indirect Detection EXperiment, MINIDEX, and the results from its first data taking period are presented as well as future plans. Neutron production in high-Z materials is of particular interest as such materials are used for shielding in low-background experiments. The design of next-generation large-scale experiments searching for neutrinoless double beta decay or direct interactions of dark matter requires reliable Monte Carlo simulations of background induced by muon interactions. The first five months of operation already provided a valuable data set on neutron production and neutron transport in lead. A first round of comparisons between MINIDEX data and Monte Carlo predictions obtained with a GEANT4-based package for two different sets of physics models of relevance for neutron production by muons is presented. The rate of muon-induced events is overall a factor three to four higher in data than predicted by the Monte Carlo packages. In addition, the time evolution of the muon-induced signal is not well described by the simulations.
Observation of Neutron Skyshine from an Accelerator Based Neutron Source
Energy Technology Data Exchange (ETDEWEB)
Franklyn, C. B. [Radiation Science Department, Necsa, PO Box 582, Pretoria 0001 (South Africa)
2011-12-13
A key feature of neutron based interrogation systems is the need for adequate provision of shielding around the facility. Accelerator facilities adapted for fast neutron generation are not necessarily suitably equipped to ensure complete containment of the vast quantity of neutrons generated, typically >10{sup 11} n{center_dot}s{sup -1}. Simulating the neutron leakage from a facility is not a simple exercise since the energy and directional distribution can only be approximated. Although adequate horizontal, planar shielding provision is made for a neutron generator facility, it is sometimes the case that vertical shielding is minimized, due to structural and economic constraints. It is further justified by assuming the atmosphere above a facility functions as an adequate radiation shield. It has become apparent that multiple neutron scattering within the atmosphere can result in a measurable dose of neutrons reaching ground level some distance from a facility, an effect commonly known as skyshine. This paper describes a neutron detection system developed to monitor neutrons detected several hundred metres from a neutron source due to the effect of skyshine.
Thermal neutron irradiation field design for boron neutron capture therapy of human explanted liver.
Bortolussi, S; Altieri, S
2007-12-01
The selective uptake of boron by tumors compared to that by healthy tissue makes boron neutron capture therapy (BNCT) an extremely advantageous technique for the treatment of tumors that affect a whole vital organ. An example is represented by colon adenocarcinoma metastases invading the liver, often resulting in a fatal outcome, even if surgical resection of the primary tumor is successful. BNCT can be performed by irradiating the explanted organ in a suitable neutron field. In the thermal column of the Triga Mark II reactor at Pavia University, a facility was created for this purpose and used for the irradiation of explanted human livers. The neutron field distribution inside the organ was studied both experimentally and by means of the Monte Carlo N-particle transport code (MCNP). The liver was modeled as a spherical segment in MCNP and a hepatic-equivalent solution was used as an experimental phantom. In the as-built facility, the ratio between maximum and minimum flux values inside the phantom ((phi(max)/phi(min)) was 3.8; this value can be lowered to 2.3 by rotating the liver during the irradiation. In this study, the authors proposed a new facility configuration to achieve a uniform thermal neutron flux distribution in the liver. They showed that a phi(max)/phi(min) ratio of 1.4 could be obtained without the need for organ rotation. Flux distributions and dose volume histograms were reported for different graphite configurations.
Albright, B. J.; Yin, L.; Favalli, A.
2016-10-01
Laser-ion-beam generation in the break-out afterburner (BOA) acceleration regime has been modeled for several deuteron-rich solid-density targets using the VPIC particle-in-cell code. Monte Carlo modeling of the transport of these beams in a beryllium converter in a pitcher-catcher neutron source configuration shows significant increases in neutron yields may be achievable through judicious choices of laser target material. Additionally, species-separation dynamics in some target materials during the BOA ion acceleration phase can be exploited to control the shapes of the neutron spectra. Work performed under the auspices of the U.S. DOE by the LANS, LLC, Los Alamos National Laboratory under Contract No. DE-AC52-06NA25396. Funding provided by the Los Alamos National Laboratory Directed Research and Development Program.
Neutron beam design for low intensity neutron and gamma-ray radioscopy using small neutron sources
Matsumoto, T
2003-01-01
Two small neutron sources of sup 2 sup 5 sup 2 Cf and sup 2 sup 4 sup 1 Am-Be radioisotopes were used for design of neutron beams applicable to low intensity neutron and gamma ray radioscopy (LINGR). In the design, Monte Carlo code (MCNP) was employed to generate neutron and gamma ray beams suited to LINGR. With a view to variable neutron spectrum and neutron intensity, various arrangements were first examined, and neutron-filter, gamma-ray shield and beam collimator were verified. Monte Carlo calculations indicated that with a suitable filter-shield-collimator arrangement, thermal neutron beam of 3,900 ncm sup - sup 2 s sup - sup 1 with neutron/gamma ratio of 7x10 sup 7 , and 25 ncm sup - sup 2 s sup - sup 1 with very large neutron/gamma ratio, respectively, could be produced by using sup 2 sup 5 sup 2 Cf(122 mu g) and a sup 2 sup 4 sup 1 Am-Be(37GBq)radioisotopes at the irradiation port of 35 cm from the neutron sources.
The spallation neutron source: New opportunities
Indian Academy of Sciences (India)
Ian S Anderson
2008-11-01
The spallation neutron source (SNS) facility became operational in the spring of 2006, and is now well on its way to become the world-leading facility for neutron scattering. Furthermore, the SNS and the HFIR reactor facility, newly outfitted with a brilliant cold source and guide hall, were brought together within a single Neutron Sciences Directorate at ORNL providing the opportunity to develop science and instrumentation programs which take advantage of the unique characteristics of each source. SNS and HFIR will both operate as scientific user facilities. Access to these facilities is being managed under an integrated proposal system, which also includes the Center for Nanophase Materials Sciences (CNMS) and the electron microscopes in the Shared Research Equipment (SHARE) program. Presently, SNS has three instruments operating in the user program and seven more will begin operations in 2008. When complete, the facility will accommodate 25 instruments enabling researchers from the United States and abroad to study materials science that forms the basis for new technologies in telecommunications, manufacturing, transportation, information technology, biotechnology, and health.
Neutron Scattering Investigations of Correlated Electron Systems and Neutron Instrumentation
DEFF Research Database (Denmark)
Holm, Sonja Lindahl
are a unique probe for studying the atomic and molecular structure and dynamics of materials. Even though neutrons are very expensive to produce, the advantages neutrons provide overshadow the price. As neutrons interact weakly with materials compared to many other probes, e.g. electrons or photons......, it is possible to make a neutron scattering experiment through sample environment equipment like cryostats or pressure cells. Another advantage of neutron experiments is that the wavelength and energy of the neutron match the inter-atomic distances and basic excitations of solid materials. The scattering cross...... magnetism. This is in contrast to what is observed as the critical temperature is slightly lower for this system compared to other co-doped systems, suggesting that the magnetic and superconducting phases co-exist. A published manuscript describes the study of magnetic and superconducting properties of Ba...
Neutron-gamma competition for $\\beta$-delayed neutron emission
Mumpower, Matthew; Moller, Peter
2016-01-01
We present a coupled Quasi-particle Random Phase Approximation and Hauser-Feshbach (QRPA+HF) model for calculating delayed particle emission. This approach uses microscopic nuclear structure information which starts with Gamow-Teller strength distributions in the daughter nucleus, and then follows the statistical decay until the initial available excitation energy is exhausted. Explicitly included at each particle emission stage is $\\gamma$-ray competition. We explore this model in the context of neutron emission of neutron-rich nuclei and find that neutron-gamma competition can lead to both increases and decreases in neutron emission probabilities, depending on the system considered. A second consequence of this formalism is a prediction of more neutrons on average being emitted after $\\beta$-decay for nuclei near the neutron dripline compared to models that do not consider the statistical decay.
Measurement of the neutron $\\beta$-asymmetry parameter $A_0$ with ultracold neutrons
Plaster, B; Back, H O; Bowles, T J; Broussard, L J; Carr, R; Clayton, S; Currie, S; Filippone, B W; Garcia, A; Geltenbort, P; Hickerson, K P; Hoagland, J; Hogan, G E; Hona, B; Holley, A T; Ito, T M; Liu, C -Y; Liu, J; Makela, M; Mammei, R R; Martin, J W; Melconian, D; Mendenhall, M P; Morris, C L; Mortensen, R; Pattie, R W; Jr.,; Galvan, A Perez; Pitt, M L; Ramsey, J C; Russell, R; Saunders, A; Schmid, R; Seestrom, S J; Sjue, S; Sondheim, W E; Tatar, E; Tipton, B; Vogelaar, R B; VornDick, B; Wrede, C; Xu, Y P; Yan, H; Young, A R; Yuan, J
2012-01-01
We present a detailed report of a measurement of the neutron $\\beta$-asymmetry parameter $A_0$, the parity-violating angular correlation between the neutron spin and the decay electron momentum, performed with polarized ultracold neutrons (UCN). UCN were extracted from a pulsed spallation solid deuterium source and polarized via transport through a 7-T magnetic field. The polarized UCN were then transported through an adiabatic-fast-passage spin-flipper field region, prior to storage in a cylindrical decay volume situated within a 1-T $2 \\times 2\\pi$ solenoidal spectrometer. The asymmetry was extracted from measurements of the decay electrons in multiwire proportional chamber and plastic scintillator detector packages located on both ends of the spectrometer. From an analysis of data acquired during runs in 2008 and 2009, we report $A_0 = -0.11966 \\pm 0.00089_{-0.00140} ^{+0.00123}$, from which we extract a value for the ratio of the weak axial-vector and vector coupling constants of the nucleon, $\\lambda = g...
CALTRANS: A parallel, deterministic, 3D neutronics code
Energy Technology Data Exchange (ETDEWEB)
Carson, L.; Ferguson, J.; Rogers, J.
1994-04-01
Our efforts to parallelize the deterministic solution of the neutron transport equation has culminated in a new neutronics code CALTRANS, which has full 3D capability. In this article, we describe the layout and algorithms of CALTRANS and present performance measurements of the code on a variety of platforms. Explicit implementation of the parallel algorithms of CALTRANS using both the function calls of the Parallel Virtual Machine software package (PVM 3.2) and the Meiko CS-2 tagged message passing library (based on the Intel NX/2 interface) are provided in appendices.
Deuterated polyethylene coatings for ultra-cold neutron applications
Energy Technology Data Exchange (ETDEWEB)
Brenner, Th.; Geltenbort, P. [Institut Laue-Langevin, 38042 Grenoble Cedex 9 (France); Fierlinger, P.; Gutsmiedl, E.; Hollering, A.; Petzoldt, G.; Ruhstorfer, D.; Stuiber, St.; Taubenheim, B.; Windmayer, D. [Physikdepartment, Technische Universität München, D-85748 Garching (Germany); Lauer, T.; Schroffenegger, J.; Zechlau, T. [Forschungsneutronenquelle Heinz Maier-Leibnitz, Technische Universität München, D-85748 Garching (Germany); Seemann, K. M. [Physik-Department E21 & Heinz Maier-Leibnitz Zentrum (MLZ), Technische Universität München, D-85747 Garching (Germany); Soltwedel, O. [Max-Planck-Institute for Solid State Research, Outstation at MLZ, Lichtenbergstr. 1, 85747 Garching (Germany)
2015-09-21
We report on the fabrication and use of deuterated polyethylene as a coating material for ultra-cold neutron (UCN) storage and transport. The Fermi potential has been determined to be 214 neV, and the wall loss coefficient η is 1.3 × 10{sup 4} per wall collision. The coating technique allows for a wide range of applications in this field of physics. In particular, flexible and quasi-massless UCN guides with slit-less shutters and seamless UCN storage volumes become possible. These properties enable the use in next-generation measurements of the electric dipole moment of the neutron.
Neutron dynamics and materials damage in inertial fusion reactors
Energy Technology Data Exchange (ETDEWEB)
Perlado, J.M.; Malerba, L.; Marian, J.; Lodi, D. [Universidad Politecnica de Madrid, Instituto de Fusion Nuclear, DENIM (Spain); Diaz de la Rubia, T.; Alonso, E. [Lawrence Livermore National Lab., Chemistry and Materials Science Div., CA (United States)
2000-07-01
Energy, tritium breeding, damage and activation in IFE blankets are affected by the neutron dynamics in the target, and the type of protection. The time-dependent structure of neutron transport (pulsed and very high intensity) led to new consequences in some key parameters (damage). A basic simulation on damage on SiC reveals an opposite response of the two sublattices: ductile Si, and fragile C. Radiation induced amorphization (accumulation of damage) is investigated and only argued for very low temperatures, and simulations of damage in metallic alloys show precipitates formation in a initial phase. (authors)
Fast neutron imaging device and method
Popov, Vladimir; Degtiarenko, Pavel; Musatov, Igor V.
2014-02-11
A fast neutron imaging apparatus and method of constructing fast neutron radiography images, the apparatus including a neutron source and a detector that provides event-by-event acquisition of position and energy deposition, and optionally timing and pulse shape for each individual neutron event detected by the detector. The method for constructing fast neutron radiography images utilizes the apparatus of the invention.
Progress of Neutron Bubble Detectors in CIAE
Institute of Scientific and Technical Information of China (English)
2002-01-01
Neutron bubble detector is the only personal neutron dosimeter which has adequate neutronsensitivity to meet the implications of the ICRP 60 recommendations for neutron dosimetry. It canmonitor the wide range of neutron energy, for example 100 eV to 10 MeV And it becomes a significanttool for neutron dose monitoring at the environment of nuclear energy.
Measurement of neutron scattering lengths using neutron interferometry
Shahi, Chandra B.
This thesis describes the details on building a new Neutron Interferometry and Optics Facility (NIOFa), the measurement of the incoherent neutron scattering length bi of 3He, and the measurement of the coherent neutron scattering length bc of 4He at National Institute of Standards and Technology (NIST) Center for Neutron Research (NCNR). A new monochromatic beamline and facility has been installed at the NCNR devoted to neutron interferometry in the research areas of spin control, spin manipulation, quantum mechanics, quantum information science, spintronics, and material science. This facility is possible in part because of advances in decoherence free subspace interferometer designs that have demonstrated consistent contrast in the presence of vibrational noise; a major environmental constraint that has prevented neutron interferometry from being applied at other neutron facilities. This new facility, NIOFa, is located in the guide hall of the NCNR upstream of the existing Neutron Interferometry and Optics Facility (NIOF) and has several advantages over the NIOF including higher incident flux, better neutron polarization, and increased accessibility. The measurement of the incoherent neutron scattering length bi of 3He was done using a (220) single silicon crystal skew symmetric interferometer. This experiment requires both a polarized beam and a polarized target. We report bi = -2.35 +/- 0.014 (stat.) +/- 0.014 (syst.). This experiment is a revision of the previous experiment which was done in 2008, and partially explains the non-zero phase shift seen in 2008 experiment even if target cell was completely unpolarized. The measurement of the coherent neutron scattering length b c of the 4He was done using a (111) single silicon crystal interferometer. The neutron interferometry and optics facility at NIST had been used previously to determine the coherent scattering lengths for n- 1H, n-2H, and n-3He to less than 1% relative uncertainty. We report bc of the 4He
Neutron electric polarizability
Alexandru, Andrei
2009-01-01
We use the background field method to extract the "connected" piece of the neutron electric polarizability. We present results for quenched simulations using both clover and Wilson fermions and discuss our experience in extracting the mass shifts and the challenges we encountered when we lowered the quark mass. For the neutron we find that as the pion mass is lowered below $500\\MeV$, the polarizability starts rising in agreement with predictions from chiral perturbation theory. For our lowest pion mass, $m_\\pi=320\\MeV$, we find that $\\alpha_n = 3.8(1.3)\\times 10^{-4}\\fm^3$, which is still only one third of the experimental value. We also present results for the neutral pion; we find that its polarizability turns negative for pion masses smaller than $500\\MeV$ which is puzzling.
Neutron beam measurement dosimetry
Energy Technology Data Exchange (ETDEWEB)
Amaro, C.R. [Idaho National Engineering Lab., Idaho Falls, ID (United States)
1995-11-01
This report describes animal dosimetry studies and phantom measurements. During 1994, 12 dogs were irradiated at BMRR as part of a 4 fraction dose tolerance study. The animals were first infused with BSH and irradiated daily for 4 consecutive days. BNL irradiated 2 beagles as part of their dose tolerance study using BPA fructose. In addition, a dog at WSU was irradiated at BMRR after an infusion of BPA fructose. During 1994, the INEL BNCT dosimetry team measured neutron flux and gamma dose profiles in two phantoms exposed to the epithermal neutron beam at the BMRR. These measurements were performed as a preparatory step to the commencement of human clinical trials in progress at the BMRR.
Energy Technology Data Exchange (ETDEWEB)
Kegel, Gunter H.R.; Egan, James J
2007-04-18
This project covers four principal areas of research: Elastic and inelastic neutron scattering studies in odd-A terbium, thulium and other highly deformed nuclei near A=160 with special regard to interband transitions and to the investigation of the direct-interaction versus the compound-nucleus excitation process in these nuclei. Examination of new, fast photomultiplier tubes suitable for use in a miniaturized neutron-time-of-flight spectrometer. Measurement of certain inelastic cross sections of 238U. Determination of the multiplicity of prompt fission gamma rays in even-A fissile actinides. Energies and mean lives of fission isomers produced by fast fission of even-Z, even-A actinides. Study of the mean life of 7Be in different host matrices and its possible astro-physical significance.
Neutrons from Antiproton Irradiation
DEFF Research Database (Denmark)
Bassler, Niels; Holzscheiter, Michael; Petersen, Jørgen B.B.
Background: Radiotherapy with Antiprotons is currently investigated by the AD-4/ACE collaboration. The hypothesis is that the additional energy released from the antiprotons annihilating at the target nuclei can enable a reduced dose in the entry channel of the primary beam. Furthermore an enhanced...... relative biological effect (RBE) has already been beam measured in spread out Bragg peaks of antiprotons, relative to that found in the plateau region. However, the antiproton annihilation process is associated with a substantial release of secondary particles which contribute to the dose outside...... the neutron spectrum. Additionally, we used a cylindrical polystyrene loaded with several pairs of thermoluminescent detectors containing Lithium-6 and Lithium-7, which effectively detects thermalized neutrons. The obtained results are compared with FLUKA imulations. Results: The results obtained...
High-resolution neutron microtomography with noiseless neutron counting detector
Energy Technology Data Exchange (ETDEWEB)
Tremsin, A.S., E-mail: ast@ssl.berkeley.edu [Space Sciences Laboratory, University of California, Berkeley, CA 94720 (United States); McPhate, J.B.; Vallerga, J.V.; Siegmund, O.H.W. [Space Sciences Laboratory, University of California, Berkeley, CA 94720 (United States); Feller, W.B. [Nova Scientific Inc., 10 Picker Road, Sturbridge, MA 01566 (United States); Lehmann, E. [Paul Scherrer Institute, CH-5232 Villigen (Switzerland); Butler, L.G. [Louisiana State University, Baton Rouge, LA 70803 (United States); Dawson, M. [Helmholtz Centre Berlin for Materials and Energy (Germany)
2011-10-01
The improved collimation and intensity of thermal and cold neutron beamlines combined with recent advances in neutron imaging devices enable high-resolution neutron radiography and microtomography, which can provide information on the internal structure of objects not achievable with conventional X-ray imaging techniques. Neutron detection efficiency, spatial and temporal resolution (important for the studies of dynamic processes) and low background count rate are among the crucial parameters defining the quality of radiographic images and tomographic reconstructions. The unique capabilities of neutron counting detectors with neutron-sensitive microchannel plates (MCPs) and with Timepix CMOS readouts providing high neutron detection efficiency ({approx}70% for cold neutrons), spatial resolutions ranging from 15 to 55 {mu}m and a temporal resolution of {approx}1 {mu}s-combined with the virtual absence of readout noise-make these devices very attractive for high-resolution microtomography. In this paper we demonstrate the capabilities of an MCP-Timepix detection system applied to microtomographic imaging, performed at the ICON cold neutron facility of the Paul Scherrer Institute. The high resolution and the absence of readout noise enable accurate reconstruction of texture in a relatively opaque wood sample, differentiation of internal tissues of a fly and imaging of individual {approx}400 {mu}m grains in an organic powder encapsulated in a {approx}700 {mu}m thick metal casing.
Neutrons in the moon. [neutron flux and production rate calculations
Kornblum, J. J.; Fireman, E. L.; Levine, M.; Aronson, A.
1973-01-01
Neutron fluxes for energies between 15 MeV and thermal at depths of 0 to 300 g/sq cm in the moon are calculated by the discrete ordinate mathod with the ANISN code. With the energy spectrum of Lingenfelter et al. (1972). A total neutron-production rate for the moon of 26 plus or minus neutrons/sq cm sec is determined from the Ar-37 activity measurements in the Apollo 16 drill string, which are found to have a depth dependence in accordance with a neutron source function that decreases exponentially with an attenuation length of 155 g/sq cm.
Are there good probes for the di-neutron correlation in light neutron-rich nuclei?
Hagino, K
2015-01-01
The di-neutron correlation is a spatial correlation with which two valence neutrons are located at a similar position inside a nucleus. We discuss possible experimental probes for the di-neutron correlation. This includes the Coulomb breakup and the pair transfer reactions of neutron-rich nuclei, and the direct two-neutron decays of nuclei beyond the neutron drip-line.
2000-01-01
View of the n_TOF tube with members of the design and construction team of the facility(from left to right: R. Magnin/LHC, E. Radermacher/EP, P. Cennini/EP and R. Cappi/PS). A new experimental facility was inaugurated at CERN on Wednesday 8 November. The neutron Time Of Flight (n_TOF) facility received its first protons from the PS at 10:55. With an intensity of 1 x 1011 protons per cycle on the n_TOF target, an intense neutron beam has been produced at CERN for the first time, opening the door to many new avenues of research including, for example, neutron induced cross-section measurements. The facility is an offspring of the work by Carlo Rubbia and his group on the novel idea of an Energy Amplifier. The basic idea was successfully tested at the PS with the FEAT experiment and later with the TARC experiment, where the feasibility of transmutation of long-lived products by Adiabatic Resonance Crossing (ARC) was confirmed. This led to the possibility of radio-isotope production for medical applications, fo...
Neutron instrumentation for biology
Energy Technology Data Exchange (ETDEWEB)
Mason, S.A. [Institut Laue-Langevin, Grenoble (France)
1994-12-31
In the October 1994 round of proposals at the ILL, the external biology review sub- committee was asked to allocate neutron beam time to a wide range of experiments, on almost half the total number of scheduled neutron instruments: on 3 diffractometers, on 3 small angle scattering instruments, and on some 6 inelastic scattering spectrometers. In the 3.5 years since the temporary reactor shutdown, the ILL`s management structure has been optimized, budgets and staff have been trimmed, the ILL reactor has been re-built, and many of the instruments up-graded, many powerful (mainly Unix) workstations have been introduced, and the neighboring European Synchrotron Radiation Facility has established itself as the leading synchrotron radiation source and has started its official user program. The ILL reactor remains the world`s most intense dedicated neutron source. In this challenging context, it is of interest to review briefly the park of ILL instruments used to study the structure and energetics of small and large biological systems. A brief summary will be made of each class of experiments actually proposed in the latest ILL proposal round.
Euratom Neutron Radiography Working Group
DEFF Research Database (Denmark)
Domanus, Joseph Czeslaw
1986-01-01
In 1979 a Neutron Radiography Working Group (NRWG) was constituted within Buratom with the participation of all centers within the European Community at which neutron facilities were available. The main purpose of NRWG was to standardize methods and procedures used in neutron radiography of nuclear...... reactor fuel as well as establish standards for radiographic image quality of neutron radiographs. The NRWG meets once a year in each of the neutron radiography centers to review the progress made and draw plans for the future. Besides, ad-hoc sub-groups or. different topics within the field of neutron...... radiography are constituted. This paper reviews the activities and achievements of the NRWG and its sub-groups....
Neutron background estimates in GESA
Directory of Open Access Journals (Sweden)
Fernandes A.C.
2014-01-01
Full Text Available The SIMPLE project looks for nuclear recoil events generated by rare dark matter scattering interactions. Nuclear recoils are also produced by more prevalent cosmogenic neutron interactions. While the rock overburden shields against (μ,n neutrons to below 10−8 cm−2 s−1, it itself contributes via radio-impurities. Additional shielding of these is similar, both suppressing and contributing neutrons. We report on the Monte Carlo (MCNP estimation of the on-detector neutron backgrounds for the SIMPLE experiment located in the GESA facility of the Laboratoire Souterrain à Bas Bruit, and its use in defining additional shielding for measurements which have led to a reduction in the extrinsic neutron background to ∼ 5 × 10−3 evts/kgd. The calculated event rate induced by the neutron background is ∼ 0,3 evts/kgd, with a dominant contribution from the detector container.
Energy Technology Data Exchange (ETDEWEB)
O' Rourke, Patrick Francis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2016-10-27
The purpose of this report is to provide the reader with an understanding of how a Monte Carlo neutron transport code was written, developed, and evolved to calculate the probability distribution functions (PDFs) and their moments for the neutron number at a final time as well as the cumulative fission number, along with introducing several basic Monte Carlo concepts.
Toward a final design for the Birmingham boron neutron capture therapy neutron beam.
Allen, D A; Beynon, T D; Green, S; James, N D
1999-01-01
This paper is concerned with the proposed Birmingham accelerator-based epithermal neutron beam for boron neutron capture therapy (BNCT). Details of the final moderator design, such as beam delimiter, shield, and beam exit surface shape are considered. Monte Carlo radiation transport simulations with a head and body phantom have shown that a simple flat moderator beam exit surface is preferable to the previously envisioned spherical design. Dose rates to individual body organs during treatment have been calculated using a standard MIRD phantom. We have shown that a simple polyethylene shield, doped with natural lithium, is sufficient to provide adequate protection to the rest of the body during head irradiations. The effect upon the head phantom dose distributions of the use of such a shield to delimit the therapy beam has been evaluated.
Cardone, Fabio; Perconti, Walter; Petrucci, Andrea; Rosada, Alberto
2013-01-01
In this paper a new type of passive neutron detector based on the already existing one, CR39, is described. Its operation was verified by three different neutron sources: an Americium-Beryllium (Am241-Be) source; a TRIGA type nuclear reactor; and a fast neutron reactor called TAPIRO. The obtained results, reported here, positively confirm its operation and the accountability of the new developed detecting technique.
Novel concept for neutron detection: proportional counter filled with 10B nanoparticle aerosol
Amaro, F. D.; Monteiro, C. M. B.; dos Santos, J. M. F.; Antognini, A.
2017-01-01
The high neutron detection efficiency, good gamma-ray discrimination and non-toxicity of 3He made of proportional counters filled with this gas the obvious choice for neutron detection, particularly in radiation portal monitors (RPM), used to control the illicit transport of nuclear material, of which neutron detectors are key components. 3He is very rare and during the last decade this gas has become increasingly difficult to acquire. With the exception of BF3, which is toxic, no other gas can be used for neutron detection in proportional counters. We present an alternative where the 3He atoms are replaced by nanoparticles made of another neutron sensitive material, 10B. The particles are dispersed in a gaseous volume, forming an aerosol with neutron sensitive properties. A proportional counter filled with such aerosol was exposed to a thermal neutron beam and the recorded response indicates that the neutrons have interacted with the particles in the aerosol. This original technique, which transforms a standard proportional gas mixture into a neutron sensitive aerosol, is a breakthrough in the field of radiation detection and has the potential to become an alternative to the use of 3He in proportional counters. PMID:28181520
An ultracold neutron source at the NC State University PULSTAR reactor
Korobkina, E.; Wehring, B. W.; Hawari, A. I.; Young, A. R.; Huffman, P. R.; Golub, R.; Xu, Y.; Palmquist, G.
2007-08-01
Research and development is being completed for an ultracold neutron (UCN) source to be installed at the PULSTAR reactor on the campus of North Carolina State University (NCSU). The objective is to establish a university-based UCN facility with sufficient UCN intensity to allow world-class fundamental and applied research with UCN. To maximize the UCN yield, a solid ortho-D 2 converter will be implemented coupled to two moderators, D 2O at room temperature, to thermalize reactor neutrons, and solid CH 4, to moderate the thermal neutrons to cold-neutron energies. The source assembly will be located in a tank of D 2O in the space previously occupied by the thermal column of the PULSTAR reactor. Neutrons leaving a bare face of the reactor core enter the D 2O tank through a 45×45 cm cross-sectional area void between the reactor core and the D 2O tank. Liquid He will cool the disk-shaped UCN converter to below 5 K. Independently, He gas will cool the cup-shaped CH 4 cold-neutron moderator to an optimum temperature between 20 and 40 K. The UCN will be transported from the converter to experiments by a guide with an inside diameter of 16 cm. Research areas being considered for the PULSTAR UCN source include time-reversal violation in neutron beta decay, neutron lifetime determination, support measurements for a neutron electric-dipole-moment search, and nanoscience applications.
Characterization of the high-energy neutron beam of the PRISMA beamline using a diamond detector
Cazzaniga, C.; Frost, C. D.; Minniti, T.; Schooneveld, E.; Perelli Cippo, E.; Tardocchi, M.; Rebai, M.; Gorini, G.
2016-07-01
The high-energy neutron component (En > 10 MeV) of the neutron spectrum of PRISMA, a beam-line at the ISIS spallation source, has been characterized for the first time. Neutron measurements using a Single-crystal Diamond Detector at a short-pulse source are obtained by a combination of pulse height and time of flight analysis. An XY scan provides a 2D map of the high-energy neutron beam which has a diameter of about 40 mm. The high neutron flux, that has been found to be (3.8 ± 0.7) · 105 cm-2s-1 for En > 10 MeV in the centre, opens up for a possible application of the beam-line as a high-energy neutron irradiation position. Results are of interest for the development of the ChipIR beam-line, which will feature an atmospheric-like neutron spectrum for chip irradiation experiment. Furthermore, these results demonstrate that diamond detectors can be used at spallation sources to investigate the transport of high-energy neutrons down instruments which is of interest in general to designers as high-energy neutrons are a source of background in thermal beamlines.
Novel concept for neutron detection: proportional counter filled with 10B nanoparticle aerosol
Amaro, F. D.; Monteiro, C. M. B.; Dos Santos, J. M. F.; Antognini, A.
2017-02-01
The high neutron detection efficiency, good gamma-ray discrimination and non-toxicity of 3He made of proportional counters filled with this gas the obvious choice for neutron detection, particularly in radiation portal monitors (RPM), used to control the illicit transport of nuclear material, of which neutron detectors are key components. 3He is very rare and during the last decade this gas has become increasingly difficult to acquire. With the exception of BF3, which is toxic, no other gas can be used for neutron detection in proportional counters. We present an alternative where the 3He atoms are replaced by nanoparticles made of another neutron sensitive material, 10B. The particles are dispersed in a gaseous volume, forming an aerosol with neutron sensitive properties. A proportional counter filled with such aerosol was exposed to a thermal neutron beam and the recorded response indicates that the neutrons have interacted with the particles in the aerosol. This original technique, which transforms a standard proportional gas mixture into a neutron sensitive aerosol, is a breakthrough in the field of radiation detection and has the potential to become an alternative to the use of 3He in proportional counters.
Radiation transport. Progress report, April 1-December 31, 1983
Energy Technology Data Exchange (ETDEWEB)
O' Dell, R.D.
1984-10-01
Research and development progress in radiation transport by the Los Alamos National Laboratory's Group X-6 for the last nine months of CY 83 is reported. Included are unclassified tasks in the areas of Fission Reactor Neutronics, Deterministic Transport Methods, Monte Carlo Radiation Transport, and Cross Sections and Physics.
Energy Technology Data Exchange (ETDEWEB)
Ribeiro, Rosane M.; Santos, Denison de S.; Queiroz Filho, Pedro P. de; Mauricio, CLaudia L.P.; Silva, Livia K. da; Pessanha, Paula R., E-mail: rosanemribeiro@oi.com.br [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil)
2014-07-01
Fluence to dose equivalent conversion coefficients provide the basis for the calculation of area and personal monitors. Recently, the ICRP has started a revision of these coefficients, including new Monte Carlo codes for benchmarking. So far, little information is available about neutron transport below 10 MeV in tissue-equivalent (TE) material performed with Monte Carlo GEANT4 code. The objective of this work is to calculate neutron fluence to personal dose equivalent conversion coefficients, H{sub p} (10)/Φ, with GEANT4 code. The incidence of monoenergetic neutrons was simulated as an expanded and aligned field, with energies ranging between thermal neutrons to 10 MeV on the ICRU slab of dimension 30 x 30 x 15 cm{sup 3}, composed of 76.2% of oxygen, 10.1% of hydrogen, 11.1% of carbon and 2.6% of nitrogen. For all incident energy, a cylindrical sensitive volume is placed at a depth of 10 mm, in the largest surface of the slab (30 x 30 cm{sup 2}). Physic process are included for neutrons, photons and charged particles, and calculations are made for neutrons and secondary particles which reach the sensitive volume. Results obtained are thus compared with values published in ICRP 74. Neutron fluence in the sensitive volume was calculated for benchmarking. The Monte Carlo GEANT4 code was found to be appropriate to calculate neutron doses at energies below 10 MeV correctly. (author)
Adaptive Nodal Transport Methods for Reactor Transient Analysis
Energy Technology Data Exchange (ETDEWEB)
Thomas Downar; E. Lewis
2005-08-31
Develop methods for adaptively treating the angular, spatial, and time dependence of the neutron flux in reactor transient analysis. These methods were demonstrated in the DOE transport nodal code VARIANT and the US NRC spatial kinetics code, PARCS.
NEUTRON IMAGING, RADIOGRAPHY AND TOMOGRAPHY.
Energy Technology Data Exchange (ETDEWEB)
SMITH,G.C.
2002-03-01
Neutrons are an invaluable probe in a wide range of scientific, medical and commercial endeavors. Many of these applications require the recording of an image of the neutron signal, either in one-dimension or in two-dimensions. We summarize the reactions of neutrons with the most important elements that are used for their detection. A description is then given of the major techniques used in neutron imaging, with emphasis on the detection media and position readout principle. Important characteristics such as position resolution, linearity, counting rate capability and sensitivity to gamma-background are discussed. Finally, the application of a subset of these instruments in radiology and tomography is described.
Centrifugal quantum states of neutrons
Nesvizhevsky, V. V.; Petukhov, A. K.; Protasov, K. V.; Voronin, A. Yu.
2008-09-01
We propose a method for observation of the quasistationary states of neutrons localized near a curved mirror surface. The bounding effective well is formed by the centrifugal potential and the mirror Fermi potential. This phenomenon is an example of an exactly solvable “quantum bouncer” problem that can be studied experimentally. It could provide a promising tool for studying fundamental neutron-matter interactions, as well as quantum neutron optics and surface physics effects. We develop a formalism that describes quantitatively the neutron motion near the mirror surface. The effects of mirror roughness are taken into account.
Ukraine experimental neutron source facility.
Energy Technology Data Exchange (ETDEWEB)
Gohar, Y.; Bolshinsky, I.; Nekludov, I.; Karnaukhov, I. (Nuclear Engineering Division); (INL); (Kharkov Institute of Physics and Technology)
2008-01-01
Kharkov Institute of Physics and Technology (KIPT) of Ukraine has a plan to construct an experimental neutron source facility. The facility has been developed for producing medical isotopes, training young nuclear professionals, supporting the Ukraine nuclear industry, providing capability for performing reactor physics, material research, and basic science experiments. Argonne National Laboratory (ANL) of USA is collaborating with KIPT on developing this facility. A driven subcritical assembly utilizing the KIPT electron accelerator with a target assembly is used to generate the neutron source. The target assembly utilizes tungsten or uranium for neutron production through photonuclear reactions with 100-KW of electron beam power. The neutron source intensity, spectrum, and spatial distribution have been studied to maximize the neutron yield and satisfy different engineering requirements. The subcritical assembly is designed to obtain the highest possible neutron flux intensity with a subcriticality of 0.98. Low enrichment uranium is used for the fuel material because it enhances the neutron source performance. Safety, reliability, and environmental considerations are included in the facility conceptual design. Horizontal neutron channels are incorporated for performing basic research including cold neutron source. This paper describes the conceptual design and summarizes some of the related analyses.
Federal Laboratory Consortium — Researchers working at the Los Alamos Neutron Science Center and eight other member institutions of an international collaboration are constructing the most intense...
Reactor neutrons in nuclear astrophysics
Reifarth, René; Glorius, Jan; Göbel, Kathrin; Heftrich, Tanja; Jentschel, Michael; Jurado, Beatriz; Käppeler, Franz; Köster, Ulli; Langer, Christoph; Litvinov, Yuri A.; Weigand, Mario
2017-09-01
The huge neutron fluxes offer the possibility to use research reactors to produce isotopes of interest, which can be investigated afterwards. An example is the half-lives of long-lived isotopes like 129I. A direct usage of reactor neutrons in the astrophysical energy regime is only possible, if the corresponding ions are not at rest in the laboratory frame. The combination of an ion storage ring with a reactor and a neutron guide could open the path to direct measurements of neutron-induced cross sections on short-lived radioactive isotopes in the astrophysically interesting energy regime.
Measurement of neutron diffraction with compact neutron source RANS
Ikeda, Y.; Takamura, M.; Taketani, A.; Sunaga, H.; Otake, Y.; Suzuki, H.; Kumagai, M.; Oba, Y.; Hama, T.
2016-11-01
Diffraction is used as a measurement technique for crystal structure. X-rays or electron beam with wavelength that is close to the lattice constant of the crystal is often used for the measurement. They have sensitivity in surface (0.01mm) of heavy metals due to the mean free path for heavy ions. Neutron diffraction has the probe of the internal structure of the heavy metals because it has a longer mean free path than that of the X-rays or the electrons. However, the neutron diffraction measurement is not widely used because large facilities are required in the many neutron sources. RANS (Riken Accelerator-driven Compact Neutron Source) is developed as a neutron source which is usable easily in laboratories and factories. In RANS, fast neutrons are generated by 7MeV protons colliding on a Be target. Some fast neutrons are moderated with polyethylene to thermal neutrons. The thermal neutrons of 10meV which have wavelength of 10nm can be used for the diffraction measurement. In this study, the texture evolution in steels was measured with RANS and the validity of the compact neutron source was proved. The texture of IF steel sheets with the thickness of 1.0mm was measured with 10minutes run. The resolution is 2% and is enough to analyze a evolution in texture due to compression/tensile deformation or a volume fraction of two phases in the steel sample. These results have proven the possibility to use compact neutron source for the analysis of mesoscopic structure of metallic materials.
Energy Technology Data Exchange (ETDEWEB)
Masuda, Y. [National Lab. for High Energy Physics, Tsukuba, Ibaraki (Japan)
1996-08-01
Experimental method to detect a T-odd correlation term in neutron propagation through a nuclear target is discussed. The correlation term is between the neutron spin, neutron momentum and nuclear spin. (author)
Fundamental physics research and neutron interferometry
Energy Technology Data Exchange (ETDEWEB)
Ioffe, A. [Hahn-Meitner-Institut Berlin GmbH (Germany)
1996-08-01
The possibility of the use of an extremely sensitive neutron interferometry technique for the study of electromagnetic structure of the neutron and the parity non-conservative effects in neutron spin rotation is discussed. (author)
Nuclear superfluidity and cooling time of neutron-star crust
Energy Technology Data Exchange (ETDEWEB)
Monrozeau, C.; Margueron, J. [Institut de Physique Nucleaire, Universite Paris Sud, F-91406 Orsay CEDEX (France); Sandulescu, N. [Institut de Physique Nucleaire, Universite Paris Sud, F-91406 Orsay CEDEX (France); Institute of Physics and Nuclear Engineering, RO-76900 Bucharest (Romania)
2007-03-15
We analyse the effect of neutron superfluidity on the cooling time of inner crust matter in neutron stars, in the case of a rapid cooling of the core. The specific heat of the inner crust, which determines the thermal response of the crust, is calculated in the framework of HFB approach at finite temperature. The calculations are performed with two paring forces chosen to simulate the pairing properties of uniform neutron matter corresponding respectively to Gogny-BCS approximation and to many-body techniques including polarisation effects. Using a simple model for the heat transport across the inner crust, it is shown that the two pairing forces give very different values for the cooling time. (authors)
Neutron fluctuations a treatise on the physics of branching processes
Pazsit, Imre; Pzsit, Imre
2007-01-01
The transport of neutrons in a multiplying system is an area of branching processes with a clear formalism. This book presents an account of the mathematical tools used in describing branching processes, which are then used to derive a large number of properties of the neutron distribution in multiplying systems with or without an external source. In the second part of the book, the theory is applied to the description of the neutron fluctuations in nuclear reactor cores as well as in small samples of fissile material. The question of how to extract information about the system under study is discussed. In particular the measurement of the reactivity of subcritical cores, driven with various Poisson and non-Poisson (pulsed) sources, and the identification of fissile material samples, is illustrated. The book gives pragmatic information for those planning and executing and evaluating experiments on such systems. - Gives a complete treatise of the mathematics of branching particle processes, and in particular n...
Total cross sections for ultracold neutrons scattered from gases
Seestrom, S. J.; Adamek, E. R.; Barlow, D.; Blatnik, M.; Broussard, L. J.; Callahan, N. B.; Clayton, S. M.; Cude-Woods, C.; Currie, S.; Dees, E. B.; Fox, W.; Hoffbauer, M.; Hickerson, K. P.; Holley, A. T.; Liu, C.-Y.; Makela, M.; Medina, J.; Morley, D. J.; Morris, C. L.; Pattie, R. W.; Ramsey, J.; Roberts, A.; Salvat, D. J.; Saunders, A.; Sharapov, E. I.; Sjue, S. K. L.; Slaughter, B. A.; Walstrom, P. L.; Wang, Z.; Wexler, J.; Womack, T. L.; Young, A. R.; Vanderwerp, J.; Zeck, B. A.
2017-01-01
We have followed up on our previous measurements of upscattering of ultracold neutrons (UCNs) from a series of gases by making measurements of total cross sections on the following gases hydrogen, ethane, methane, isobutene, n -butane, ethylene, water vapor, propane, neopentane, isopropyl alcohol, and 3He . The values of these cross sections are important for estimating the loss rate of trapped neutrons due to residual gas and are relevant to neutron lifetime measurements using UCNs. The effects of the UCN velocity and path-length distributions were accounted for in the analysis using a Monte Carlo transport code. Results are compared to our previous measurements and with the known absorption cross section for 3He scaled to our UCN energy. We find that the total cross sections for the hydrocarbon gases are reasonably described by a function linear in the number of hydrogen atoms in the molecule.
Neutron mirror development for VCN/UCN sources
Energy Technology Data Exchange (ETDEWEB)
Kawabata, Y.; Tasaki, S.; Hino, M. [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst; Suzuki, M. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Somemiya, K. [Kyoto Univ. (Japan). Faculty of Engineering; Wakata, A. [R and D center, Mitsubishi pencil Co., Fujioka, Gumma (Japan); Nakayama, M. [R and D center, TDK Co., Ichikawa, Chiba (Japan)
2001-03-01
Several types of neutron mirrors and monochromator have been developed for VCN and UCN facilities. As the VCN guide tube must be set very close to the CNS cell, it will suffer a severe irradiation. Neutron mirrors enduring a hard environment are essential. Replica supermirrors and polished glassy carbon mirrors have been developed for VCN extraction. A wide band monochromator consisting of a stack of four multilayers on two Si wafers has been developed. One multilayer has 201 Ni/Ti layers. The layer thickness is gradually changed in order to extend the neutron reflection wavelength range similar to a supermirror. It is required as blades of a proposed new type UCN turbine. Development of deuterated diamond-like carbon mirrors is also in progress for the UCN transportation. (author)
Design of Testing Set-up for Nuclear Fuel Rod by Neutron Radiography at CARR
Institute of Scientific and Technical Information of China (English)
WEI; Guo-hai; HAN; Song-bai; WANG; Hong-li; HAO; Li-jie; WU; Mei-mei; HE; Lin-feng; WANG; Yu; LIU; Yun-tao; SUN; Kai; CHEN; Dong-feng
2012-01-01
<正>An experimental set-up dedicated to non-destructively test a 15 cm long pressurized water reactor (PWR) nuclear fuel rod by neutron radiography (NR) is designed and fabricated. It consists of three parts: Transport container, imaging block and steel support. The design of the transport container was optimized with Monte-Carlo simulation by the MCNP code.
Development and performance tests of NSTAR, a new type of compact neutron detector
Pawelczak, Iwona Agnieszka
capture time, a parameter determining the detector speed, turned out to be = (21.7 +/- 0.2)micros. The experimental values are in good agreement with theoretical estimates based on simulation calculations obtained with a modified version of a neutron transport code (DENIS(E)).
Calculating and measuring thermal neutrons exiting from neutron diffractometers collimators
Tafazolee, K
2000-01-01
process, effectiveness of them are studied for the enhancement of the available system. Final conclusion from the simulation process, indicates that the heavy water with the thickness of 50 to 60 cm. is the best moderator for gaining the better thermal neutrons flux for enhancement of P.N.D. in the T.R.R. Powder Neutron Diffractometer y (P.N.D.) is relatively good and practical way for identification of the 3 dimensional construction of materials. In order to exploit the capabilities of this method, in one of the neutron beam of the Tehran Research Reactor (T.R.R.), a collimator embedded inside the concrete wall, direct the neutrons produced in the core reactor towards a monochromator e. Neutrons having been monochromated by 2 nd collimator are then directed towards the sample. Then the pattern of diffracted neutrons from the sample are studied. In order to make the best out of it, neutrons coming to sit on the sample must be of the thermal type. That means the number/amount of thermal neutrons flux in compar...
Probing neutron star physics using accreting neutron stars
Directory of Open Access Journals (Sweden)
Patruno A.
2010-10-01
Full Text Available We give an obervational overview of the accreting neutron stars systems as probes of neutron star physics. In particular we focus on the results obtained from the periodic timing of accreting millisecond X-ray pulsars in outburst and from the measurement of X-ray spectra of accreting neutron stars during quiescence. In the ﬁrst part of this overview we show that the X-ray pulses are contaminated by a large amount of noise of uncertain origin, and that all these neutron stars do not show evidence of spin variations during the outburst. We present also some recent developments on the presence of intermittency in three accreting millisecond X-ray pulsars and investigate the reason why only a small number of accreting neutron stars show X-ray pulsations and why none of these pulsars shows sub-millisecond spin periods. In the second part of the overview we introduce the observational technique that allows the study of neutron star cooling in accreting systems as probes of neutron star internal composition and equation of state. We explain the phenomenon of the deep crustal heating and present some recent developments on several quasi persistent X-ray sources where a cooling neutron star has been observed.
Sequential measurements of environmental neutron energy spectrum and neutron dose
Energy Technology Data Exchange (ETDEWEB)
Nunomiya, Tomoya; Nakamura, Takashi; Suzuki, Hiroyuki; Terunuma, Kazutaka; Hirabayashi, Naoya; Sato, Youichi; Abe, Sigeru; Rasolonjatovo A.H, Danielle [Tohoku Univ., Dept. of Quantum Science and Energy Engineering, Sendai, Miyagi (Japan)
2003-03-01
From April 2001, neutron energy spectra and neutron dose were sequentially measured using 5'' -rem counter and {sup 3}He multi-moderator spectrometer (Boner boll) at Kawauchi-campus of Tohoku University. These data were collected about the relation between the dose level and the solar activities. (author)
Magnetic correlations in oxides: Neutron diffraction and neutron depolarization study
Indian Academy of Sciences (India)
S M Yusuf
2008-10-01
We have studied magnetic correlations in several oxide materials that belong to colossal magnetoresistive, naturally occurring layered oxide showing low-dimensional magnetic ordering, solid oxide fuel cell interconnect materials, and magnetic nanoparticles using neutron diffraction and neutron depolarization techniques. In this paper, an overview of some of these results is given.
Synovectomy by Neutron capture; Sinovectomia por captura de neutrones
Energy Technology Data Exchange (ETDEWEB)
Vega C, H.R.; Torres M, C. [Centro Regional de Estudios Nucleares, Universidad Autonoma de Zacatecas, C. Cipres 10, Fracc. La Penuela, 98000 Zacatecas (Mexico)
1998-12-31
The Synovectomy by Neutron capture has as purpose the treatment of the rheumatoid arthritis, illness which at present does not have a definitive curing. This therapy requires a neutron source for irradiating the articulation affected. The energy spectra and the intensity of these neutrons are fundamental since these neutrons induce nuclear reactions of capture with Boron-10 inside the articulation and the freely energy of these reactions is transferred at the productive tissue of synovial liquid, annihilating it. In this work it is presented the neutron spectra results obtained with moderator packings of spherical geometry which contains in its center a Pu{sup 239} Be source. The calculations were realized through Monte Carlo method. The moderators assayed were light water, heavy water base and the both combination of them. The spectra obtained, the average energy, the neutron total number by neutron emitted by source, the thermal neutron percentage and the dose equivalent allow us to suggest that the moderator packing more adequate is what has a light water thickness 0.5 cm (radius 2 cm) and 24.5 cm heavy water (radius 26.5 cm). (Author)
Energy Technology Data Exchange (ETDEWEB)
Caldeira, A.D.; Dias, A.F.; Garcia, R.D.M. [Centro Tecnico Aeroespacial (CTA-IEAv), Sao Jose dos Campos, SP (Brazil). Inst. de Estudos Avancados
1997-12-01
A new version of the P{sub N} method for solving the two-group neutron transport equation in plane geometry without up scattering is reported. Contrasting with the conventional matrix approach, the developed version is based on a group-by-group solution of the transport equation which makes it specially well-suited for neutrons slowing-down calculations. (author). 14 refs., 2 tabs.
Time-resolved neutron imaging at ANTARES cold neutron beamline
Tremsin, A S; Tittelmeier, K; Schillinger, B; Schulz, M; Lerche, M; Feller, W B
2015-01-01
In non-destructive evaluation with X-rays light elements embedded in dense, heavy (or high-Z) matrices show little contrast and their structural details can hardly be revealed. Neutron radiography, on the other hand, provides a solution for those cases, in particular for hydrogenous materials, owing to the large neutron scattering cross section of hydrogen and uncorrelated dependency of neutron cross section on the atomic number. The majority of neutron imaging experiments at the present time is conducted with static objects mainly due to the limited flux intensity of neutron beamline facilities and sometimes due to the limitations of the detectors. However, some applications require the studies of dynamic phenomena and can now be conducted at several high intensity beamlines such as the recently rebuilt ANTARES beam line at the FRM-II reactor. In this paper we demonstrate the capabilities of time resolved imaging for repetitive processes, where different phases of the process can be imaged simultaneously and...
Application of imaging plate neutron detector to neutron radiography
Fujine, S; Kamata, M; Etoh, M
1999-01-01
As an imaging plate neutron detector (IP-ND) has been available for thermal neutron radiography (TNR) which has high resolution, high sensitivity and wide range, some basic characteristics of the IP-ND system were measured at the E-2 facility of the KUR. After basic performances of the IP were studied, images with high quality were obtained at a neutron fluence of 2 to 7x10 sup 8 n cm sup - sup 2. It was found that the IP-ND system with Gd sub 2 O sub 3 as a neutron converter material has a higher sensitivity to gamma-ray than that of a conventional film method. As a successful example, clear radiographs of the flat view for the fuel side plates with boron burnable poison were obtained. An application of the IP-ND system to neutron radiography (NR) is presented in this paper.
Cnudde, Veerle; Dierick, Manuel; Vlassenbroeck, Jelle; Masschaele, Bert; Lehmann, Eberhard; Jacobs, Patric; Van Hoorebeke, Luc
2008-01-01
Fluid flow through porous natural building stones is of great importance when studying their weathering processes. Many traditional experiments based on mass changes are available for studying liquid transport in porous stones, such as the determination of the water absorption coefficient by capillarity. Because thermal neutrons experience a strong attenuation by hydrogen, neutron radiography is a suitable technique for the study of water absorption by capillarity in porous stones. However, image contrast can be impaired because hydrogen mainly scatters neutrons rather than absorbing them, resulting in a blurred image. Capillarity results obtained by neutron radiography and by the European Standard 1925 for the determination of the water absorption coefficient by capillarity for natural building stones with a variable porosity were compared. It is illustrated that high-speed neutron radiography can be a useful research tool for the visualization of internal fluid flow inside inorganic building materials such as limestones and sandstones.
Radioactivity measurements of ITER materials using the TFTR D-T neutron field
Energy Technology Data Exchange (ETDEWEB)
Kumar, A.; Abdou, M.A. [California Univ., Los Angeles, CA (United States). School of Engineering and Applied Science; Barnes, C.W. [Los Alamos National Lab., NM (United States); Kugel, H.W. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Loughlin, M.J. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking
1994-08-01
The availability of high D-T fusion neutron yields at TFTR has provided a useful opportunity to directly measure D-T neutron-induced radioactivity in a realistic tokamak fusion reactor environment for materials of vital interest to ITER. These measurements are valuable for characterizing radioactivity in various ITER candidate materials. for validating complex neutron transport calculations, and for meeting fusion reactor licensing requirements. The radioactivity measurements at TFTR involve potential ITER materials including stainless steel 316, vanadium, titanium, chromium, silicon, iron, cobalt, nickel, molybdenum, aluminum, copper, zinc. zirconium, niobium, and tungsten. Small samples of these materials were irradiated close to the plasma and just outside the vacuum vessel wall of TFTR, locations of different neutron energy spectra. Saturation activities for both threshold and capture reactions were measured. Data from dosimetric reactions have been used to obtain preliminary neutron energy spectra. Spectra from the first wall were compared to calculations from ITER and to measurements from accelerator-based tests.
Dworak, D; Drabina, A; Woźnicka, U
2006-07-01
The semi-empirical method of neutron logging tool calibration developed by Prof. J.A. Czubek uses the real and so-called apparent integral neutron parameters of geological formations. To this end, Czubek proposed a few separated calculation methods commonly based on analytical solutions of the neutron transport problem. A new calculation method for the neutron integral parameters is proposed. Quantities like slowing-down length, diffusion and migration lengths, probability to avoid absorption during slowing down, and thermal neutron absorption cross section can be easily approximated using Monte Carlo simulations. A comparison with the results of the analytical method developed by Czubek has been performed for many cases and the observed differences have been explained.
Small angle neutron scattering
Directory of Open Access Journals (Sweden)
Cousin Fabrice
2015-01-01
Full Text Available Small Angle Neutron Scattering (SANS is a technique that enables to probe the 3-D structure of materials on a typical size range lying from ∼ 1 nm up to ∼ a few 100 nm, the obtained information being statistically averaged on a sample whose volume is ∼ 1 cm3. This very rich technique enables to make a full structural characterization of a given object of nanometric dimensions (radius of gyration, shape, volume or mass, fractal dimension, specific area… through the determination of the form factor as well as the determination of the way objects are organized within in a continuous media, and therefore to describe interactions between them, through the determination of the structure factor. The specific properties of neutrons (possibility of tuning the scattering intensity by using the isotopic substitution, sensitivity to magnetism, negligible absorption, low energy of the incident neutrons make it particularly interesting in the fields of soft matter, biophysics, magnetic materials and metallurgy. In particular, the contrast variation methods allow to extract some informations that cannot be obtained by any other experimental techniques. This course is divided in two parts. The first one is devoted to the description of the principle of SANS: basics (formalism, coherent scattering/incoherent scattering, notion of elementary scatterer, form factor analysis (I(q→0, Guinier regime, intermediate regime, Porod regime, polydisperse system, structure factor analysis (2nd Virial coefficient, integral equations, characterization of aggregates, and contrast variation methods (how to create contrast in an homogeneous system, matching in ternary systems, extrapolation to zero concentration, Zero Averaged Contrast. It is illustrated by some representative examples. The second one describes the experimental aspects of SANS to guide user in its future experiments: description of SANS spectrometer, resolution of the spectrometer, optimization of
Neutronic calculations for a final focus system
Energy Technology Data Exchange (ETDEWEB)
Mainardi, E. E-mail: enrico@nuc.berkeley.edu; Premuda, F.; Lee, E
2001-05-21
For heavy-ion fusion and for 'liquid-protected' reactor designs such as HYLIFE-II (Moir et al., Fusion Technol. 25 (1994); HYLIFE-II-Progress Report, UCID-21816, 4-82-100), a mixture of molten salts made of F{sup 10}, Li{sup 6}, Li{sup 7}, Be{sup 9} called flibe allows highly compact target chambers. Smaller chambers will have lower costs and will allow the final-focus magnets to be closer to the target with decreased size of the focus spot and of the driver, as well as drastically reduced costs of IFE electricity. Consequently the superconducting coils of the magnets closer to the chamber will suffer higher radiation damage though they can stand only a certain amount of energy deposited before quenching. The scope of our calculations is essentially the total energy deposited on the magnetic lens system by fusion neutrons and induced {gamma}-rays. Such a study is important for the design of the final focus system itself from the neutronic point of view and indicates some guidelines for a design with six magnets in the beam line. The entire chamber consists of 192 beam lines to provide access of heavy ions that will implode the pellet. A 3-D transport calculation of the radiation penetrating through ducts that takes into account the complexity of the system, requires Monte Carlo methods. The development of efficient and precise models for geometric representation and nuclear analysis is necessary. The parameters are optimized thanks to an accurate analysis of six geometrical models that are developed starting from the simplest. Different configurations are examined employing TART 98 (D.E. Cullen, Lawrence Livermore National Laboratory, UCRL-ID-126455, Rev. 1, November, 1997) and MCNP 4B (Briesmeister (Ed.), Version 4B, La-12625-m, March 1997, Los Alamos National Laboratory): two Monte Carlo codes for neutrons and photons. The quantities analyzed include: energy deposited by neutrons and gamma photons, values of the total fluence integrated on the whole
Electron Accelerator Shielding Design of KIPT Neutron Source Facility
Directory of Open Access Journals (Sweden)
Zhaopeng Zhong
2016-06-01
Full Text Available The Argonne National Laboratory of the United States and the Kharkov Institute of Physics and Technology of the Ukraine have been collaborating on the design, development and construction of a neutron source facility at Kharkov Institute of Physics and Technology utilizing an electron-accelerator-driven subcritical assembly. The electron beam power is 100 kW using 100-MeV electrons. The facility was designed to perform basic and applied nuclear research, produce medical isotopes, and train nuclear specialists. The biological shield of the accelerator building was designed to reduce the biological dose to less than 5.0e-03 mSv/h during operation. The main source of the biological dose for the accelerator building is the photons and neutrons generated from different interactions of leaked electrons from the electron gun and the accelerator sections with the surrounding components and materials. The Monte Carlo N-particle extended code (MCNPX was used for the shielding calculations because of its capability to perform electron-, photon-, and neutron-coupled transport simulations. The photon dose was tallied using the MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is very small, ∼0.01 neutron for 100-MeV electron and even smaller for lower-energy electrons. This causes difficulties for the Monte Carlo analyses and consumes tremendous computation resources for tallying the neutron dose outside the shield boundary with an acceptable accuracy. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were utilized for this study. The generated neutrons were banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron dose. The weight windows variance reduction technique was also utilized for both
Electron accelerator shielding design of KIPT neutron source facility
Energy Technology Data Exchange (ETDEWEB)
Zhong, Zhao Peng; Gohar, Yousry [Argonne National Laboratory, Argonne (United States)
2016-06-15
The Argonne National Laboratory of the United States and the Kharkov Institute of Physics and Technology of the Ukraine have been collaborating on the design, development and construction of a neutron source facility at Kharkov Institute of Physics and Technology utilizing an electron-accelerator-driven subcritical assembly. The electron beam power is 100 kW using 100-MeV electrons. The facility was designed to perform basic and applied nuclear research, produce medical isotopes, and train nuclear specialists. The biological shield of the accelerator building was designed to reduce the biological dose to less than 5.0e-03 mSv/h during operation. The main source of the biological dose for the accelerator building is the photons and neutrons generated from different interactions of leaked electrons from the electron gun and the accelerator sections with the surrounding components and materials. The Monte Carlo N-particle extended code (MCNPX) was used for the shielding calculations because of its capability to perform electron-, photon-, and neutron-coupled transport simulations. The photon dose was tallied using the MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is very small, ∼0.01 neutron for 100-MeV electron and even smaller for lower-energy electrons. This causes difficulties for the Monte Carlo analyses and consumes tremendous computation resources for tallying the neutron dose outside the shield boundary with an acceptable accuracy. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were utilized for this study. The generated neutrons were banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron dose. The weight windows variance reduction technique was also utilized for both neutron and photon dose
Utsuro, Masahiko
2010-01-01
Written by authors with an international reputation, acknowledged expertise and teaching experience, this is the most up-to-date resource on the field. The text is clearly structured throughout so as to be readily accessible, and begins by looking at scattering of a scalar particle by one-dimensional systems. The second section deals with the scattering of neutrons with spin in one-dimensional potentials, while the third treats dynamical diffraction in three-dimensional periodic media. The final two sections conclude with incoherent and small angle scattering, and some problems of quantum mech
Christl, Mark; Dobson, Chris; Norwood, Joseph; Kayatin, Matthew; Apple, Jeff; Gibson, Brian; Dietz, Kurt; Benson, Carl; Smith, Dennis; Howard, David;
2013-01-01
Energetic neutron measurements remain a challenge for space science investigations and radiation monitoring for human exploration beyond LEO. We are investigating a new composite scintillator design that uses Li6 glass scintillator embedded in a PVT block. A comparison between Li6 and Boron 10 loaded scintillators are being studied to assess the advantages and shortcomings of these two techniques. We present the details of the new Li6 design and results from the comparison of the B10 and Li6 techniques during exposures in a mixed radiation field produced by high energy protons interacting in a target material.
Lopes, Luiz L
2014-01-01
In this work we review the formalism normally used in the literature about the effects of density-dependent magnetic fields on the properties of neutron stars, expose some ambiguities that arise and propose a way to solve the related problem. Our approach uses a different prescription for the calculation of the pressure based on the chaotic field formalism for the stress tensor and also a different way of introducing a variable magnetic field, which depends on the energy density rather than on the baryonic density.
The Muon-Induced Neutron Indirect Detection EXperiment, MINIDEX
Abt, I; Carissimo, C; Gooch, C; Kneissl, R; Langford, J; Liu, X; Majorovits, B; Palermo, M; Schulz, O; Vanhoefer, L
2016-01-01
A new experiment to quantitatively measure neutrons induced by cosmic-ray muons in selected high-Z materials is introduced. The design of the Muon-Induced Neutron Indirect Detection EXperiment, MINIDEX, and the results from its first data taking period are presented as well as future plans. Neutron production in high-Z materials is of particular interest as such materials are used for shielding in low-background experiments. The design of next-generation large-scale experiments searching for neutrinoless double beta decay or direct interactions of dark matter requires reliable Monte Carlo simulations of background induced by muon interactions. The first five months of operation already provided a valuable data set on neutron production and neutron transport in lead. A first round of comparisons between MINIDEX data and Monte Carlo predictions obtained with two GEANT4- based packages is presented. The rate of muon-induced events is overall a factor three to four higher in data than predicted by the Monte Carlo...
High Energy Telescope With Neutron Detection Capabilities (HETn)
Posner, A.; Wimmer-Schweingruber, R. F.; Böhm, E.; Böttcher, s.; Connell, J. J.; Dröge, W.; Hassler, D. M.; Heber, B.; Lopate, C.; McKibben, R. B.; Steigies, C. T.
2007-01-01
The High-Energy Telescope with neutron detection capabilities (HETn) for the Solar Orbiter will measure and resolve energetic charged particles, in particular electrons, proton, and heavy ions up to Fe including selected isotopes up to energies equivalen to the penetration depth of 100 MeV protons. The full active anti-coincidence encloses detectors sensitive to 1-30 MeV neutrons and 0.5-5 MeV X-/gamma-rays. The sensor consists of the angle-detecting inclined sensors (ADIS) solid-state detector detector telescope utilizing a shared calorimeter for total energy and X-/gamma-ray measurement. A separate plastic detector provides sensitivity to neutrons via the recoil process. HETn will open a new window on solar eruptive events with its neutron detection capability and allows determination of high-energy close to the Sun. Timing and spectral information on neutral particles (neutrons and X-/gamma rays ), on relativistic electrons and high-energy heavy ions will provide new insights into the processes which accelerate particles to high energies at the sun and into transport processes between the source and the spacecraft in the near-Sun environment.
Monte Carlo simulations of a D-T neutron generator shielding for landmine detection
Energy Technology Data Exchange (ETDEWEB)
Reda, A.M., E-mail: amreda2005@yahoo.com [College of Science, Shaqra University, Al-Dawadme, P.O. Box 1040 (Saudi Arabia)
2011-10-15
Shielding for a D-T sealed neutron generator has been designed using the MCNP5 Monte Carlo radiation transport code. The neutron generator will be used in field for the detection of explosives, landmines, drugs and other 'threat' materials. The optimization of the detection of buried objects was started by studying the signal-to-noise ratio for different geometric conditions. - Highlights: > A landmine detection system based on neutron fast/slow analysis has been designed. > Shielding for a D-T sealed neutron generator tube has been designed using Monte Carlo radiation transport code. > Detection of buried objects was started by studying the signal-to-noise ratio for different geometric conditions. > The signal-to-background ratio optimized at one position for all depths.