WorldWideScience

Sample records for whole-core neutron transport

  1. A whole-core coarse mesh neutron transport method in 2D cylindrical (R, theta) geometry

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Dingkang; Rahnema, Farzad, E-mail: farzad@gatech.edu

    2013-12-15

    Highlights: • The coarse mesh radiation transport method COMET is extended to 2D cylindrical geometry. • Two new functions are also developed to facilitate the extension of COMET. • These functions are used to expand the incident angular flux on the (r, Θ) surfaces. • The new method has accuracy close to direct Monte Carlo with 3 orders of magnitude faster speed. - Abstract: In this paper, the hybrid stochastic deterministic coarse mesh radiation transport method COMET is extended to 2D cylindrical geometry for whole-core neutronics calculations. To facilitate the extension new functions in cylindrical geometry were developed for expanding the incident angular flux on the (r, Θ) surfaces. The new COMET method is tested in a set of simplified pebble bed reactor cores in 2D (r, Θ) geometry, consisting of an inner reflector, an annular fuel region, a controlled outer reflector. The comparisons indicate that COMET achieves an accuracy close to the Monte Carlo method (MCNP), with computational efficiency for core calculations that is at least 3 orders of magnitude faster than that of MCNP when a precomputed response library with a 4th order expansion in space and a 2nd order expansion in the polar and azimuthal angles is used.

  2. Prospects in deterministic three dimensional whole-core transport calculations

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, Richard [Commissariat a l' Energie Atomique et aux Energies Alternatives Direction de l' Energie Nucleaire, Service d.Etudes de Reacteurs et de Mathematiques Appliquees CEA de Saclay, Cedex (France)

    2012-03-15

    The point we made in this paper is that, although detailed and precise three-dimensional (3D) whole-core transport calculations may be obtained in the future with massively parallel computers, they would have an application to only some of the problems of the nuclear industry, more precisely those regarding multiphysics or for methodology validation or nuclear safety calculations. On the other hand, typical design reactor cycle calculations comprising many one-point core calculations can have very strict constraints in computing time and will not directly benefit from the advances in computations in large scale computers. Consequently, in this paper we review some of the deterministic 3D transport methods which in the very near future may have potential for industrial applications and, even with low-order approximations such as a low resolution in energy, might represent an advantage as compared with present industrial methodology, for which one of the main approximations is due to power reconstruction. These methods comprise the response-matrix method and methods based on the two-dimensional (2D) method of characteristics, such as the fusion method.

  3. Verification of Cell-Homogenized Whole-Core Transport Calculations in Actual PWR Core Geometry

    Energy Technology Data Exchange (ETDEWEB)

    Akio Yamamoto; Masahiro Tatsumi; Tatsuya Kimoto; Shinya Kosaka; Etsuro Saji

    2000-11-12

    In this paper, a cell-homogenized whole-core transport calculation is applied to a pressurized water reactor (PWR) core and its prediction accuracy on reactivity and power distribution is verified in comparison with cell-heterogeneous whole-core transport calculation results obtained by the method of characteristics. Recent progress in common computer hardware promotes continuous improvements on core calculation methods. Among several different approaches, a three-dimensional pin-by-pin (fine-mesh) core calculation is considered as one of the candidates of advanced methods. In general, although the fine-mesh core calculation explicitly treats each individual pin, it still has three approximations: cell-homogenization, low-level transport theory (usually the SPN or the diffusion approximation) and few-group treatment. Therefore, the impact of these approximations on core characteristics should be evaluated to confirm the feasibility of the three-dimensional fine-mesh core calculation method. This paper focuses on two former approximations, and the impact of these approximations is evaluated.

  4. 3-D Whole-Core Transport Calculation with 3D/2D Rotational Plane Slicing Method

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Han Jong; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-10-15

    Use of the method of characteristics (MOC) is very popular due to its capability of heterogeneous geometry treatment and widely used for 2-D core calculation, but direct extension of MOC to 3-D core is not so attractive due to huge calculational cost. 2-D/1-D fusion method was very successful for 3-D calculation of current generation reactor types (highly heterogeneous in radial direction but piece-wise homogeneous in axial direction). In this paper, 2-D MOC concept is extended to 3-D core calculation with little modification of an existing 2-D MOC code. The key idea is to suppose 3-D geometry as a set of many 2-D planes like a phone-directory book. Dividing 3-D structure into a large number of 2-D planes and solving each plane with a simple 2-D SN transport method would give the solution of a 3-D structure. This method was developed independently at KAIST but it is found that this concept is similar with that of 'plane tracing' in the MCCG-3D code. The method developed was tested on the 3-D C5G7 OECD/NEA benchmark problem and compared with the 2-D/1-D fusion method. Results show that the proposed method is worth investigating further. A new approach to 3-D whole-core transport calculation is described and tested. By slicing 3-D structure along characteristic planes and solving each 2-D plane problem, we can get 3-D solution. The numerical test results indicate that the new method is comparable with the 2D/1D fusion method and outperforms other existing methods. But more fair comparison should be done in similar discretization level.

  5. Verification of the Advanced Nodal Method on BWR Core Analyses by Whole-Core Heterogeneous Transport Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Shinya Kosaka

    2000-11-12

    Recent boiling water reactor (BWR) core and fuel designs have become more sophisticated and heterogeneous to improve fuel cycle cost, thermal margin, etc. These improvements, however, tend to lead to a strong interference effect among fuel assemblies, and it my cause some inaccuracies in the BWR core analyses by advanced nodal codes. Furthermore, the introduction of mixed-oxide (MOX) fuel will lead to a much stronger interference effect between MOX and UO{sub 2} fuel assemblies. However, the CHAPLET multiassembly characteristics transport code was developed recently to solve two-dimensional cell-heterogeneous whole-core problems efficiently, and its results can be used as reference whole-core solutions to verify the accuracy of nodal core calculations. In this paper, the results of nodal core calculations were compared with their reference whole-core transport solutions to verify their accuracy (in k{sub eff}, assembly power and pin power via pin power reconstruction) of the advanced nodal method on both UO{sub 2} and MOX BWR whole-core analyses. Especially, it was investigated if there were any significant differences in the accuracy between MOX and UO{sub 2} results.

  6. Whole core pin-by-pin coupled neutronic-thermal-hydraulic steady state and transient calculations using COBAYA3 code

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, J.; Herrero, J. J.; Cuervo, D.; Aragones, J. M., E-mail: jimenez@din.upm.e [Universidad Politecnica de Madrid, Departamento de Ingenieria Nuclear, Jose Gutierrez Abascal No. 2, 28006 Madrid (Spain)

    2010-10-15

    Nowadays coupled 3-dimensional neutron kinetics and thermal-hydraulic core calculations are performed by applying a radial average channel approach using a meshing of one quarter of assembly in the best case. This approach does not take into account the subchannels effects due to the averaging of the physical fields and the loose of heterogeneity in the thermal-hydraulic model. Therefore the models do not have enough resolution to predict those subchannels effects which are important for the fuel design safety margins, because it is in the local scale, where we can search the hottest pellet or the maximum heat flux. The Polytechnic University of Madrid advanced multi-scale neutron-kinetics and thermal-hydraulics methodologies being implemented in COBAYA3 include domain decomposition by alternate core dissections for the local 3-dimensional fine-mesh scale problems (pin cells/subchannels) and an analytical nodal diffusion solver for the coarse mesh scale coupled with the thermal-hydraulic using a model of one channel per assembly or per quarter of assembly. In this work, we address the domain decomposition by the alternate core dissections methodology applied to solve coupled 3-dimensional neutronic-thermal-hydraulic problems at the fine-mesh scale. The neutronic-thermal-hydraulic coupling at the cell-subchannel scale allows the treatment of the effects of the detailed thermal-hydraulic feedbacks on cross-sections, thus resulting in better estimates of the local safety margins at the pin level. (Author)

  7. A stylized three dimensional PWR whole-core benchmark problem with Gadolinium

    Energy Technology Data Exchange (ETDEWEB)

    Douglass, Steven, E-mail: douglass.steven@gmail.co [Nuclear and Radiological Engineering/Medical Physics Programs, George W. Woodruff School, Georgia Institute of Technology, Atlanta, GA 30332-0405 (United States); Rahnema, Farzad, E-mail: farzad@gatech.ed [Nuclear and Radiological Engineering/Medical Physics Programs, George W. Woodruff School, Georgia Institute of Technology, Atlanta, GA 30332-0405 (United States); Margulies, Johann, E-mail: johann.margulies@me.gatech.ed [Nuclear and Radiological Engineering/Medical Physics Programs, George W. Woodruff School, Georgia Institute of Technology, Atlanta, GA 30332-0405 (United States)

    2010-10-15

    Recent advancements in computer technology have led to an increased ability to solve complicated reactor problems, and as a result, there has been an interest in codes and methods that utilize whole-core transport in three-dimensional, large-scale, highly-heterogeneous configurations. Most current reactor benchmark problems are either small-scale or involve spatial homogenization (pin cell or full-assembly) for diffusion codes; therefore, new whole-core benchmark problems must be designed in order to validate whole-core transport methods. This study focuses on an PWR case, with a low leakage loading pattern and Gadolinium as a burnable absorber. In this paper, a detailed description of a new 3D whole-core numerical benchmark problem based on a 4-loop layout is presented with geometric, material and cross section specifications. A set of 2-group and 4-group region-wise neutron cross section libraries is generated using the lattice depletion code HELIOS, and an MCNP model is described for three configurations (all rods in, all rods out, power-shaping rods inserted). Eigenvalues are presented for the 2-group calculations of all three configurations, and detailed pin fission density results are presented for the power-shaping configuration.

  8. Stability and accuracy of 3D neutron transport simulations using the 2D/1D method in MPACT

    Science.gov (United States)

    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.

  9. 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.

  10. Influence of the neutron transport tube on neutron resonance densitometry

    Science.gov (United States)

    Kitatani, Fumito; Tsuchiya, Harufumi; Koizumi, Mitsuo; Takamine, Jun; Hori, Junichi; Sano, Tadafumi

    2017-09-01

    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.

  11. Coupled Neutron Transport for HZETRN

    Science.gov (United States)

    Slaba, Tony C.; Blattnig, Steve R.

    2009-01-01

    Exposure estimates inside space vehicles, surface habitats, and high altitude aircrafts exposed to space radiation are highly influenced by secondary neutron production. The deterministic transport code HZETRN has been identified as a reliable and efficient tool for such studies, but improvements to the underlying transport models and numerical methods are still necessary. In this paper, the forward-backward (FB) and directionally coupled forward-backward (DC) neutron transport models are derived, numerical methods for the FB model are reviewed, and a computationally efficient numerical solution is presented for the DC model. Both models are compared to the Monte Carlo codes HETC-HEDS, FLUKA, and MCNPX, and the DC model is shown to agree closely with the Monte Carlo results. Finally, it is found in the development of either model that the decoupling of low energy neutrons from the light particle transport procedure adversely affects low energy light ion fluence spectra and exposure quantities. A first order correction is presented to resolve the problem, and it is shown to be both accurate and efficient.

  12. Neutron Transport Simulations for NIST Neutron Lifetime Experiment

    Science.gov (United States)

    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.

  13. 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.

  14. An improved neutron transport algorithm for HZETRN

    Science.gov (United States)

    Slaba, T. C.; Blattnig, S. R.; Clowdsley, M. S.; Walker, S. A.; Badavi, F. F.

    2010-09-01

    Long-term human presence in space requires the inclusion of radiation constraints in mission planning and the design of shielding materials, structures and vehicles. It is necessary to expose the numerical tools commonly used in radiation analyses to extensive verification, validation and uncertainty quantification. In this paper, the numerical error associated with energy discretization in HZETRN is addressed. An inadequate numerical integration scheme in the transport algorithm is shown to produce large errors in the low energy portion of the neutron and light ion fluence spectra. It is further shown that the errors result from the narrow energy domain of the neutron elastic cross section spectral distributions and that an extremely fine energy grid is required to resolve the problem under the current formulation. Since adding a sufficient number of energy points will render the code computationally inefficient, we revisit the light ion and neutron transport theory developed for HZETRN and focus on neutron elastic interactions. Two numerical methods (average value and collocation) are developed to provide adequate resolution in the energy domain and more accurately resolve the neutron elastic interactions. An energy grid convergence study is conducted to demonstrate the improved stability of the new methods. Based on the results of the convergence study and the ease of implementation, the average value method with a 100 point energy grid is found to be suitable for future use in HZETRN.

  15. An Improved Neutron Transport Algorithm for HZETRN

    Science.gov (United States)

    Slaba, Tony C.; Blattnig, Steve R.; Clowdsley, Martha S.; Walker, Steven A.; Badavi, Francis F.

    2010-01-01

    Long term human presence in space requires the inclusion of radiation constraints in mission planning and the design of shielding materials, structures, and vehicles. In this paper, the numerical error associated with energy discretization in HZETRN is addressed. An inadequate numerical integration scheme in the transport algorithm is shown to produce large errors in the low energy portion of the neutron and light ion fluence spectra. It is further shown that the errors result from the narrow energy domain of the neutron elastic cross section spectral distributions, and that an extremely fine energy grid is required to resolve the problem under the current formulation. Two numerical methods are developed to provide adequate resolution in the energy domain and more accurately resolve the neutron elastic interactions. Convergence testing is completed by running the code for various environments and shielding materials with various energy grids to ensure stability of the newly implemented method.

  16. Cosmic-ray neutron transport at a forest field site

    DEFF Research Database (Denmark)

    Andreasen, Mie; Jensen, Karsten Høgh; Desilets, Darin

    2017-01-01

    -ray neutron intensity is essential (e.g., the effect of vegetation, litter layer and soil type). In this study the environmental effect is examined by performing a sensitivity analysis using neutron transport modeling. We use a neutron transport model with various representations of the forest and different...

  17. Multi-domain/multi-method numerical approach for neutron transport equation; Couplage de methodes et decomposition de domaine pour la resolution de l'equation du transport des neutrons

    Energy Technology Data Exchange (ETDEWEB)

    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.

  18. Orthogonal polynomials in neutron transport theory

    Energy Technology Data Exchange (ETDEWEB)

    Dehesa, J.S. (Granada Univ. (Spain). Facultad de Ciencias)

    1982-01-01

    The asymptotic average properties of zeros of the polynomials gsub(k)sup(m) (x), which play a fundamental role in neutron transport and radiative transfer theories, are investigated analytically in terms of the angular expansion coefficients wsub(k) of the scattering kernel for three wide classes of scattering models. In particular it is found that the scattering models of Eccleston-McCormick (J. Nucl. Energy.; 24:23 (1970)), Shultis et al (Nucl. Sci. Eng.; 59:53 (1976)) and Henyey-Greenstein (Astrophys. J.; 93:70 (1941)) belong in one of the above-mentioned classes, and their associated polynomials gsub(k)sup(m) (x) have the same asymptotic density of zeros.

  19. Whole Core Transport Calculation Methodology for a Hexagonal Core

    Energy Technology Data Exchange (ETDEWEB)

    Cho, J. Y.; Kim, K. S.; Lee, C. C.; Zee, S. Q.; Joo, H. G

    2007-07-15

    This report discusses the hexagonal module implemented to the DeCART code and the performance of them. The implemented hexagonal module includes the hexagonal ray tracing and the CMFD acceleration modules. The performance of the implemented hexagonal module is examined for 4 tests of: (1) CMFD acceleration test, (2) the accuracy test of the hexagonal module, (3) the performance test for 2-D NGNP problem and (4) the applicability test for 3-D NGNP problem. The features of the implemented hexagonal modules are: (1) The Modular ray tracing scheme based on a hexagonal assembly and a path linking scheme between the modular rays. (2) Segment generation based on the structure unit. (3) Cell ray approximation: This feature is developed to reduce the memory required to store the segment information. (4) Modified cycle ray scheme that begins the ray tracing at a given surface and finishes if the reflected ray meets the starting surface. This feature is developed to reduce the memory required for the angular flux at the core boundary. (5) Fixed assembly geometry. The pin geometry of the single pin per assembly problem is different from that of the multi-pin problem. The core geometry of a single assembly problem is also different from that of the multi-assembly problem. (6) CMFD module based on unstructured cell. This feature is to deal with the irregular gap cells that are positioned at the assembly boundaries. The examination results of the 4 tests can be summarized as: (1) The CMFD acceleration test shows that the CMFD module speedups about greater than 200 for the core problem. (2) The accuracy test shows that the hexagonal MOC module produces an accurate solution of less than 60 pcm of eigenvalue and less than 2 % of local pin power errors. (3) The performance test for 2-D NGNP problem shows that the implemented hexagonal module works soundly and produces a reasonable solution by cooperating with the existing DeCART library and the other modules. (4) The applicability test for 3-D NGNP problem shows that DeCART can be applied to the real 3-D hexagonal core problem by using the MPI based parallel computation. These test results mean that the hexagonal module is well developed and well implemented to the DeCART code, and cooperates with the existing DeCART modules. Therefore, it is concluded that the DeCART code contains a practical capability to solve a realistic hexagonal core problem within an affordable computing time.

  20. Magnetic field devices for neutron spin transport and manipulation in precise neutron spin rotation measurements

    Energy Technology Data Exchange (ETDEWEB)

    Maldonado-Velázquez, M. [Posgrado en Ciencias Físicas, Universidad Nacional Autónoma de México, 04510 (Mexico); Barrón-Palos, L., E-mail: libertad@fisica.unam.mx [Instituto de Física, Universidad Nacional Autónoma de México, Apartado Postal 20-364, 01000 (Mexico); Crawford, C. [University of Kentucky, Lexington, KY 40506 (United States); Snow, W.M. [Indiana University, Bloomington, IN 47405 (United States)

    2017-05-11

    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{sup −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.

  1. Composite finite element solutions for neutron transport

    Energy Technology Data Exchange (ETDEWEB)

    Ackroyd, R.T.; Wilson, W.E.

    1988-01-01

    Composite solutions for neutron transport employ a finite element representation for a spatial dependence of the angular flux and spherical-harmonic expansions for its directional dependence. The orders of the spherical harmonic expansions are varied from region to region, according to the physical characteristics of the regions, so that the massive calculations inherent with high-order expansions can be confined to the regions for which they are essential. The transitions from low- to high-order expansions are made at the interfaces of elements. The K/sup +-/ boundary-free maximum principle provides a convenient means of optimizing a composite approximate solution for the angular flux. The way in which penalty weighting of interfaces is used to optimize composite solutions is investigated theoretically and numerically. The method is illustrated by solutions of two benchmark problems. An estimate for penalty weightings is obtained by deriving a central difference analogue of the even-parity transport equation. This provides also a demonstration of the use of admissible, wildly nonconforming elements in a maximum principle.

  2. 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.

  3. Technical notes. Spherical harmonics approximations of neutron transport

    Energy Technology Data Exchange (ETDEWEB)

    Demeny, A.; Dede, K.M.; Erdei, K.

    1976-12-01

    A double-range spherical harmonics approximation obtained by expanding the angular flux separately in the two regions combined with the conventional single-range spherical harmonics is found to give superior description of neutron transport.

  4. 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.

  5. Foundations of finite element applications to neutron transport

    Energy Technology Data Exchange (ETDEWEB)

    Ackroyd, R.T. [Imperial Coll. of Science and Technology, London (United Kingdom). Dept. of Mechanical Engineering

    1995-06-01

    The use of finite elements in neutron transport gives deterministic treatments a geometrical capability with a flexibility comparable to that achievable by the stochastic Monte Carlo method. The principles used in obtaining solutions can be used also to assess the accuracy of numerical solutions. These assessment methods are illustrated by close bracketing bounds for local error and disadvantage factors. Principles are described for the steady state which guarantee neutron conservation for the whole finite element mesh. Some new principles are described which guarantee neutron conservation over each element of a mesh. For time dependent transport a principle guarantees neutron conservation with time over the whole mesh. The paper concludes with a sketch of fundamental work with the potential to exploit recent developments in computer architecture. (author).

  6. Transportable, Low-Dose Active Fast-Neutron Imaging

    Energy Technology Data Exchange (ETDEWEB)

    Mihalczo, John T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wright, Michael C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McConchie, Seth M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Archer, Daniel E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Palles, Blake A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    This document contains a description of the method of transportable, low-dose active fast-neutron imaging as developed by ORNL. The discussion begins with the technique and instrumentation and continues with the image reconstruction and analysis. The analysis discussion includes an example of how a gap smaller than the neutron production spot size and detector size can be detected and characterized depending upon the measurement time.

  7. 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)

  8. Considerations in the design of an improved transportable neutron spectrometer

    CERN Document Server

    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.

  9. Optimization study of a transportable neutron radiography unit based on a compact neutron generator

    Energy Technology Data Exchange (ETDEWEB)

    Fantidis, J.G. [Laboratory of Nuclear Technology, School of Engineering, ' Democritus' University of Thrace, Xanthi (Greece); Nicolaou, G.E., E-mail: nicolaou@ee.duth.g [Laboratory of Nuclear Technology, School of Engineering, ' Democritus' University of Thrace, Xanthi (Greece); Tsagas, N.F. [Laboratory of Nuclear Technology, School of Engineering, ' Democritus' University of Thrace, Xanthi (Greece)

    2010-06-21

    A transportable fast and thermal neutron radiography system, incorporating a compact DD neutron generator, has been simulated using the MCNPX code. 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. Appropriate collimators were simulated for each of the radiography modes. With suitable aperture and collimator designs, it was possible to optimize the parameters for both fast and thermal neutron radiographies, for a wide range of values of the collimator ratio. The system simulated allows different object sizes to be studied with a wide range of radiography parameters.

  10. Neutron Transport Associated with the Galactic Cosmic Ray Cascade

    Science.gov (United States)

    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

  11. Parallel solution of the neutron multigroup transport equation; Resolution parallele de l`equation du transport multigroupe des neutrons

    Energy Technology Data Exchange (ETDEWEB)

    N`kaoua, T.; Chaigneau, C.; Coulomb, F. [CEA Centre d`Etudes de Limeil, 94 - Villeneuve-Saint-Georges (France)

    1997-04-01

    We are interested in the solution of the Multigroup Neutron Transport Equation. After the presentation of the multigroup treatment, the angular, time and space discretization, we expose the modifications that have been made in order to get an efficient parallel method. (author)

  12. 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.

  13. Homogenization of the critically spectral equation in neutron transport

    Energy Technology Data Exchange (ETDEWEB)

    Allaire, G. [CEA Saclay, 91 - Gif-sur-Yvette (France). Dept. de Mecanique et de Technologie]|[Paris-6 Univ., 75 (France). Lab. d' Analyse Numerique; Bal, G. [Electricite de France (EDF), 92 - Clamart (France). Direction des Etudes et Recherches

    1998-07-01

    We address the homogenization of an eigenvalue problem for the neutron transport equation in a periodic heterogeneous domain, modeling the criticality study of nuclear reactor cores. We prove that the neutron flux, corresponding to the first and unique positive eigenvector, can be factorized in the product of two terms, up to a remainder which goes strongly to zero with the period. On terms is the first eigenvector of the transport equation in the periodicity cell. The other term is the first eigenvector of a diffusion equation in the homogenized domain. Furthermore, the corresponding eigenvalue gives a second order corrector for the eigenvalue of the heterogeneous transport problem. This result justifies and improves the engineering procedure used in practice for nuclear reactor cores computations. (author)

  14. Recent improvements to a transportable neutron spectrometer (TNS)

    CERN Document Server

    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...

  15. Bicubic spline function approximation of the solution of the fast neutron transport equation

    CERN Document Server

    Chamayou, J M F

    1975-01-01

    The numerical method of approximation of the fast neutron stationary transport equation by means of bicubic cardinal splines is investigated in order to calculate the neutron flux in the one- dimensional plane geometry. A numerical example is given. (5 refs).

  16. Neutrons for fuel cell membranes: Structure, sorption and transport properties

    Science.gov (United States)

    Lyonnard, S.; Gebel, G.

    2012-11-01

    A molecular level understanding of structure and transport properties in fuel cell ionomer membranes is essential for designing new electrolytes with improved performance. Scattering techniques are suited tools for this purpose. In particular, neutron scattering, which has been extensively used in hydrogen-containing systems, is well adapted to investigate water-dependent complex polymeric morphologies. We report Small-Angle Neutron Scattering (SANS) studies on different types of fuel cell polymers: perfluorinated, radiation-grafted and sulfonated polyphosphazene membranes. We show that contrast variation methods can be efficiently employed to provide new insights on membrane microstructure and reveal ionic condensation effects. Neutrons have been used also as non-intrusive diagnosis tool to probe water properties and distribution inside membranes. Recently, in-situ neutronography and SANS experiments on operating fuel cells have been reported. In-plane cartography of water distribution at the surface of bipolar plates and water profiles across membrane thickness have been obtained and studied as a function of operating conditions. The last section of the article is devoted to the use of Quasi-Elastic Neutron Scattering to study water dynamics at molecular scale. We show that analysis with an appropriate sophisticated diffusion model allows to extract diffusion coefficients, characteristic times and length-scales of molecular motions. This quantitative information is fruitfully integrated in multi-scale modelling and usefully compared with numerical simulations. QENS also permits to compare alternative polymers and relate dynamical properties to chemical composition and membrane nanostructure.

  17. A response-based time-dependent neutron transport method

    Energy Technology Data Exchange (ETDEWEB)

    Pounders, Justin M. [Nuclear and Radiological Engineering/Medical Physics Programs, George W. Woodruff School, Georgia Institute of Technology, Atlanta, GA 30332-0405 (United States); Rahnema, Farzad, E-mail: farzad@gatech.ed [Nuclear and Radiological Engineering/Medical Physics Programs, George W. Woodruff School, Georgia Institute of Technology, Atlanta, GA 30332-0405 (United States)

    2010-11-15

    An efficient response-based solution to the time-dependent neutron transport equation in a semi-infinite slab is derived. The solution is based on polynomial expansions of the source terms and neutron flux in the time domain. The expansion coefficients of the flux solution are computed in terms of response functions, which are special cases of Green's functions for arbitrary in-volume and surface sources. The resulting response equation, which is a convolution integral equation in time, is reduced to a linear algebraic system of equations in the expansion coefficients. Two example problems are solved using the response-based method, and the extension of the method to general (finite, heterogeneous) problems is discussed.

  18. In situ quantification and visualization of lithium transport with neutrons.

    Science.gov (United States)

    Liu, Danny X; Wang, Jinghui; Pan, Ke; Qiu, Jie; Canova, Marcello; Cao, Lei R; Co, Anne C

    2014-09-01

    A real-time quantification of Li transport using a nondestructive neutron method to measure the Li distribution upon charge and discharge in a Li-ion cell is reported. By using in situ neutron depth profiling (NDP), we probed the onset of lithiation in a high-capacity Sn anode and visualized the enrichment of Li atoms on the surface followed by their propagation into the bulk. The delithiation process shows the removal of Li near the surface, which leads to a decreased coulombic efficiency, likely because of trapped Li within the intermetallic material. The developed in situ NDP provides exceptional sensitivity in the temporal and spatial measurement of Li transport within the battery material. This diagnostic tool opens up possibilities to understand rates of Li transport and their distribution to guide materials development for efficient storage mechanisms. Our observations provide important mechanistic insights for the design of advanced battery materials. © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  19. Structures of the fractional spaces generated by the difference neutron transport operator

    Energy Technology Data Exchange (ETDEWEB)

    Ashyralyev, Allaberen [Department of Elementary Mathematics Education, Fatih University, 34500, Istanbul (Turkey); Department of Mathematics, ITTU, Ashgabat (Turkmenistan); Taskin, Abdulgafur [Department of Mathematics, Fatih University, 34500, Istanbul (Turkey)

    2015-09-18

    The initial boundary value problem for the neutron transport equation is considered. The first, second and third order of accuracy difference schemes for the approximate solution of this problem are presented. Highly accurate difference schemes for neutron transport equation based on Padé approximation are constructed. In applications, stability estimates for solutions of difference schemes for the approximate solution of the neutron transport equation are obtained.The positivity of the neutron transport operator in Slobodeckij spaces is proved. Numerical techniques are developed and algorithms are tested on an example in MATLAB.

  20. High order variational solutions of time dependent neutron transport problems

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, B.C.

    1985-01-01

    High order numerical solutions of the time-dependent one speed neutron transport equation are developed using cubic hermite polynomial trial functions, variational techniques, and exponential matrix operators. Two new numerical solutions are developed that are high order with respect to both time and space variables. In the first method, the time-dependent P/sub N/ equations are transformed into Generalized Telegrapher's Equations (GTE) that are valid for any order P/sub N/ approximation. The Generalized Telegrapher's Equations form a coupled set of second order differential equations with respect to both time and space. In the second method, the time-dependent P/sub N/ equations are transformed into coupled Transport Diffusion Equations (TDE), keeping the additional terms that maintain the transport nature of the solution. The Transport Diffusion Equations are first order in time and second order in space. Numerically evaluated time-dependent analytic solutions are also developed for homogeneous media transport problems in the P/sub 1/ and P/sub 3/ approximations via Laplace Transforms in order to validate the variational GTE and TDE solutions. The analytic solutions allow anisotropic scattering, up to the appropriate P/sub N/ order. The analytic solutions are not limited to the non-precise extrapolation boundary condition, like many time-dependent analytic P/sub N/ solutions, but allow any of the standard transport vacuum boundary condition approximations.

  1. Commercial exploitation of finite element codes for neutron transport

    Energy Technology Data Exchange (ETDEWEB)

    Ackroyd, R.T.; Issa, J.G.

    1989-01-01

    The finite element method for neutron transport is in commercial use in the United Kingdom for shielding and criticality problems. It is also applied to time-dependent problems arising in the field of oil-well logging. The present capability of the method is described, and the developments in hand to accelerate calculations are outlined. The codes that have been developed in the United Kingdom are FELTRAN, TRIPAC, and MARC. Complex shapes can be modeled with finite elements as realistically as in Monte Carlo calculations. Thus the finite element and Monte Carlo methods provide an independent means of cross checking a safety assessment for finalized design. The finite element method is particularly useful for scoping design studies, because it gives the angular fluxes everywhere in the system. These fluxes provide guidance on the efficacy of the design features.

  2. 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...

  3. 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)

  4. A method for transient, three-dimensional neutron transport calculations

    Energy Technology Data Exchange (ETDEWEB)

    Waddell, M.W. Jr. (Oak Ridge Y-12 Plant, TN (United States)); Dodds, H.L. (Tennessee Univ., Knoxville, TN (United States))

    1992-12-28

    This paper describes the development and evaluation of a method for solving the time-dependent, three-dimensional Boltzmann transport model with explicit representation of delayed neutrons. A hybrid stochastic/deterministic technique is utilized with a Monte Carlo code embedded inside of a quasi-static kinetics framework. The time-dependent flux amplitude, which is usually fast varying, is computed deterministically by a conventional point kinetics algorithm. The point kinetics parameters, reactivity and generation time as well as the flux shape, which is usually slowly varying in time, are computed stochastically during the random walk of the Monte Carlo calculation. To verify the accuracy of this new method, several computational benchmark problems from the Argonne National Laboratory benchmark book, ANL-7416, were calculated. The results are shown to be in reasonably good agreement with other independently obtained solutions. The results obtained in this work indicate that the method/code is working properly and that it is economically feasible for many practical applications provided a dedicated high performance workstation is available.

  5. A method for transient, three-dimensional neutron transport calculations

    Energy Technology Data Exchange (ETDEWEB)

    Waddell, M.W. Jr. (Martin Marietta Energy Systems, Inc. (United States)); Dodds, H.L. (Univ. of Tennessee (United States))

    1993-04-01

    This paper describes the development and evaluation of a method for solving the time-dependent, three-dimensional Boltzmann transport model with explicit representation of delayed neutrons. A hybrid stochastic/deterministic technique is utilized with a Monte Carlo code embedded inside of a quasi-static kinetics framework. The time-dependent flux amplitude, which is usually fast varying, is computed deterministically by a conventional point kinetics algorithm. The point kinetics parameters, reactivity and generation time as well as the flux shape, which is usually slowly varying in time, are computed stochastically during the random walk of the Monte Carlo calculation. To verify the accuracy of this new method, several computational benchmark problems from the Argonne National Laboratory benchmark book, ANL-7416, were calculated. The results are shown to be in reasonably good agreement with other independently obtained solutions. The results obtained in this work indicate that the method/code is working properly and that it is economically feasible for many practical applications provided a dedicated high performance workstation is available. (orig.)

  6. 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.)

  7. Cooperative learning of neutron diffusion and transport theories

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, Michael A.

    1999-04-30

    A cooperative group instructional strategy is being used to teach a unit on neutron transport and diffusion theory in a first-year-graduate level, Reactor Theory course that was formerly presented in the traditional lecture/discussion style. Students are divided into groups of two or three for the duration of the unit. Class meetings are divided into traditional lecture/discussion segments punctuated by cooperative group exercises. The group exercises were designed to require the students to elaborate, summarize, or practice the material presented in the lecture/discussion segments. Both positive interdependence and individual accountability are fostered by adjusting individual grades on the unit exam by a factor dependent upon group achievement. Group collaboration was also encouraged on homework assignments by assigning each group a single grade on each assignment. The results of the unit exam have been above average in the two classes in which the cooperative group method was employed. In particular, the problem solving ability of the students has shown particular improvement. Further,the students felt that the cooperative group format was both more educationally effective and more enjoyable than the lecture/discussion format.

  8. Two level calculation of assembly neutronic data libraries; Schema de calcul de bibliotheques a deux niveaux

    Energy Technology Data Exchange (ETDEWEB)

    Benomar, M

    1998-09-01

    The neutronic modeling of a nuclear reactor core requires 2 steps. The first step that is called transport calculation, is an accurate modeling of each type of assemblies put in a simple configuration. APOLLO2, a French neutronic code is used. This step allows the constitution of assembly data libraries. The second step represents the computing of the whole core by the diffusion theory and by using the data libraries defined in the first step. This work is dedicated to the improvement of the first step by allowing both a 172 group energy meshing and a two-dimension spatial processing. (A.C.) 7 refs.

  9. Modeling of the GELINA neutron target using coupled electron-photon-neutron transport with the MCNP4C3 code

    Energy Technology Data Exchange (ETDEWEB)

    Flaska, M.; Borella, A.; Lathouwers, D.; Mihailescu, L.C.; Mondelaers, W. E-mail: willy.mondelaers@cec.eu.int; Plompen, A.J.M.; Dam, H. van; Hagen, T.H.J.J. van der

    2004-10-01

    Coupled electron-photon-neutron transport calculations were performed with the Monte Carlo code MCNP4C3 to model the GELINA neutron target. Our calculations were compared with measurements of the direct and moderated flux distributions and agreement was achieved within 20%. Further, the present simulations were compared with the resolution functions established earlier with a dedicated program and with measured data for {sup 56}Fe. Here too, the agreement is excellent. New IAEA photonuclear libraries were implemented to perform these calculations. The simulations crucially rely on the new MCNP4C3 photonuclear capabilities and are made efficient by using the variance reduction and TALLYX features.

  10. Neutron transport calculation for Activation Evaluation for Decommissioning of PET cyclotron Facility

    Science.gov (United States)

    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.

  11. Neutron transport calculation for Activation Evaluation for Decommissioning of PET cyclotron Facility

    Directory of Open Access Journals (Sweden)

    Nobuhara Fumiyoshi

    2017-01-01

    Full Text Available 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.

  12. Computational Transport Modeling of High-Energy Neutrons Found in the Space Environment

    Science.gov (United States)

    Cox, Brad; Theriot, Corey A.; Rohde, Larry H.; Wu, Honglu

    2012-01-01

    The high charge and high energy (HZE) particle radiation environment in space interacts with spacecraft materials and the human body to create a population of neutrons encompassing a broad kinetic energy spectrum. As an HZE ion penetrates matter, there is an increasing chance of fragmentation as penetration depth increases. When an ion fragments, secondary neutrons are released with velocities up to that of the primary ion, giving some neutrons very long penetration ranges. These secondary neutrons have a high relative biological effectiveness, are difficult to effectively shield, and can cause more biological damage than the primary ions in some scenarios. Ground-based irradiation experiments that simulate the space radiation environment must account for this spectrum of neutrons. Using the Particle and Heavy Ion Transport Code System (PHITS), it is possible to simulate a neutron environment that is characteristic of that found in spaceflight. Considering neutron dosimetry, the focus lies on the broad spectrum of recoil protons that are produced in biological targets. In a biological target, dose at a certain penetration depth is primarily dependent upon recoil proton tracks. The PHITS code can be used to simulate a broad-energy neutron spectrum traversing biological targets, and it account for the recoil particle population. This project focuses on modeling a neutron beamline irradiation scenario for determining dose at increasing depth in water targets. Energy-deposition events and particle fluence can be simulated by establishing cross-sectional scoring routines at different depths in a target. This type of model is useful for correlating theoretical data with actual beamline radiobiology experiments. Other work exposed human fibroblast cells to a high-energy neutron source to study micronuclei induction in cells at increasing depth behind water shielding. Those findings provide supporting data describing dose vs. depth across a water-equivalent medium. This

  13. 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.

  14. 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.

  15. Natural iron isotopes influence on the neutron transport.

    Science.gov (United States)

    Schulc, Martin; Jánský, Bohumil; Harutyunyan, Davit; Novák, Evžen

    2017-12-01

    As an iron is the main structural component of nuclear power plants as well as future fusion power plants, the validation of neutron incident data libraries of iron is a must. Presented paper fits into ongoing validation activities and presents measuring neutron leakage spectra in the 0.1-1.0MeV region from iron sphere of 100cm in diameter by hydrogen proportional detectors. The experimental result is compared with ENDF/B-VII.1, JEFF-3.2 and CIELO nuclear data libraries. No library reasonably well describes whole region under study. Furthermore, elastic and inelastic XS sensitivity analysis for all iron isotopes was carried out. 54Fe isotope elastic XS influence is comparable with 56Fe XS influence up to 0.8MeV. 57Fe isotope elastic XS is significant in the region of 0.14-0.15MeV. Additionally, there are large differences among libraries in both elastic and inelastic XS for 57Fe. Furthermore, it was found that 58Fe isotope XS has negligible influence on the results. As a neutron source, 252Cf with initial emission rate of 9.53E8 n/s was used in this experiment. Copyright © 2017 Elsevier Ltd. All rights reserved.

  16. Quantifying moisture transport in cementitious materials using neutron radiography

    Science.gov (United States)

    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

  17. Sensitive and transportable gadolinium-core plastic scintillator sphere for neutron detection and counting

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Shafii, Mohammad Ali, E-mail: mashafii@fmipa.unand.ac.id; Meidianti, Rahma, E-mail: mashafii@fmipa.unand.ac.id; Wildian,, E-mail: mashafii@fmipa.unand.ac.id; Fitriyani, Dian, E-mail: mashafii@fmipa.unand.ac.id [Department of Physics, Andalas University Padang West Sumatera Indonesia (Indonesia); Tongkukut, Seni H. J. [Department of Physics, Sam Ratulangi University Manado North Sulawesi Indonesia (Indonesia); Arkundato, Artoto [Department of Physics, Jember University Jember East Java Indonesia (Indonesia)

    2014-09-30

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

  19. Radiation Transport Simulation for Boron Neutron Capture Therapy (BNCT)

    Energy Technology Data Exchange (ETDEWEB)

    Ziegner, M.; Blaickner, M. [AIT Austrian Institute of Technology GmbH, Health and Environment Department, Molecular Medicine, Muthgasse 11, 1190 Wien (Austria); Ziegner, M.; Khan, R.; Boeck, H. [Vienna University of Technology, Institute of Atomic and Subatomic Physics, Stadionallee 2, 1020 Wien (Austria); Bortolussi, S.; Altieri, S. [Department of Nuclear and Theoretical Physics, University of Pavia, National Institute of Nuclear Physics (INFN) Pavia Section, Pavia (Italy); Schmitz, T.; Hampel, G. [Nuclear Chemistry, University of Mainz, Fritz Strassmann Weg 2, 55099 Mainz (Germany)

    2011-07-01

    This work is part of a larger project initiated by the University of Mainz and aiming to use the university's TRIGA reactor to develop a treatment for liver metastases based on Boron Neutron Capture Therapy (BNCT). Diffuse distribution of cancerous cells within the organ makes complete resection difficult and the vicinity to radiosensitive organs impedes external irradiation. Therefore the method of 'autotransplantation', first established at the University of Pavia, is used. The liver is taken out of the body, irradiated in the thermal column of the reactor, therewith purged of metastases and then reimplanted. A highly precise dosimetry system is to be developed by means of measurements at the University of Mainz and computational calculations at the AIT. The stochastic MCNP-5 Monte Carlo-Code, developed by Los Alamos Laboratories, is applied. To verify the calculations of the flux and the absorbed dose in matter a number of measurements are performed irradiating different phantoms and liver sections in a 20cm x 20cm beam tube, which was created by removing graphite blocks from the thermal column of the reactor. The detector material consists of L- {alpha} -alanine pellets which are thought to be the most suitable because of their good tissue equivalence, small size and their wide response range. Another experiment focuses on the determination of the relative biological effectiveness (RBE-factor) of the neutron and photon dose for liver cells. Therefore cell culture plates with the cell medium enriched with {sup 157}Gd and {sup 10}B at different concentrations are irradiated. With regard to the alanine pellets MCNP-5 calculations give stable results. Nevertheless the absorbed dose is underestimated compared to the measurements, a phenomenon already observed in previous works. The cell culture calculations showed the enormous impact of the added isotopes with high thermal neutron cross sections, especially {sup 157}Gd, on the absorbed dose

  20. 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.

  1. In situ neutron depth profiling: A powerful method to probe lithium transport in micro-batteries

    NARCIS (Netherlands)

    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

  2. Analytical three-dimensional neutron transport benchmarks for verification of nuclear engineering codes. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Ganapol, B.D.; Kornreich, D.E. [Univ. of Arizona, Tucson, AZ (United States). Dept. of Nuclear Engineering

    1997-07-01

    Because of the requirement of accountability and quality control in the scientific world, a demand for high-quality analytical benchmark calculations has arisen in the neutron transport community. The intent of these benchmarks is to provide a numerical standard to which production neutron transport codes may be compared in order to verify proper operation. The overall investigation as modified in the second year renewal application includes the following three primary tasks. Task 1 on two dimensional neutron transport is divided into (a) single medium searchlight problem (SLP) and (b) two-adjacent half-space SLP. Task 2 on three-dimensional neutron transport covers (a) point source in arbitrary geometry, (b) single medium SLP, and (c) two-adjacent half-space SLP. Task 3 on code verification, includes deterministic and probabilistic codes. The primary aim of the proposed investigation was to provide a suite of comprehensive two- and three-dimensional analytical benchmarks for neutron transport theory applications. This objective has been achieved. The suite of benchmarks in infinite media and the three-dimensional SLP are a relatively comprehensive set of one-group benchmarks for isotropically scattering media. Because of time and resource limitations, the extensions of the benchmarks to include multi-group and anisotropic scattering are not included here. Presently, however, enormous advances in the solution for the planar Green`s function in an anisotropically scattering medium have been made and will eventually be implemented in the two- and three-dimensional solutions considered under this grant. Of particular note in this work are the numerical results for the three-dimensional SLP, which have never before been presented. The results presented were made possible only because of the tremendous advances in computing power that have occurred during the past decade.

  3. 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.

  4. 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.

  5. Hybrid variational principles and synthesis method for finite element neutron transport calculations

    Energy Technology Data Exchange (ETDEWEB)

    Ackroyd, R.T. (Queen Mary Coll., London (UK). Dept. of Nuclear Engineering); Nanneh, M.M. (Imperial Coll. of Science and Technology, London (UK). Dept. of Mechanical Engineering)

    1990-01-01

    A family of hybrid variational principles is derived using a generalised least squares method. Neutron conservation is automatically satisfied for the hybrid principles employing two trial functions. No interfaces or reflection conditions need to be imposed on the independent even-parity trial function. For some hybrid principles a single trial function can be employed by relating one parity trial function to the other, using one of the parity transport equation in relaxed form. For other hybrid principles the trial functions can be employed sequentially. Synthesis of transport solutions, starting with the diffusion theory approximation, has been used as a way of reducing the scale of the computation that arises with established finite element methods for neutron transport. (author).

  6. Atmospheric transport of neutrons and gamma rays from a high-altitude nuclear detonation

    Energy Technology Data Exchange (ETDEWEB)

    Byrd, R.C.

    1995-07-01

    Although radiation outputs from nuclear detonations in free space are well established, few studies exist of effect of atmospheric transport on the resulting intensity, energy, and time signatures. This report presents calculations for generic sources at high altitudes, 20-50 km above the Earth`s surface, in an atmosphere whose density decreases almost exponentially with height. The sources are instantaneous time bursts with simple energy dependences: gamma rays use an evaporation spectrum; neutrons use either a Gaussian fusion or a Maxwell fission spectrum. The observation angles vary from vertical to 5{degrees} below the horizon, and detectors are placed in either geosynchronous or low Earth orbits (100 km). All calculations use the Monte Carlo N-Particle (MCNP) transport code in either its photon, neutron, or coupled neutron-photon modes, with the coupled mode being applied to the production of gamma rays by neutron inelastic scattering. The standard MCNP outputs are analyzed to extract the intensity, energy, and time dependences of the fluence as functions of either source altitude or observation angle. In general, the intensities drop rapidly below about 30-km source altitude or +5` slant angle. Above these limits, the gamma-ray signal loses substantial intensity but still contains most of the original source information. In contrast, neutron scattering produces little or no decrease in intensity, but it rapidly degrades much of the information about the original source spectrum. Finally, although there is abundant gamma-ray production from neutron inelastic scattering, the resulting signatures appear to provide little additional information.

  7. 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.

  8. Hybrid principle with applications to synthesis. [Neutron transport theory

    Energy Technology Data Exchange (ETDEWEB)

    Nanneh, M.M. (Imperial Coll. of Science and Technology, London (United Kingdom). Dept. of Mechanical Engineering); Ackroyd, R.T. (Queen Mary Coll., London (United Kingdom))

    1991-01-01

    The theory of hybrid principles is presented together with the transformation rule for converting odd-parity approximations into even-parity approximations. This rule leads to a method which provides rigorous upper and lower bounds for the disadvantage factor for a reactor lattice cell. With these bounds very precise benchmarks have been constructed for representative lattices. It is found that a combination of even and odd-parity solutions for the neutron flux is much more efficient than solutions based on either the even-parity or odd-parity. This is the basis of one synthesis scheme. In another synthesis method, a hybrid principle with trial functions for both the even- and odd- parity angular flux is used in conjunction with a classical principle with an odd-parity trial function. The synthesis process is efficient because the largest set of equations to be solved, i.e. the frame work, involves as few as one unknown per node of the finite element mesh. The effectiveness of the synthesis method is demonstrated for a thick shield problem. (author).

  9. A transportable fast neutron and dual gamma-ray system for the detection of illicit materials

    Energy Technology Data Exchange (ETDEWEB)

    Fantidis, J.G. [Laboratory of Nuclear Technology, School of Engineering, ' Democritus' University of Thrace, Vas. Sofias 12, Xanthi 67100 (Greece); Nicolaou, G.E., E-mail: nicolaou@ee.duth.gr [Laboratory of Nuclear Technology, School of Engineering, ' Democritus' University of Thrace, Vas. Sofias 12, Xanthi 67100 (Greece)

    2011-08-21

    A transportable FNGR radiography system has been simulated using the MCNPX Monte Carlo code. The system is envisaged to be applied to the material characterisation of a suspicious bulky object, in view of identifying illegal materials. The system combines a neutron and two gamma-ray sources achieving characterisation of the material of the object through two ratios, namely {sup 137}Cs/DD and {sup 60}Co/DD. Hence, the system discriminates materials of similar or even the same of either of the two ratios. The proposed unit complies with radiation protection requirements achieving a safe occupational environment. - Highlights: > Transportable radiography system. > Neutron- and dual energy photon-beams available. > Discrimination of materials. > Detection of illicit materials.

  10. Effect of Fast Neutron Irradiation on Current Transport Properties of HTS Materials

    CERN Document Server

    Ballarino, A; Kruglov, V S; Latushkin, S T; Lubimov, A N; Ryazanov, A I; Shavkin, S V; Taylor, T M; Volkov, P V

    2004-01-01

    The effect of fast neutron irradiation with energy up to 35 MeV and integrated fluence of up to 5 x 10**15 cm-2 on the current transport properties of HTS materials Bi-2212 and Bi-2223 has been studied, both at liquid nitrogen and at room temperatures. The samples irradiated were selected after verification of the stability of their superconducting properties after temperature cycling in the range of 77 K - 293 K. It has been found that the irradiation by fast neutrons up to the above dose does not produce a significant degradation of critical current. The effect of room temperature annealing on the recovery of transport properties of the irradiated samples is also reported, as is a preliminary microstructure investigation of the effect of irradiation on the soldered contacts.

  11. Apparatus, Method and Program Storage Device for Determining High-Energy Neutron/Ion Transport to a Target of Interest

    Science.gov (United States)

    Wilson, John W. (Inventor); Tripathi, Ram K. (Inventor); Badavi, Francis F. (Inventor); Cucinotta, Francis A. (Inventor)

    2012-01-01

    An apparatus, method and program storage device for determining high-energy neutron/ion transport to a target of interest. Boundaries are defined for calculation of a high-energy neutron/ion transport to a target of interest; the high-energy neutron/ion transport to the target of interest is calculated using numerical procedures selected to reduce local truncation error by including higher order terms and to allow absolute control of propagated error by ensuring truncation error is third order in step size, and using scaling procedures for flux coupling terms modified to improve computed results by adding a scaling factor to terms describing production of j-particles from collisions of k-particles; and the calculated high-energy neutron/ion transport is provided to modeling modules to control an effective radiation dose at the target of interest.

  12. Modelling of neutron and photon transport in iron and concrete radiation shieldings by the Monte Carlo method - Version 2

    CERN Document Server

    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.

  13. 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.)

  14. Monte Carlo transport simulation for a long counter neutron detector employed as a cosmic rays induced neutron monitor at ground level

    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)

  15. 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.

  16. Finite element methods for neutron transport based on maximum and minimum principles for discontinuous trial functions

    Energy Technology Data Exchange (ETDEWEB)

    Ackroyd, R.T. (Imperial Coll. of Science and Technology, London (United Kingdom). Dept. of Mechanical Engineering)

    Variational principles are given for the first order Boltzmann equation for neutron transport and its equivalent second order forms for even and odd-parity transport. For these extremum principles there are no boundary conditions to be imposed, and across interfaces the trial functions can be discontinuous. Numerical solutions are presented for the even-parity principle. The scale of interface discontinuities is controlled by a penalty parameter [lambda]. As [lambda] is increased for a given basis the degree of satisfaction of the transport equation by the optimised trial function decreases to that for a classical trial function, whereas the interface discontinuities decrease. These trends lend themselves to a simple interpretation in terms of vectors in a Hilbert space. In applications the spatial dependence of a trial function is represented by finite elements, and a spherical harmonic expansion gives the directional dependence of the trial function. The principles are applied in the method of composite solutions in which the order of the spherical harmonic expansion can be varied according to the physical needs of the regions of a system. Large values of [lambda] are used in the method of composite solutions. On the other hand specific values of [lambda] when used with the simplest finite element lead to the well-known central difference equations for neutron transport. (author).

  17. CAD-Based Monte Carlo Neutron Transport KSTAR Analysis for KSTAR

    Science.gov (United States)

    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.

  18. Geant4 simulations of the neutron production and transport in the n_TOF spallation target

    CERN Document Server

    Lerendegui-Marco, J; Guerrero, C; Quesada,, J , M

    2016-01-01

    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 1 GeV, 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.

  19. Curved finite elements and acceleration for the neutron transport; Elements finis courbes et acceleration pour le transport de neutrons

    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

  20. Evaluation of HYLIFE-II and Sombrero using 175- and 566-group neutron transport and activation cross sections

    CERN Document Server

    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.

  1. Modeling of neutron and photon transport in iron and concrete radiation shields by using Monte Carlo method

    CERN Document Server

    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.

  2. Modular, object-oriented redesign of a large-scale Monte Carlo neutron transport program

    Energy Technology Data Exchange (ETDEWEB)

    Moskowitz, B.S.

    2000-02-01

    This paper describes the modular, object-oriented redesign of a large-scale Monte Carlo neutron transport program. This effort represents a complete 'white sheet of paper' rewrite of the code. In this paper, the motivation driving this project, the design objectives for the new version of the program, and the design choices and their consequences will be discussed. The design itself will also be described, including the important subsystems as well as the key classes within those subsystems.

  3. 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.

  4. Equations of the surface harmonics method for solving time-dependent neutron transport problems and their verification

    Science.gov (United States)

    Boyarinov, V. F.; Kondrushin, A. E.; Fomichenko, P. A.

    2013-12-01

    Time-dependent equations of the surface harmonics method (SHM) are obtained for planar one-dimensional geometry. The equations are verified by calculations of test problems from Benchmark Problem Book ANL-7416, and the capabilities and efficiency of applying the SHM for solving the time-dependent neutron transport equation in the diffusion approximation are demonstrated. The results of the work show that the implementation of the SHG for full-scale computations will make possible substantial progress in the efficient solution of time-dependent problems of neutron transport in nuclear reactors.

  5. Analytical synthetic methods of solution of neutron transport equation with diffusion theory approaches energy multigroup; Metodos sinteticos analiticos de solucao da equacao de transporte de neutrons com aproximacoes da teoria da difusao multigrupo de energia

    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.

  6. Comparison of preconditioned generalized conjugate gradient methods to two-dimensional neutron and photon transport equation

    Energy Technology Data Exchange (ETDEWEB)

    Chen, G.S.; Yang, D.Y

    1998-01-01

    We apply and compare the preconditioned generalized conjugate gradient methods to solve the linear system equation that arises in the two-dimensional neutron and photon transport equation in this paper. Several subroutines are developed on the basis of preconditioned generalized conjugate gradient methods for time-independent, two-dimensional neutron and photon transport equation in the transport theory. These generalized conjugate gradient methods are used: TFQMR (transpose free quasi-minimal residual algorithm) CGS (conjugate gradient square algorithm), Bi-CGSTAB (bi-conjugate gradient stabilized algorithm) and QMRCGSTAB (quasi-minimal residual variant of bi-conjugate gradient stabilized algorithm). These subroutines are connected to computer program DORT. Several problems are tested on a personal computer with Intel Pentium CPU. The reasons to choose the generalized conjugate gradient methods are that the methods have better residual (equivalent to error) control procedures in the computation and have better convergent rate. The pointwise incomplete LU factorization ILU, modified pointwise incomplete LU factorization MILU, block incomplete factorization BILU and modified blockwise incomplete LU factorization MBILU are the preconditioning techniques used in the several testing problems. In Bi-CGSTAB, CGS, TFQMR and QMRCGSTAB method, we find that either CGS or Bi-CGSTAB method combined with preconditioner MBILU is the most efficient algorithm in these methods in the several testing problems. The numerical solution of flux by preconditioned CGS and Bi-CGSTAB methods has the same result as those from Cray computer, obtained by either the point successive relaxation method or the line successive relaxation method combined with Gaussian elimination.

  7. 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.

  8. 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)

  9. A hybrid method for coupled neutron-ion transport calculations for {sup 10}B and {sup 6}LiF coated and perforated detector efficiencies

    Energy Technology Data Exchange (ETDEWEB)

    Solomon, C.J. [Department of Mechanical and Nuclear Engineering, Kansas State University, Manhattan, KS 66506 (United States)], E-mail: cjs8888@ksu.edu; Shultis, J.K.; McNeil, W.J.; Unruh, T.C.; Rice, B.B.; McGregor, D.S. [Department of Mechanical and Nuclear Engineering, Kansas State University, Manhattan, KS 66506 (United States)

    2007-09-21

    A hybrid method for neutron-ion transport has been developed and applied to modeling coated and perforated neutron semiconductor detectors. The method couples the MCNP transport code developed by Los Alamos National Laboratory and a specialized ion transport code. Output from MCNP is passed to the ion transport code to perform ion energy deposition calculations, and in this manner the two codes are married. The process is controlled by a PERL script. Angle dependent efficiency calculation results for perforated rod detectors are presented, and neutron absorption efficiencies are presented for channel and chevron type perforations to show how problems with rod perforations may be overcome.

  10. Completely boundary-free minimum and maximum principles for neutron transport and their least-squares and Galerkin equivalents

    Energy Technology Data Exchange (ETDEWEB)

    Ackroyd, R.T. (UKAEA Risley Nuclear Power Development Establishment. Technical Services and Planning Directorate)

    1982-01-01

    Some minimum and maximum variational principles for even-parity neutron transport are reviewed and the corresponding principles for odd-parity transport are derived by a simple method to show why the essential boundary conditions associated with these maximum principles have to be imposed. The method also shows why both the essential and some of the natural boundary conditions associated with these minimum principles have to be imposed. These imposed boundary conditions for trial functions in the variational principles limit the choice of the finite element used to represent trial functions. The reasons for the boundary conditions imposed on the principles for even- and odd-parity transport point the way to a treatment of composite neutron transport, for which completely boundary-free maximum and minimum principles are derived from a functional identity. In general a trial function is used for each parity in the composite neutron transport, but this can be reduced to one without any boundary conditions having to be imposed.

  11. 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.

  12. Finite-element method for multigroup neutron transport: anisotropic scattering in 1-D slab geometry

    Energy Technology Data Exchange (ETDEWEB)

    Riyait, N.S.; Ackroyd, R.T.

    1987-01-01

    Proof-tests on 1-D multigroup neutron transport problems are reported for strong anisotropic scattering. These tests have been undertaken as part of the validation of the 3-D multigroup finite-element transport code FELTRAN for anisotropic scattering media. To illustrate the treatment of within-group and intergroup anisotropic scattering in the finite-element method the relevant theory is outlined. Ingroup scattering is checked using the backward-forward-isotropic (BFI) scattering law for source and eigenvalue problems. With this law anisotropic scattering problems can be transformed into equivalent isotropic scattering problems. In this way the well-validated isotropic scattering version of FELTRAN is used to validate the anisotropic version. Intergroup scattering effects are checked by solving few-group source problems for P/sub 1/ and P/sub 3/ scattering and the BFI scattering law. For P/sub 1/ and P/sub 3/ scattering checks are made with the discrete-ordinate finite-difference code ANISN and the spherical harmonics finite-difference code MARC/PN. For the BFI scattering law comparison is made with two-group exact solutions of Williams (1985) for 1-D systems.

  13. 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.

  14. Study of the response of a lithium yttrium borate scintillator based neutron rem counter by Monte Carlo radiation transport simulations

    Energy Technology Data Exchange (ETDEWEB)

    Sunil, C., E-mail: csunil11@gmail.com [Accelerator Radiation Safety Section, Health Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Tyagi, Mohit [Technical Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Biju, K.; Shanbhag, A.A.; Bandyopadhyay, T. [Accelerator Radiation Safety Section, Health Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2015-12-11

    The scarcity and the high cost of {sup 3}He has spurred the use of various detectors for neutron monitoring. A new lithium yttrium borate scintillator developed in BARC has been studied for its use in a neutron rem counter. The scintillator is made of natural lithium and boron, and the yield of reaction products that will generate a signal in a real time detector has been studied by FLUKA Monte Carlo radiation transport code. A 2 cm lead introduced to enhance the gamma rejection shows no appreciable change in the shape of the fluence response or in the yield of reaction products. The fluence response when normalized at the average energy of an Am–Be neutron source shows promise of being used as rem counter.

  15. A Novel Algorithm for Solving the Multidimensional Neutron Transport Equation on Massively Parallel Architectures

    Energy Technology Data Exchange (ETDEWEB)

    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.

  16. Improvements in Neutronics/Thermal-Hydraulics Coupling in Two-Phase Flow Systems Using Stochastic-Mixture Transport Models

    CERN Document Server

    Palmer, T S

    2003-01-01

    In this NEER project, researchers from Oregon State University have investigated the limitations of the treatment of two-phase coolants as a homogeneous mixture in neutron transport calculations. Improved methods of calculating the neutron distribution in binary stochastic mixtures have been developed over the past 10-15 years and are readily available in the transport literature. These methods are computationally more expensive than the homogeneous (or atomic mix) models, but can give much more accurate estimates of ensemble average fluxes and reaction rates provided statistical descriptions of the distributions of the two materials are know. A thorough review of the two-phase flow literature has been completed and the relevant mixture distributions have been identified. Using these distributions, we have performed Monte Carlo criticality calculations of fuel assemblies to assess the accuracy of the atomic mix approximation when compared to a resolved treatment of the two-phase coolant. To understand the ben...

  17. An Innovative Three-Dimensional Heterogeneous Coarse-Mesh Transport Method for Advanced and Generation IV Reactor Core Analysis and Design

    Energy Technology Data Exchange (ETDEWEB)

    Farzad Rahnema

    2009-11-12

    This project has resulted in a highly efficient method that has been shown to provide accurate solutions to a variety of 2D and 3D reactor problems. The goal of this project was to develop (1) an accurate and efficient three-dimensional whole-core neutronics method with the following features: based sollely on transport theory, does not require the use of cross-section homogenization, contains a highly accurate and self-consistent global flux reconstruction procedure, and is applicable to large, heterogeneous reactor models, and to (2) create new numerical benchmark problems for code cross-comparison.

  18. A discrete ordinates approximation to the neutron transport equation applied to generalized geometries

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, Mark David [Texas A & M Univ., College Station, TX (United States)

    1992-12-01

    A method for applying the discrete ordinates method for solution of the neutron transport equation in arbitary two-dimensional meshes has been developed. The finite difference approach normally used to approximate spatial derivatives in extrapolating angular fluxes across a cell is replaced by direct solution of the characteristic form of the transport equation for each discrete direction. Thus, computational cells are not restricted to the traditional shape of a mesh element within a given coordinate system. However, in terms of the treatment of energy and angular dependencies, this method resembles traditional discrete ordinates techniques. Using the method developed here, a general two-dimensional space can be approximated by an irregular mesh comprised of arbitrary polygons. The present work makes no assumptions about the orientations or the number of sides in a given cell, and computes all geometric relationships between each set of sides in each cell for each discrete direction. A set of non-reentrant polygons can therefore be used to represent any given two dimensional space. Results for a number of test problems have been compared to solutions obtained from traditional methods, with good agreement. Comparisons include benchmarks against analytical results for problems with simple geometry, as well numerical results obtained from traditional discrete ordinates methods by applying the ANISN and TWOTRAN computer programs. Numerical results were obtained for problems ranging from simple one-dimensional geometry to complicated multidimensional configurations. These results have demonstrated the ability of the developed method to closely approximate complex geometrical configurations and to obtain accurate results for problems that are extremely difficult to model using traditional methods.

  19. Performance Analysis of Fission and Surface Source Iteration Method for Domain Decomposed Monte Carlo Whole-Core Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Yu Gwon; Oh, Yoo Min; Park, Hyang Kyu; Park, Kang Soon; Cho, Nam Zin [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    In this paper, two issues in the FSS iteration method, i.e., the waiting time for surface source data and the variance biases in local tallies are investigated for the domain decomposed, 3-D continuous-energy whole-core calculation. The fission sources are provided as usual, while the surface sources are provided by banking MC particles crossing local domain boundaries. The surface sources serve as boundary conditions for nonoverlapping local problems, so that each local problem can be solved independently. In this paper, two issues in the FSS iteration are investigated. One is quantifying the waiting time of processors to receive surface source data. By using nonblocking communication, 'time penalty' to wait for the arrival of the surface source data is reduced. The other important issue is underestimation of the sample variance of the tally because of additional inter-iteration correlations in surface sources. From the numerical results on a 3-D whole-core test problem, it is observed that the time penalty is negligible in the FSS iteration method and that the real variances of both pin powers and assembly powers are estimated by the HB method. For those purposes, three cases; Case 1 (1 local domain), Case 2 (4 local domains), Case 3 (16 local domains) are tested. For both Cases 2 and 3, the time penalties for waiting are negligible compared to the source-tracking times. However, for finer divisions of local domains, the loss of parallel efficiency caused by the different number of sources for local domains in symmetric locations becomes larger due to the stochastic errors in source distributions. For all test cases, the HB method very well estimates the real variances of local tallies. However, it is also noted that the real variances of local tallies estimated by the HB method show slightly smaller than the real variances obtained from 30 independent batch runs and the deviations become larger for finer divisions of local domains. The batch size used

  20. Neutron transport in hexagonal reactor cores modeled by trigonal-geometry diffusion and simplified P{sub 3} nodal methods

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. Neutron Albedo

    CERN Document Server

    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.

  2. Analytical solution of neutron transport equation in an annular reactor with a rotating pulsed source; Resolucao analitica da equacao de transporte de neutrons em um reator anelar com fonte pulsada rotativa

    Energy Technology Data Exchange (ETDEWEB)

    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)

  3. Neutronics - thermal-hydraulics coupling: application to the helium-cooled fast reactor; Couplage neutronique - thermohydraulique: application au reacteur a neutrons rapides refroidi a l'helium

    Energy Technology Data Exchange (ETDEWEB)

    Vaiana, F.

    2009-11-15

    This thesis focuses on the study of interactions between neutron-kinetics and thermal-hydraulics. Neutron-kinetics allow to calculate the power in a nuclear reactor and the temperature evolution of materials where this power is deposited is known thanks to thermal-hydraulics. Moreover, when the temperatures evolve, the densities and cross sections change. These two disciplines are thus coupled. The first part of this work corresponds to the study and development of a method which allows to simulate transients in nuclear reactors and especially with a Monte-Carlo code for neutron-kinetics. An algorithm for the resolution of the neutron transport equation has been established and validated with a benchmark. In thermal-hydraulics, a porous media approach, based on another thesis, is considered. This gives the opportunity to solve the equations on the whole core without unconscionable computation time. Finally, a theoretical study has been performed on the statistical uncertainties which result from the use of a Monte-Carlo code and which spread from the reactivity to the power and from the power to the temperatures. The second part deals with the study of a misplaced control rod withdrawing in a GFR (helium-cooled fast reactor), a fourth generation reactor. Some models allowing to calculate neutron-kinetics and thermal-hydraulics in the core (which contains assemblies built up with fuel plates) were defined. In thermal-hydraulics, a model for the core based on the porous media approach and a fuel plate homogenization model have been set up. A similar homogenization model has been studied for neutron-kinetics. Finally, the control rod withdrawing transient where we can observe the power raising and the stabilisation by thermal feedback has been performed with the Monte-Carlo code Tripoli for neutron-kinetics and the code Trio-U for thermal-hydraulics. (author)

  4. Neutron Diffraction and Electrical Transport Studies on Magnetic Transition in Terbium at High Pressures and Low Temperatures

    Science.gov (United States)

    Thomas, Sarah; Montgomery, Jeffrey; Tsoi, Georgiy; Vohra, Yogesh; Weir, Samuel; Tulk, Christopher; Moreira Dos Santos, Antonio

    2013-06-01

    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 its transition from a helical antiferromagnetic to a ferromagnetic ordered phase as a function of pressure. The electrical resistance measurements using designer diamonds show a change in slope as the temperature is lowered through the ferromagnetic Curie temperature. The temperature of the ferromagnetic transition decreases at a rate of -16.7 K/GPa till 3.6 GPa, where terbium undergoes a structural transition from hexagonal close packed (hcp) to an α-Sm phase. Above this pressure, the electrical resistance measurements no longer exhibit a change in slope. In order to confirm the change in magnetic phase suggested by the electrical resistance measurements, neutron diffraction measurements were conducted at the SNAP beamline at the Oak Ridge National Laboratory. Measurements were made at pressures to 5.3 GPa and temperatures as low as 90 K. An abrupt increase in peak intensity in the neutron diffraction spectra signaled the onset of magnetic order below the Curie temperature. A magnetic phase diagram of rare earth metal terbium will be presented to 5.3 GPa and 90 K based on these studies.

  5. TARTNP: a coupled neutron--photon Monte Carlo transport code. [10-/sup 9/ to 20 MeV; in LLL FORTRAN

    Energy Technology Data Exchange (ETDEWEB)

    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.

  6. Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    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

  7. The coupling of the neutron transport application RATTLESNAKE to the nuclear fuels performance application BISON under the MOOSE framework

    Energy Technology Data Exchange (ETDEWEB)

    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.

  8. Development of a three-dimensional neutron transport code DFEM based on the double finite element method

    Energy Technology Data Exchange (ETDEWEB)

    Fujimura, Toichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1996-01-01

    A three-dimensional neutron transport code DFEM has been developed by the double finite element method to analyze reactor cores with complex geometry as large fast reactors. Solution algorithm is based on the double finite element method in which the space and angle finite elements are employed. A reactor core system can be divided into some triangular and/or quadrangular prism elements, and the spatial distribution of neutron flux in each element is approximated with linear basis functions. As for the angular variables, various basis functions are applied, and their characteristics were clarified by comparison. In order to enhance the accuracy, a general method is derived to remedy the truncation errors at reflective boundaries, which are inherent in the conventional FEM. An adaptive acceleration method and the source extrapolation method were applied to accelerate the convergence of the iterations. The code structure is outlined and explanations are given on how to prepare input data. A sample input list is shown for reference. The eigenvalue and flux distribution for real scale fast reactors and the NEA benchmark problems were presented and discussed in comparison with the results of other transport codes. (author).

  9. 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

  10. Analysis of the neutrons dispersion in a semi-infinite medium based in transport theory and the Monte Carlo method; Analisis de la dispersion de neutrones en un medio semi-infinito en base a teoria de transporte y el metodo de Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    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)

  11. Identification and properties of the fundamental expansion functions for neutron transport in an infinite homogeneous medium

    Energy Technology Data Exchange (ETDEWEB)

    Milgram, Michael S. [P.O. Box 1484, Deep River, Ont., K0J 1P0 (Canada)]. E-mail: mike@geometrics-unlimited.com

    2005-07-15

    Starting from the basic expression for the neutron flux due to a point source in an infinite homogeneous scattering and absorbing medium, the first few fundamental expansion functions corresponding to successive collisions are identified, and their analytic properties are presented, in spherical and plane geometry. Various representations of the functions are obtained in the form of power series, an expansion in a series of exponential integrals, and other integrals. The adequacy of traditional asymptotic forms is considered.

  12. Standard practice for qualification and acceptance of boron based metallic neutron absorbers for nuclear criticality control for dry cask storage systems and transportation packaging

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This practice provides procedures for qualification and acceptance of neutron absorber materials used to provide criticality control by absorbing thermal neutrons in systems designed for nuclear fuel storage, transportation, or both. 1.2 This practice is limited to neutron absorber materials consisting of metal alloys, metal matrix composites (MMCs), and cermets, clad or unclad, containing the neutron absorber boron-10 (10B). 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  13. Fast Neutron Transport in the Biological Shielding Model and Other Regions of the VVER-1000 Mock-Up on the LR-0 Research Reactor

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. Parallel computing for homogeneous diffusion and transport equations in neutronics; Calcul parallele pour les equations de diffusion et de transport homogenes en neutronique

    Energy Technology Data Exchange (ETDEWEB)

    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)

  16. TART 2000 A Coupled Neutron-Photon, 3-D, Combinatorial Geometry, Time Dependent, Monte Carlo Transport Code

    CERN Document Server

    Cullen, D

    2000-01-01

    TART2000 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. TART2000 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. TART2000 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. TART2000 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART2000 and its data files.

  17. Using 2H labeling with neutron radiography for the study of solid polymer electrolyte water transport properties.

    Science.gov (United States)

    Boillat, P; Oberholzer, P; Seyfang, B C; Kästner, A; Perego, R; Scherer, G G; Lehmann, E H; Wokaun, A

    2011-06-15

    A method combining (2)H labeling of different sources of H atoms (hydrogen, water vapor) with neutron imaging for the analysis of transport parameters in the bulk and at the interfaces of Nafion polymer electrolyte membranes is proposed. The use of different isotope compositions in the steady state allows evaluation of the relation between bulk and interface transport parameters, but relies on literature data for evaluating absolute values. By using transients of isotope composition, absolute values of these parameters including the self-diffusion coefficient of H can be extracted, making this method an attractive alternative to self-diffusion measurements using nuclear magnetic resonance (NMR), allowing measurements in precisely controlled conditions in real fuel cell structures. First measurements were realized on samples with and without electrodes and we report values of the self-diffusion coefficient of the same order of magnitude as values measured using NMR, although with slightly higher numbers. In our particular case, lower interfacial exchange rates for water transport were observed for samples with an electrode.

  18. ACCELERATING FUSION REACTOR NEUTRONICS MODELING BY AUTOMATIC COUPLING OF HYBRID MONTE CARLO/DETERMINISTIC TRANSPORT ON CAD GEOMETRY

    Energy Technology Data Exchange (ETDEWEB)

    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).

  19. Multi-Sensor Core Logger (MSCL) P-wave velocity, gamma-ray density, and magnetic susceptibility whole-core logs of sediment cores collected in 2009 offshore from Palos Verdes, California

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — This part of the data release includes Multi-Sensor Core Logger (MSCL) P-wave velocity, gamma-ray density, and magnetic susceptibility whole-core logs of sediment...

  20. Neutron/gamma coupled library generation and gamma transport calculation with KARMA 1.2

    Energy Technology Data Exchange (ETDEWEB)

    Hong, S. G. [Dept. of Nuclear Engineering, Kyung Hee Univ., 446-701 Deogyeong-daero, GiHeung-gu, Yongin, Gyeonggi-do (Korea, Republic of); Kim, K. S.; Cho, J. Y.; Lee, K. H. [Korea Atomic Energy Research Inst., 305-353 Duckjin-dong, Yuseong-gu, Daejon (Korea, Republic of)

    2012-07-01

    KAERI has developed a lattice transport calculation code KARMA and its multi-group cross section library generation system. Recently, the multi-group cross section library generation system has included a gamma cross section generation capability and KARMA also has been improved to include a gamma transport calculation module. This paper addresses the multi-group gamma cross section generation capability for the KARMA 1.2 code and the preliminary test results of the KARMA 1.2 gamma transport calculations. The gamma transport calculation with KARMA 1.2 gives the gamma flux, gamma smeared power, and gamma energy deposition distributions. The results of the KARMA gamma calculations were compared with those of HELIOS and they showed that KARMA 1.2 gives reasonable gamma transport calculation results. (authors)

  1. Solution of the Neutron transport equation in hexagonal geometry using strongly discontinuous nodal schemes; Solucion de la Ecuacion de transporte de neutrones en geometria hexagonal usando esquemas nodales fuertemente discontinuos

    Energy Technology Data Exchange (ETDEWEB)

    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)

  2. Trade Study for Neutron Transport at Low Earth Orbit: Adding Fidelity to DIORAMA

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. Analytical Tests for Ray Effect Errors in Discrete Ordinate Methods for Solving the Neutron Transport Equation

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. Neutron transport-burnup code MCORGS and its application in fusion fission hybrid blanket conceptual research

    Science.gov (United States)

    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.

  5. Analysis of a HP-refinement method for solving the neutron transport equation using two error estimators

    Energy Technology Data Exchange (ETDEWEB)

    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)

  6. Transport theory calculation for a heterogeneous multi-assembly problem by characteristics method with direct neutron path linking technique

    Energy Technology Data Exchange (ETDEWEB)

    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)

  7. The bidimensional neutron transport code TWOTRAN-GG. Users manual and input data TWOTRAN-TRACA version; El codigo de transporte bidimensional TWOTRAN-GG. Manual de usuario y datos de entrada version TWOTRAN-TRACA

    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.

  8. An overview of the Boltzmann transport equation solution for neutrons, photons and electrons in cartesian geometry

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez, Barbara D. do Amaral, E-mail: barbara.arodriguez@gmail.co [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica; Vilhena, Marco Tullio, E-mail: vilhena@mat.ufrgs.b [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Dept. de Matematica Pura e Aplicada

    2009-07-01

    Questions regarding accuracy and efficiency of deterministic transport methods are still on our mind today, even with modern supercomputers. The most versatile and widely used deterministic methods are the P{sub N} approximation, the S{sub N} method (discrete ordinates method) and their variants. In the discrete ordinates (S{sub N}) formulations of the transport equation, it is assumed that the linearized Boltzmann equation only holds for a set of distinct numerical values of the direction-of-motion variables. In this work, looking forward to confirm the capabilities of deterministic methods in obtaining accurate results, we present a general overview of deterministic methods to solve the Boltzmann transport equation for neutral and charged particles. First, we describe a review in the Laplace transform technique applied to S{sub N} two dimensional transport equation in a rectangular domain considering Compton scattering. Next, we solved the Fokker-Planck (FP) equation, an alternative approach for the Boltzmann transport equation, assuming a monoenergetic electron beam in a rectangular domain. The main idea relies on applying the P{sub N} approximation, a recent advance in the class of deterministic methods, in the angular variable, to the two dimensional Fokker-Planck equation and then applying the Laplace Transform in the spatial x-variable. Numerical results are given to illustrate the accuracy of deterministic methods presented. (author)

  9. The use of symbolic computation in radiative, energy, and neutron transport calculations. Final report

    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.

  10. Hydrodynamics and Radiation Transport in Radiation Dominated Accretion onto Neutron Stars

    Science.gov (United States)

    Wolfram, Kenneth D.; Becker, P. A.

    2011-01-01

    The problem of spectral formation in accretion-powered X-ray pulsars was solved for the first time using an analytical model based on the fundamental physics in 2007, and the resulting spectra were shown to agree rather closely with those observed from several of the most luminous sources. However, in order to derive the analytical solutions, simplifying assumptions were made regarding the inflow velocity profile, the thermal structure of the plasma, the boundary conditions, and the geometry of the column. In this paper, the problem is revisited using a new numerical approach that facilitates the solution of a more realistic, coupled radiative-hydrodynamical model. The new model utilizes a conical geometry for the accretion flow and applies a robust free-streaming boundary condition at the top of the column, along with a "mirror" boundary condition at the neutron star surface. The temperature of the electrons is computed based on inverse-Compton equilibration, and the hydrodynamical structure of the column is determined by solving the coupled set of conservation equations for momentum, energy, and mass. The column-integrated spectra computed using the new model are compared with the data for several sources, and the resulting source parameters are compared with those computed using the original analytical model.

  11. Semiconductor neutron detector

    Science.gov (United States)

    Ianakiev, Kiril D [Los Alamos, NM; Littlewood, Peter B [Cambridge, GB; Blagoev, Krastan B [Arlington, VA; Swinhoe, Martyn T [Los Alamos, NM; Smith, James L [Los Alamos, NM; Sullivan, Clair J [Los Alamos, NM; Alexandrov, Boian S [Los Alamos, NM; Lashley, Jason Charles [Santa Fe, NM

    2011-03-08

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

  12. Asymptotic analysis of the several competitive equations to solve the time-dependent neutron transport equation

    Energy Technology Data Exchange (ETDEWEB)

    Shin, U.; Miller, W.F. Jr. [California Univ., Berkeley, CA (United States)]|[Los Alamos National Lab., NM (United States); Morel, J.E. [Los Alamos National Lab., NM (United States)

    1994-10-01

    Using conventional diffusion limit analysis, we asymptotically compare three competitive time-dependent equations (the telegrapher`s equation, the time-dependent Simplified P{sub 2} (SP{sub 2}) equation, and the time-dependent Simplified Evcn-Parity (SEP) equation). The time-dependent SP{sub 2} equation contains higher order asymptotic approximations of the time-dependent transport equation than the other equations in a physical regime in which the time-dependent diffusion equation is the leading order approximation. In addition, we derive the multigroup modified time-dependent SP{sub 2} equation from the multigroup time-dependent transport equation by means of an asymptotic expansion in which the multigroup time-dependent diffusion equation is the leading, order approximation. Numerical comparisons of the timedependent diffusion, the telegrapher`s, the time-dependent SP{sub 2}, and S{sub 8} solutions in 2-D X-Y geometry show that, in most cases, the SP{sub 2} solutions contain most of the transport corrections for the diffusion approximation.

  13. Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR

    Directory of Open Access Journals (Sweden)

    Brovchenko Mariya

    2017-01-01

    Full Text Available The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR. The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

  14. A Posteriori Error Estimation for a Nodal Method in Neutron Transport Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Azmy, Y.Y.; Buscaglia, G.C.; Zamonsky, O.M.

    1999-11-03

    An a posteriori error analysis of the spatial approximation is developed for the one-dimensional Arbitrarily High Order Transport-Nodal method. The error estimator preserves the order of convergence of the method when the mesh size tends to zero with respect to the L{sup 2} norm. It is based on the difference between two discrete solutions that are available from the analysis. The proposed estimator is decomposed into error indicators to allow the quantification of local errors. Some test problems with isotropic scattering are solved to compare the behavior of the true error to that of the estimated error.

  15. Development of a transient three-dimensional neutron transport code with feedback

    Energy Technology Data Exchange (ETDEWEB)

    Waddell, M.W. Jr.

    1994-07-19

    A new code is being developed at the Y-12 Plant for solving the time-dependent, three-dimensional Boltzmann transport model with feedback. The new code, PADK, uses the quasi-static method in its adiabatic form and is to be utilized to analyze hypothetical criticality accidents. A description of the code along with preliminary results without feedback are presented in this paper. The code is applied to 2 standard benchmark problems and the results are compared to another method. Also, the code is used to model the GODIVA reactor. Further work needed to be completed is described.

  16. Development of a transient three-dimensional neutron transport code with feedback

    Energy Technology Data Exchange (ETDEWEB)

    Waddell, M.W. Jr.

    1994-12-31

    A new code is being developed at the Y-12 plant for solving the time-dependent, three-dimensional Boltzmann transport model with feedback. The new code, PADK, uses the quasi-static method in its adiabatic form and is to be utilized to analyze hypothetical criticality accidents. A description of the code along with preliminary results without feedback are presented in this paper. The code is applied to two standard benchmark problems, and the results are compared to another method. Also, the code is used to model the GODIVA reactor. Further work needed to be completed is described.

  17. Spectral nodal methodology for multigroup slab-geometry discrete ordinates neutron transport problems with linearly anisotropic scattering

    Energy Technology Data Exchange (ETDEWEB)

    Oliva, Amaury M.; Filho, Hermes A.; Silva, Davi M.; Garcia, Carlos R., E-mail: aoliva@iprj.uerj.br, E-mail: halves@iprj.uerj.br, E-mail: davijmsilva@yahoo.com.br, E-mail: cgh@instec.cu [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Instituto Politecnico. Departamento de Modelagem Computacional; Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba)

    2017-07-01

    In this paper, we propose a numerical methodology for the development of a method of the spectral nodal class that will generate numerical solutions free from spatial truncation errors. This method, denominated Spectral Deterministic Method (SDM), is tested as an initial study of the solutions (spectral analysis) of neutron transport equations in the discrete ordinates (S{sub N}) formulation, in one-dimensional slab geometry, multigroup approximation, with linearly anisotropic scattering, considering homogeneous and heterogeneous domains with fixed source. The unknowns in the methodology are the cell-edge, and cell average angular fluxes, the numerical values calculated for these quantities coincide with the analytic solution of the equations. These numerical results are shown and compared with the traditional ne- mesh method Diamond Difference (DD) and the coarse-mesh method spectral Green's function (SGF) to illustrate the method's accuracy and stability. The solution algorithms problems are implemented in a computer simulator made in C++ language, the same that was used to generate the results of the reference work. (author)

  18. A stochastic/deterministic method for transient, three-dimensional neutron transport

    Energy Technology Data Exchange (ETDEWEB)

    Waddell, M.W. Jr. [Oak Ridge Y-12 Plant, TN (United States); Dodds, H.L. [Tennessee Univ., Knoxville, TN (United States). Dept. of Nuclear Engineering

    1992-06-11

    This paper describes the development of a method for solving the time-dependent, three-dimensional Boltzmann transport model. A hybrid stochastic/deterministic technique is utilized with a Monte Carlo code embedded inside of a quasi-static kinetics framework. The amplitude function is computed deterministically by a conventional point kinetics algorithm. The point kinetics parameters, reactivity and generation time, as well as the flux shape, are computed stochastically during the random walk of the Monte Carlo calculation. The code which has been developed based on this new method could serve as a benchmarking tool for other more approximate and less expensive kinetics codes. It would also be useful for modeling transients of systems with complex geometries. Furthermore, 3-D transport theory is needed in areas such as reactor accident studies which involve coolant voiding and/or core disassembly. Preliminary results are presented in this paper. The code is applied to a standard benchmark problem and the results are compared to another method.

  19. A hybrid stochastic/deterministic method for transient, three-dimensional neutron transport

    Energy Technology Data Exchange (ETDEWEB)

    Waddell, M.W. Jr. (Martin Marietta Energy Systems, Inc., Oak Ridge, TN (United States)); Dodds, H.L. (Univ. of Tennessee, Knoxville (United States))

    1992-01-01

    This paper describes the development of a method for solving the time-dependent, three-dimensional Boltzmann transport model. A hybrid stochastic/deterministic technique is utilized with a Monte Carlo code embedded inside of a quasistatic kinetics framework. The amplitude function is computed deterministically by a conventional point-kinetics algorithm. The point-kinetics parameters, reactivity and generation time, as well as the flux shape, are computed stochastically during the random walk of the Monte Carlo calculation. The code (TDKENO), which was developed based on this new method, could serve as a benchmarking tool for other more approximate and less expensive kinetics codes. It would also be useful for modeling transients of systems with complex geometries. Furthermore, three-dimensional transport theory is needed in areas such as reactor accident studies that involve coolant voiding and/or core disassembly. Preliminary results are presented in this paper. The code is applied to a standard benchmark problem, and the results are compared with another method. 7 refs., 1 fig.

  20. Development of modern methods with respect to neutron transport and uncertainty analyses for reactor core calculations. Interim report; Weiterentwicklung moderner Verfahren zu Neutronentransport und Unsicherheitsanalysen fuer Kernberechnungen. Zwischenbericht

    Energy Technology Data Exchange (ETDEWEB)

    Zwermann, Winfried; Aures, Alexander; Bostelmann, Friederike; Pasichnyk, Ihor; Perin, Yann; Velkov, Kiril; Zilly, Matias

    2016-12-15

    This report documents the status of the research and development goals reached within the reactor safety research project RS1536 ''Development of modern methods with respect to neutron transport and uncertainty analyses for reactor core calculations'' as of the 3{sup rd} quarter of 2016. 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, in particular fast reactors cooled by liquid metal. The contributing individual goals are the further optimization and validation of deterministic calculation methods with high spatial and energy resolution, the development of a coupled calculation system using the Monte Carlo method for the neutron transport to describe time-dependent reactor core states, the processing and validation of nuclear data, particularly with regard to covariance data, the development, validation, and application of sampling-based methods for uncertainty and sensitivity analyses, the creation of a platform for performing systematic uncertainty analyses for fast reactor systems, as well as the description of states of severe core damage with the Monte Carlo method. Moreover, work regarding the European NURESAFE project, started in the preceding project RS1503, are being continued and completed.

  1. Comet whole-core solution to a stylized 3-dimensional pressurized water reactor benchmark problem with UO{sub 2}and MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, D.; Rahnema, F. [Georgia Inst. of Technology, 770 State Street, Atlanta, GA 30332-0745 (United States)

    2012-07-01

    A stylized pressurized water reactor (PWR) benchmark problem with UO{sub 2} and MOX fuel was used to test the accuracy and efficiency of the coarse mesh radiation transport (COMET) code. The benchmark problem contains 125 fuel assemblies and 44,000 fuel pins. The COMET code was used to compute the core eigenvalue and assembly and pin power distributions for three core configurations. In these calculations, a set of tensor products of orthogonal polynomials were used to expand the neutron angular phase space distribution on the interfaces between coarse meshes. The COMET calculations were compared with the Monte Carlo code MCNP reference solutions using a recently published an 8-group material cross section library. The comparison showed both the core eigenvalues and assembly and pin power distributions predicated by COMET agree very well with the MCNP reference solution if the orders of the angular flux expansion in the two spatial variables and the polar and azimuth angles on the mesh boundaries are 4, 4, 2 and 2. The mean and maximum differences in the pin fission density distribution ranged from 0.28%-0.44% and 3.0%-5.5%, all within 3-sigma uncertainty of the MCNP solution. These comparisons indicate that COMET can achieve accuracy comparable to Monte Carlo. It was also found that COMET's computational speed is 450 times faster than MCNP. (authors)

  2. 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.

  3. Analysis and development of spatial hp-refinement methods for solving the neutron transport equation; Analyse et developpement de methodes de raffinement hp en espace pour l'equation de transport des neutrons

    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

  4. Coupled multi-group neutron photon transport for the simulation of high-resolution gamma-ray spectroscopy applications

    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.

  5. Three-dimensional whole-core transport calculation of the OECD benchmark problem C5G7 MOX by the CRX code

    Energy Technology Data Exchange (ETDEWEB)

    Gil, Soo Lee; Nam, Zin Cho [Korea Advanced Institute of Science and Technology, Dept. of Nuclear and Quantum Engineering, Yusong-gu, Daejeon (Korea, Republic of)

    2005-07-01

    The OECD 3-dimensional benchmark problem C5G7 MOX was calculated by the CRX code. For 3-dimensional heterogeneous calculation, the CRX code uses a fusion technique of 2-dimensional/1-dimensional methods: the method of characteristics for radial 2-dimensional calculation and diamond difference scheme (DD) that is an S{sub N}-like method for axial 1-dimensional calculation. We improve the fusion method by using a linear characteristics (LC) solver in the 1-dimensional calculation. Here, we present brief structure of 2-dimensional/1-dimensional fusion method and the results of 3 configurations of benchmark problem. We also present results of several different 1-dimensional calculation options. Numerical results show that the LC scheme presents better performance than DD. In the results of the benchmark problem, k(eff) errors are less than 0.05% and the averages of pin power errors are less than 1% for all calculations.

  6. A new approach to three-dimensional neutron transport solution based on the method of characteristics and linear axial approximation

    Science.gov (United States)

    Zheng, Youqi; Choi, Sooyoung; Lee, Deokjung

    2017-12-01

    A new approach based on the method of characteristics (MOC) is proposed to solve the neutron transport equation. A new three-dimensional (3D) spatial discretization is applied to avoid the instability issue of the transverse leakage iteration of the traditional 2D/1D approach. In this new approach, the axial and radial variables are discretized in two different ways: the linear expansion is performed in the axial direction, then, the 3D solution of the angular flux is transformed to be the planar solution of 2D angular expansion moments, which are solved by the planar MOC sweeping. Based on the boundary and interface continuity conditions, the 2D expansion moment solution is equivalently transformed to be the solution of the axially averaged angular flux. Using the piecewise averaged angular flux at the top and bottom surfaces of 3D meshes, the planes are coupled to give the 3D angular flux distribution. The 3D CMFD linear system is established from the surface net current of every 3D pin-mesh to accelerate the convergence of power iteration. The STREAM code is extended to be capable of handling 3D problems based on the new approach. Several benchmarks are tested to verify its feasibility and accuracy, including the 3D homogeneous benchmarks and heterogeneous benchmarks. The computational sensitivity is discussed. The results show good accuracy in all tests. With the CMFD acceleration, the convergence is stable. In addition, a pin-cell problem with void gap is calculated. This shows the advantage compared to the traditional 2D/1D MOC methods.

  7. Development and application of neutron transport methods and uncertainty analyses for reactor core calculations. Technical report; Entwicklung und Einsatz von Neutronentransportmethoden und Unsicherheitsanalysen fuer Reaktorkernberechnungen. Technischer Bericht

    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.

  8. An analytical representation for the solution of neutron kinetic transport equation in slab-geometry multigroup discrete ordinates formulation

    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)

  9. Tracking techniques for the characteristics method applied to the resolution of the neutrons transport equation in multi scale domains; Techniques de tracage pour la methode des caracteristiques appliquee a la resolution de l'equation du transport des neutrons en domaines multi-dimensionnels

    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.)

  10. The Numerical Nuclear Reactor for High-Fidelity Integrated Simulation of Neutronic, Thermal-Hydraulic, and Thermo-Mechanical Phenomena

    Energy Technology Data Exchange (ETDEWEB)

    Kim, K. S.; Ju, H. G.; Jeon, T. H. and others

    2005-03-15

    A comprehensive high fidelity reactor core modeling capability has been developed for detailed analysis of current and advanced reactor designs as part of a US-ROK collaborative I-NERI project. High fidelity was accomplished by integrating highly refined solution modules for the coupled neutronic, thermal-hydraulic, and thermo-mechanical phenomena. Each solution module employs methods and models that are formulated faithfully to the first-principles governing the physics, real geometry, and constituents. Specifically, the critical analysis elements that are incorporated in the coupled code capability are whole-core neutron transport solution, ultra-fine-mesh computational fluid dynamics/heat transfer solution, and finite-element-based thermo-mechanics solution, all obtained with explicit (fuel pin cell level) heterogeneous representations of the components of the core. The vast computational problem resulting from such highly refined modeling is solved on massively parallel computers, and serves as the 'numerical nuclear reactor'. Relaxation of modeling parameters were also pursued to make problems run on clusters of workstations and PCs for smaller scale applications as well.

  11. Neutronic reactor

    Science.gov (United States)

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

    1983-01-01

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  12. Neutronic reactor

    Energy Technology Data Exchange (ETDEWEB)

    Babcock, D.F.; Menegus, R.L.; Wende, C.W.

    1983-01-04

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  13. Sodium Ion Transport Mechanisms in Antiperovskite Electrolytes Na3OBr and Na4OI2: An in Situ Neutron Diffraction Study.

    Science.gov (United States)

    Zhu, Jinlong; Wang, Yonggang; Li, Shuai; Howard, John W; Neuefeind, Jörg; Ren, Yang; Wang, Hui; Liang, Chengdu; Yang, Wenge; Zou, Ruqiang; Jin, Changqing; Zhao, Yusheng

    2016-06-20

    Na-rich antiperovskites are recently developed solid electrolytes with enhanced sodium ionic conductivity and show promising functionality as a novel solid electrolyte in an all solid-state battery. In this work, the sodium ionic transport pathways of the parent compound Na3OBr, as well as the modified layered antiperovskite Na4OI2, were studied and compared through temperature-dependent neutron diffraction combined with the maximum entropy method. In the cubic Na3OBr antiperovskite, the nuclear density distribution maps at 500 K indicate that sodium ions hop within and among oxygen octahedra, and Br(-) ions are not involved. In the tetragonal Na4OI2 antiperovskite, Na ions, which connect octahedra in the ab plane, have the lowest activation energy barrier. The transport of sodium ions along the c axis is assisted by I(-) ions.

  14. An analytical approach for a nodal formulation of a two-dimensional fixed-source neutron transport problem in heterogeneous medium

    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.

  15. Finite element method for neutron transport. Part IV. A variational principle giving an upper bound for the lowest eigenvalue of the Boltzmann Equation

    Energy Technology Data Exchange (ETDEWEB)

    Ackroyd, R.T.

    1978-02-01

    A maximum principle for neutron transport in systems with extraneous sources is used with the method of source iteration to suggest a functional for a variational principle for self-sustaining systems. By using the general properties of the leakage and removal operators of the even-parity transport equation the variational principle is shown to give an upper bound to the lowest eigenvalue of the one-speed Boltzmann equation. Thus by making use of the method of Part III for a lower bound, the lowest eigenvalue can be bracketed. The variational principle leads to the finite element equations identical to those arising in the Williams/Galliara finite element formulation of the source-iteration methods, thus showing that the latter method always gives an upper bound to the lowest eigenvalue. Their upper bounds are very close to the exact value for some benchmark calculations.

  16. Domain decomposition and CMFD acceleration applied to discrete-ordinate methods for the solution of the neutron transport equation in XYZ geometries

    Energy Technology Data Exchange (ETDEWEB)

    Masiello, Emiliano; Martin, Brunella; Do, Jean-Michel, E-mail: emiliano.masiello@cea.fr, E-mail: brunella.martin@gmail.com, E-mail: jean-michel.do@cea.fr [Commissariat a l' Energie Atomique et aux Energies Alternatives, Direction de l' Energie Nucleaire, Service d' Etudes de Reacteurs et de Mathematiques Appliquees, Gif sur Yvette, (France)

    2011-07-01

    A new development for the IDT solver is presented for large reactor core applications in XYZ geometries. The multigroup discrete-ordinate neutron transport equation is solved using a Domain-Decomposition (DD) method coupled with the Coarse-Mesh Finite Differences (CMFD). The later is used for accelerating the DD convergence rate. In particular, the external power iterations are preconditioned for stabilizing the oscillatory behavior of the DD iterative process. A set of critical 2-D and 3-D numerical tests on a single processor will be presented for the analysis of the performances of the method. The results show that the application of the CMFD to the DD can be a good candidate for large 3D full-core parallel applications. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-01-01

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

  18. MRI-guided neutron capture therapy by use of a dual gadolinium/boron agent targeted at tumour cells through upregulated low-density lipoprotein transporters.

    Science.gov (United States)

    Geninatti-Crich, Simonetta; Alberti, Diego; Szabo, Ibolya; Deagostino, Annamaria; Toppino, Antonio; Barge, Alessandro; Ballarini, Francesca; Bortolussi, Silva; Bruschi, Piero; Protti, Nicoletta; Stella, Sabrina; Altieri, Saverio; Venturello, Paolo; Aime, Silvio

    2011-07-18

    The upregulation of low-density lipoprotein (LDL) transporters in tumour cells has been exploited to deliver a sufficient amount of gadolinium/boron/ligand (Gd/B/L) probes for neutron capture therapy, a binary chemio-radiotherapy for cancer treatment. The Gd/B/L probe consists of a carborane unit (ten B atoms) bearing an aliphatic chain on one side (to bind LDL particles), and a Gd(III)/1,4,7,10-tetraazacyclododecane monoamide complex on the other (for detection by magnetic resonance imaging (MRI)). Up to 190 Gd/B/L probes were loaded per LDL particle. The uptake from tumour cells was initially assessed on cell cultures of human hepatoma (HepG2), murine melanoma (B16), and human glioblastoma (U87). The MRI assessment of the amount of Gd/B/L taken up by tumour cells was validated by inductively coupled plasma-mass-spectrometric measurements of the Gd and B content. Measurements were undertaken in vivo on mice bearing tumours in which B16 tumour cells were inoculated at the base of the neck. From the acquisition of magnetic resonance images, it was established that after 4-6 hours from the administration of the Gd/B/L-LDL particles (0.1 and 1 mmol kg(-1) of Gd and (10)B, respectively) the amount of boron taken up in the tumour region is above the threshold required for successful NCT treatment. After neutron irradiation, tumour growth was followed for 20 days by MRI. The group of treated mice showed markedly lower tumour growth with respect to the control group. Copyright © 2011 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  19. Comparative study of Monte Carlo particle transport code PHITS and nuclear data processing code NJOY for recoil cross section spectra under neutron irradiation

    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.

  20. Comparative study of Monte Carlo particle transport code PHITS and nuclear data processing code NJOY for recoil cross section spectra under neutron irradiation

    Science.gov (United States)

    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.

  1. 3D Monte-Carlo transport calculations of whole slab reactor cores: validation of deterministic neutronic calculation routes

    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)

  2. Discontinuous finite element and characteristics methods for neutrons transport equation solution in heterogeneous grids; Resolution de l'equation du transport des neutrons par les methodes des elements finis discontinus et des caracteristiques structurees appliquees a des maillages heterogenes

    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)

  3. Visualizing oxygen anion transport pathways in NdBaCo2O5+δ by in situ neutron diffraction

    Energy Technology Data Exchange (ETDEWEB)

    Cox-Galhotra, Rosemary A. [Univ. of Virginia, Charlottesville, VA (United States). Dept. of Chemical Engineering; Huq, Ashfia [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Spallation Neutron Source; Hodges, Jason P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Spallation Neutron Source; Kim, Jung-Hyun [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Spallation Neutron Source; Yu, Chengfei [Univ. of Houston, TX (United States). Dept. of Chemistry; Wang, Xiqu [Univ. of Houston, TX (United States). Dept. of Chemistry; Jacobson, Allan J. [Univ. of Houston, TX (United States). Dept. of Chemistry; McIntosh, Steven [Univ. of Virginia, Charlottesville, VA (United States). Dept. of Chemical Engineering; Lehigh Univ., Bethlehem, PA (United States). Dept. of Chemical Engineering

    2013-01-14

    We characterized the layered perovskite NdBaCo2O5+δ (NBCO) using neutron powder diffraction under in situ conditions from 577–852 °C and in 10-1 to 10-4 atm oxygen. The best fit to the data was obtained in the tetragonal (P4/mmm) space group. No oxygen atom vacancy ordering was observed that warranted a lowering of the symmetry to orthorhombic (Pmmm). Two P4/mmm structural models were investigated: Model 1 (no split sites) and Model 2 (split Nd and O2 sites). Furthermore, transport of oxygen through the material via the vacancy hopping mechanism will likely involve the nearest-neighbor oxygen atom sites in the Nd layer. Total oxygen stoichiometry values were in the range 5.51 ≤ δ ≤ 5.11. The tetragonal lattice parameters increased with temperature as expected. But, the a-axis expands while the c-axis contracts with decreasing pO2 at a given temperature due to increasing vacancy concentration in the Nd layer.

  4. Calculations of reactivity based in the solution of the Neutron transport equation in X Y geometry and Lineal perturbation theory; Calculos de reactividad basados en la solucion de la Ecuacion de transporte de neutrones en geometria XY y Teoria de perturbacion lineal

    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)

  5. Analysis of space and energy homogenization techniques for the solving of the neutron transport equation in nuclear reactor; Analyse des techniques d'homogeneisation spatiale et energetique dans la resolution de l'equation du transport des neutrons dans les reacteurs nucleaires

    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.)

  6. Neutron detector

    Science.gov (United States)

    Stephan, Andrew C [Knoxville, TN; Jardret,; Vincent, D [Powell, TN

    2011-04-05

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

  7. Fusion neutronics

    CERN Document Server

    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...

  8. Solution of the neutron transport equation by the collision probability for 3D geometries; Resolution de l`equation du transport pour les neutrons par la methode des probabilites de collision dans le geometries 3D

    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.

  9. Neutron structural characterization, inversion degree and transport properties of NiMn{sub 2}O{sub 4} spinel prepared by the hydroxide route

    Energy Technology Data Exchange (ETDEWEB)

    Sagua, A.; Lescano, Gabriela M. [Departamento de Química, Laboratorio de Fisicoquímica Inorgánica, Universidad Nacional del Sur, INQUISUR, 8000 Bahía Blanca (Argentina); Alonso, J.A., E-mail: jaalonso@icmm.csic.es [Instituto de Ciencia de Materiales de Madrid, C.S.I.C., Cantoblanco, E-28049 Madrid (Spain); Martínez-Coronado, R. [Instituto de Ciencia de Materiales de Madrid, C.S.I.C., Cantoblanco, E-28049 Madrid (Spain); Fernández-Díaz, M.T. [Institut Laue Langevin, BP 156X, Grenoble F-38042 (France); Morán, E. [Departamento de Química Inorgánica, Facultad de Ciencias Químicas, Universidad Complutense, 28040 Madrid (Spain)

    2012-06-15

    Graphical abstract: A pure specimen has been synthesized by the hydroxide route. This spinel, studied by NPD, shows an important inversion degree, λ = 0.80. A bond-valence study shows that the tetrahedral Mn ions are divalent whereas the octahedral Mn and Ni are slightly oxidized from the expected 3+ and 2+ values, respectively. The mixed valence Mn{sup 3+}/Mn{sup 4+} accounts for a hopping mechanism between adjacent octahedral sites, leading to a significant conductivity. Highlights: ► A low-temperature hydroxide route allowed preparing almost pure specimens of NiMn{sub 2}O{sub 4}. ► NPD essential to determine inversion degree; contrasting Ni and Mn for neutrons. ► Bond valence establishes valence state of octahedral and tetrahedral Ni and Mn ions. ► Thermal analysis, transport measurements complement characterization of this oxide. ► A structure–properties relationship is established. -- Abstract: The title compound has been synthesized by the hydroxide route. The crystal structure has been investigated at room temperature from high-resolution neutron powder diffraction (NPD) data. It crystallizes in a cubic spinel structure, space group Fd3{sup ¯}m, Z = 8, with a = 8.3940(2) Å at 295 K. The crystallographic formula is (Ni{sub 0.202(1)}Mn{sub 0.798(1)}){sub 8a}(Ni{sub 0.790(1)}Mn{sub 1.210(1)}){sub 16d}O{sub 4} where 8a and 16d stand for the tetrahedral and octahedral sites of the spinel structure, respectively. There is a significant inversion degree of the spinel structure, λ = 0.80. In fact, the variable parameter for the oxygen position, u = 0.2636(4), is far from that expected (u = 0.25) for normal spinels. From a bond-valence study, it seems that the valence distribution in NiMn{sub 2}O{sub 4} spinel is not as trivial as expected (Ni{sup 2+} and Mn{sup 3+}), but clearly the tetrahedral Mn ions are divalent whereas the octahedral Mn and Ni are slightly oxidized from the expected +3 and +2 values, respectively. The mixed valence observed at

  10. Neutron Stars

    Science.gov (United States)

    Cottam, J.

    2007-01-01

    Neutron stars were discovered almost 40 years ago, and yet many of their most fundamental properties remain mysteries. There have been many attempts to measure the mass and radius of a neutron star and thereby constrain the equation of state of the dense nuclear matter at their cores. These have been complicated by unknown parameters such as the source distance and burning fractions. A clean, straightforward way to access the neutron star parameters is with high-resolution spectroscopy. I will present the results of searches for gravitationally red-shifted absorption lines from the neutron star atmosphere using XMM-Newton and Chandra.

  11. Neutron tubes

    Science.gov (United States)

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

    2008-03-11

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

  12. 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....

  13. New Multi-group Transport Neutronics (PHISICS) Capabilities for RELAP5-3D and its Application to Phase I of the OECD/NEA MHTGR-350 MW Benchmark

    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.

  14. Transportation

    Science.gov (United States)

    2006-01-01

    container. It now permits free transit of shipping containers from their western ports, if transported by rail directly to the U.S. ( Mireles , 2005, p...Transportation Industry Study Seminar. Mireles , Richard, Castillo. (2005, January). A Cure for West Coast Congestion. Logistics Today, Vol. 46, Issue 1. 1

  15. RCPO1 - A Monte Carlo program for solving neutron and photon transport problems in three dimensional geometry with detailed energy description and depletion capability

    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.

  16. Cyclotron-based neutron source for BNCT

    Science.gov (United States)

    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.

  17. 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.

  18. 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.

  19. Neutron fluence in antiproton radiotherapy, measurements and simulations

    DEFF Research Database (Denmark)

    Bassler, Niels; Holzscheiter, Michael H.; Petersen, Jørgen B.B.

    2010-01-01

    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...

  20. Transportes

    Directory of Open Access Journals (Sweden)

    Hidalgo Fernández-Cano, Amalio

    1960-01-01

    Full Text Available El movimiento de materiales dentro de la Factoría está atendido por tres principales medios de transporte, en consonancia con las características del material y de los desplazamientos. Así se han establecido: sistemas de cintas transportadoras, una red ferroviaria de ancho normal y una completa malla de caminos enlazando funcionalmente las instalaciones.

  1. Sodium Ion Transport Mechanisms in Antiperovskite Electrolytes Na 3 OBr and Na 4 OI 2 : An in Situ Neutron Diffraction Study

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, Jinlong; Wang, Yonggang; Li, Shuai; Howard, John W.; Neuefeind, Jörg; Ren, Yang; Wang, Hui; Liang, Chengdu; Yang, Wenge; Zou, Ruqiang; Jin, Changqing; Zhao, Yusheng

    2016-06-20

    Na-rich antiperovskites are recently developed solid electrolytes with enhanced sodium ionic conductivity and show promising functionality as a novel solid electrolyte in an all solid-state battery. In this work, the sodium ionic transport pathways of the parent compound Na3OBr, as well as the modified layered antiperovskite Na4OI2, were studied and compared through temperature dependent neutron diffraction combined with the maximum entropy method. In the cubic Na3OBr antiperovskite, the nuclear density distribution maps at 500 K indicate that sodium ions hop within and among oxygen octahedra, and Br- ions are not involved. In the tetragonal Na4OI2 antiperovskite, Na ions, which connect octahedra in the ab plane, have the lowest activation energy barrier. The transport of sodium ions along the c axis is assisted by I- ions.

  2. Neutron diffraction

    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.

  3. Neutronics of pulsed spallation neutron sources

    CERN Document Server

    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...

  4. Anomalous hydrodynamics kicks neutron stars

    Energy Technology Data Exchange (ETDEWEB)

    Kaminski, Matthias, E-mail: mski@ua.edu [Department of Physics and Astronomy, University of Alabama, Tuscaloosa, AL 35487 (United States); Department of Physics and Astronomy, University of Victoria, Victoria, BC, V8P 5C2 (Canada); Uhlemann, Christoph F. [Department of Physics, University of Washington, Seattle, WA 98195 (United States); Bleicher, Marcus [Frankfurt Institute for Advanced Studies, Goethe-Universität Frankfurt (Germany); Institut für Theoretische Physik, Goethe Universität Frankfurt (Germany); Schaffner-Bielich, Jürgen [Institut für Theoretische Physik, Goethe Universität Frankfurt (Germany)

    2016-09-10

    Observations show that, at the beginning of their existence, neutron stars are accelerated briskly to velocities of up to a thousand kilometers per second. We argue that this remarkable effect can be explained as a manifestation of quantum anomalies on astrophysical scales. To theoretically describe the early stage in the life of neutron stars we use hydrodynamics as a systematic effective-field-theory framework. Within this framework, anomalies of the Standard Model of particle physics as underlying microscopic theory imply the presence of a particular set of transport terms, whose form is completely fixed by theoretical consistency. The resulting chiral transport effects in proto-neutron stars enhance neutrino emission along the internal magnetic field, and the recoil can explain the order of magnitude of the observed kick velocities.

  5. Neutron activation system for spectral measurements of pulsed ion diode neutron production

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, D.L.; Kruse, L.W.

    1980-02-01

    A neutron energy spectrometer has been developed to study intense ion beam-target interactions in the harsh radiation environment of a relativistic electron beam source. The main component is a neutron threshold activation system employing two multiplexed high efficiency Ge(Li) detectors, an annihilation gamma coincidence system, and a pneumatic sample transport. Additional constraints on the neutron spectrum are provided by total neutron yield and time-of-flight measurements. A practical lower limit on the total neutron yield into 4..pi.. required for a spectral measurement with this system is approx. 10/sup 10/ n where the neutron yield is predominantly below 4 MeV and approx. 10/sup 8/ n when a significant fraction of the yield is above 4 MeV. Applications of this system to pulsed ion diode neutron production experiments on Hermes II are described.

  6. Neutron cross sections for fusion. [Review

    Energy Technology Data Exchange (ETDEWEB)

    Haight, R.C.

    1979-10-01

    First generation fusion reactors will most likely be based on the /sup 3/H(d,n)/sup 4/He reaction, which produces 14-MeV neutrons. In these reactors, both the number of neutrons and the average neutron energy will be significantly higher than for fission reactors of the same power. Accurate neutron cross section data are therefore of great importance. They are needed in present conceptual designs to calculate neutron transport, energy deposition, nuclear transmutation including tritium breeding and activation, and radiation damage. They are also needed for the interpretation of radiation damage experiments, some of which use neutrons up to 40 MeV. In addition, certain diagnostic measurements of plasma experiments require nuclear cross sections. The quality of currently available data for these applications will be reviewed and current experimental programs will be outlined. The utility of nuclear models to provide these data also will be discussed. 65 references.

  7. Methods for absorbing neutrons

    Science.gov (United States)

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

    2012-07-24

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

  8. Tracking techniques for the method of characteristics applied to the neutron transport problem in multi-dimensional domains; Techniques de tracage pour la methode des caracteristiques appliquee a la resolution de l'equation du transport des neutrons en domaines multi-dimensionnels

    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)

  9. Subcritical neutron production using multiple accelerators

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, W.Y.; Jones, J.L. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Harmon, J.F. [Idaho State Univ., Pocatello, ID (United States)

    1994-12-31

    A subcritical neutron production technique using multiple accelerators is being developed to provide a selective alternative (for small volumes) to nuclear reactor neutron production. The concept combines the capabilities of multiple commercially-available linear accelerators and a compact subcritical assembly design to generate reactor-like thermal neutron fluxes (i.e., 10{sup 13}-10{sup 14} n/cm{sup 2}/s) in small irradiation volumes of up to 500 cm{sup 3}. In addition, fast and epithermal neutron fluxes will also be available. The neutron source utilizes radially-oriented, pulsed, electron accelerators. The subcritical neutron production assembly is in the form of a compact right-cylinder (approximately 20-cm dia.). This assembly uses an outer ring of graphite (i.e., reflector) with re-entrant holes to enable penetration of the electron beam to the internal structure which comprises of uranium as an electron convertor/neutron multiplier followed by H{sub 2}O beryllium, H{sub 2}O aluminum, and D{sub 2}O in succession toward the center. The inner-most region filled with D{sub 2}O is the central irradiation volume. The material configuration and overall design is to maximize thermal neutron fluxes in the central irradiation volume based on photoneutron/photofission and neutron multiplication processes as well as neutron transport. This assembly will be designed not to reach a nuclear critical state under any normal and/or accidental condition.

  10. Basic of Neutron NDA

    Energy Technology Data Exchange (ETDEWEB)

    Trahan, Alexis Chanel [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-09-15

    The objectives of this presentation are to introduce the basic physics of neutron production, interactions and detection; identify the processes that generate neutrons; explain the most common neutron mechanism, spontaneous and induced fission and (a,n) reactions; describe the properties of neutron from different sources; recognize advantages of neutron measurements techniques; recognize common neutrons interactions; explain neutron cross section measurements; describe the fundamental of 3He detector function and designs; and differentiate between passive and active assay techniques.

  11. Neutronic analysis of a high power density hybrid reactor using ...

    Indian Academy of Sciences (India)

    Natural lithium is also utilized for comparison to these two innovative coolants. Neutron transport calculations are performed on a simple experimental hybrid blanket with cylindrical geometry with the help of the SCALE 4·3 System by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups ...

  12. Spallation Neutron Source (SNS)

    Data.gov (United States)

    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...

  13. REFRACTIVE NEUTRON LENS

    OpenAIRE

    Petrov, P. V.; Kolchevsky, N. N.

    2013-01-01

    Compound concave refractive lenses are used for focusing neutron beam. Investigations of spectral and focusing properties of a refractive neutron lens are presented. Resolution of the imaging system on the base of refractive neutron lenses depends on material properties and parameters of neutron source. Model of refractive neutron lens are proposed. Results of calculation diffraction resolution and focal depth of refractive neutron lens are discussed.

  14. Neutron scattering. Lectures

    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)

  15. Neutron Therapy Facility

    Data.gov (United States)

    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)...

  16. 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.

  17. Development of compact accelerator neutron source

    Science.gov (United States)

    Letourneau, Alain; Marchix, Anthony; Tran, Ngoc-Hoang; Chauvin, Nicolas; Menelle, Alain; Ott, Frédéric; Schwindling, Jérôme

    2017-09-01

    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.

  18. Neutron dosimetry at the reactor facility VENUS

    Energy Technology Data Exchange (ETDEWEB)

    Deboodt, P.; Vermeersch, F.; Vanhavere, F.; Minsart, G. [SCK-CEN, Belgian Nuclear Research Centre, Mol (Belgium)

    1997-09-01

    The reactor VENUS is a zero-power research reactor mainly devoted to studies on light water fuels. The need for undertaking a neutron spectrometric and dosimetric study became apparent when locally high neutron dose rates were measured. The spectrometric study is based on two approaches. The first is an experimental one in which the neutron spectrum was measured at three positions around the facility. The second is a theoretical one in which a numerical modelling of the neutron transport at the reactor site was performed in order to determine neutron spectra and fluence rates at different positions around the site. The measured and calculated spectra are interpreted in terms of the responses of different individual and environmental dosemeters. These responses are confronted with the in situ measurements. The impact of the ICRP 60 recommendations on the determined dose rates is also studied. (author).

  19. 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.

  20. Development of neutron shielding material for cask

    Energy Technology Data Exchange (ETDEWEB)

    Najima, K.; Ohta, H.; Ishihara, N. [Mitsubishi Heavy Industries Ltd., Takasago Research and Development Center, Hyogo (Japan); Matsuoka, T.; Kuri, S.; Ohsono, K.; Hode, S. [Mitsubishi Heavy Industries Ltd., Kobe Shipyard and Machinery Works, Kobe (Japan)

    2001-07-01

    Since 1980's Mitsubishi Heavy Industries, Ltd (MHI) has established transport and storage cask design 'MSF series' which makes higher payload and reliability for long term storage. MSF series transport and storage cask uses new-developed neutron shielding material. This neutron shielding material has been developed for improving durability under high condition for long term. Since epoxy resin contains a lot of hydrogen and is comparatively resistant to heat, many casks employ epoxy base neutron shielding material. However, if the epoxy base neutron shielding material is used under high temperature condition for a long time, the material deteriorates and the moisture contained in it is released. The loss of moisture is in the range of several percents under more than 150 C. For this reason, our purpose was to develop a high durability epoxy base neutron shielding material which has the same self-fire-extinction property, high hydrogen content and so on as conventional. According to the long-time heating test, the weight loss of this new neutron shielding material after 5000 hours heating has been lower than 0.04% at 150 C and 0.35% at 170 C. A thermal test was also performed: a specimen of neutron shielding material covered with stainless steel was inserted in a furnace under condition of 800 C temperature for 30 minutes then was left to cool down in ambient conditions. The external view of the test piece shows that only a thin layer was carbonized.

  1. 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.

  2. Superfluid neutron stars

    OpenAIRE

    Langlois, David

    2001-01-01

    Neutron stars are believed to contain (neutron and proton) superfluids. I will give a summary of a macroscopic description of the interior of neutron stars, in a formulation which is general relativistic. I will also present recent results on the oscillations of neutron stars, with superfluidity explicitly taken into account, which leads in particular to the existence of a new class of modes.

  3. Neutron scattering. Lectures

    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)

  4. Nuclear reactor neutron shielding

    Science.gov (United States)

    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.

  5. NEUTRONIC REACTORS

    Science.gov (United States)

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

    1958-10-14

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

  6. Weapons Neutron Research Facility (WNR)

    Data.gov (United States)

    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...

  7. Study of neutron generation in the compact tokamak TUMAN-3M in support of a tokamak-based fusion neutron source

    Science.gov (United States)

    Kornev, V. A.; Askinazi, L. G.; Belokurov, A. A.; Chernyshev, F. V.; Lebedev, S. V.; Melnik, A. D.; Shabelsky, A. A.; Tukachinsky, A. S.; Zhubr, N. A.

    2017-12-01

    The paper presents DD neutron flux measurements in neutron beam injection (NBI) experiments aimed at the optimization of target plasma and heating beam parameters to achieve maximum neutron flux in the TUMAN-3M compact tokamak. Two ion sources of different design were used, which allowed the separation of the beam’s energy and power influence on the neutron rate. Using the database of experiments performed with the two ion sources, an empirical scaling was derived describing the neutron rate dependence on the target plasma and heating beam parameters. Numerical modeling of the neutron rate in the NBI experiments performed using the ASTRA transport code showed good agreement with the scaling.

  8. Impact of nuclear data on fast neutron therapy

    Energy Technology Data Exchange (ETDEWEB)

    Hartmann Siantar, C.L.; Chandler, W.P.; Rathkopf, J.A.; Resler, D.A.; Cox, L.J.; Chadwick, M.B.; White, R.M.

    1994-05-12

    By combining a new, all-particle Monte Carlo radiation transport code, PEREGRINE, with the Lawrence Livermore National Laboratory (LLNL) nuclear data base, we have studied the importance of various neutron reactions on dose distributions in biological materials. Monte Carlo calculations have been performed for 5--20 MeV neutron pencil beams incident on biologically relevant materials arranged in several simple geometries. Results highlight the importance of nuclear data used for calculating dose distributions resulting from fast neutron therapy.

  9. Layered semiconductor neutron detectors

    Science.gov (United States)

    Mao, Samuel S; Perry, Dale L

    2013-12-10

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

  10. Neutron streak camera

    Science.gov (United States)

    Wang, Ching L.

    1983-09-13

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

  11. Neutron in biology

    Energy Technology Data Exchange (ETDEWEB)

    Niimura, Nobuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-11-01

    Neutron in biology can provide an experimental method of directly locating relationship of proteins and DNA. However, there are relatively few experimental study of such objects since it takes a lot of time to collect a sufficient number of Bragg reflections and inelastic spectra due to the low flux of neutron illuminating the sample. Since a next generation neutron source of JAERI will be 5MW spallation neutron source and its effective neutron flux will be 10{sup 2} to 10{sup 3} times higher than the one of JRR-3M, neutron in biology will open a completely new world for structural biology. (author)

  12. Neutronic Reactor Design to Reduce Neutron Loss

    Science.gov (United States)

    Miles, F. T.

    1961-05-01

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

  13. Basics of Neutrons for First Responders

    Energy Technology Data Exchange (ETDEWEB)

    Rees, Brian G. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2018-02-05

    These are slides from a presentation on the basics of neutrons. A few topics covered are: common origins of terrestrial neutron radiation, neutron sources, neutron energy, interactions, detecting neutrons, gammas from neutron interactions, neutron signatures in gamma-ray spectra, neutrons and NaI, neutron fluence to dose (msV), instruments' response to neutrons.

  14. Numerical solution of uncertain neutron diffusion equation for ...

    Indian Academy of Sciences (India)

    Department of Mathematics, National Institute of Technology, ... The concept of fuzziness is hybridised with traditional finite element method to propose fuzzy finite element method. The proposed fuzzy finite element method has ..... In general when neutrons undergo scattering, the neutron transport equation involves uncer-.

  15. Report by the AERES on the unit: Unit of researches on neutron transport and radioactivity confinement in nuclear installations under the supervision of the establishments and bodies: IRSN; Rapport de l'AERES sur l'unite: Unite de recherches en neutronique et confinement de la radioactivite dans les installations nucleaires sous tutelle des etablissements et organismes: IRSN

    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

  16. A geometry preserving, conservative, mesh-to-mesh isogeometric interpolation algorithm for spatial adaptivity of the multigroup, second-order even-parity form of the neutron transport equation

    Science.gov (United States)

    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.

  17. Neutron anatomy

    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.

  18. Neutrons from Antiproton Irradiation

    DEFF Research Database (Denmark)

    Bassler, Niels; Holzscheiter, Michael; Petersen, Jørgen B.B.

    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...

  19. Neutron flux assessment of a neutron irradiation facility based on inertial electrostatic confinement fusion.

    Science.gov (United States)

    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. Copyright © 2015 Elsevier Ltd. All rights reserved.

  20. Investigation of Isfahan miniature neutron source reactor (MNSR for boron neutron capture therapy by MCNP simulation

    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.

  1. 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)

  2. 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...

  3. 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.

  4. 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.

  5. Grazing Incidence Neutron Optics

    Science.gov (United States)

    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.

  6. 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)

  7. Neutron scatter camera

    Science.gov (United States)

    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.

  8. Development of the neutron-transport code TransRay and studies on the two- and three-dimensional calculation of effective group cross sections; Entwicklung des Neutronentransportcodes TransRay und Untersuchungen zur zwei- und dreidimensionalen Berechnung effektiver Gruppenwirkungsquerschnitte

    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

  9. 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)

  10. Neutron structural biology

    Energy Technology Data Exchange (ETDEWEB)

    Niimura, Nobuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Neutron diffraction provides an experimental method of directly locating hydrogen atoms in protein which play important roles in physiological functions. However, there are relatively few examples of neutron crystallography in biology since it takes a lot of time to collect a sufficient number of Bragg reflections due to the low flux of neutrons illuminating the sample. In order to overcome the flux problem, we have successfully developed the neutron IP, where the neutron converter, {sup 6}Li or Gd, was mixed with a photostimulated luminescence material on flexible plastic support. Neutron Laue diffraction 2A data from tetragonal lysozyme were collected for 10 days with neutron imaging plates, and 960 hydrogen atoms in the molecule and 157 bound water molecules were identified. These results explain the proposed hydrolysis mechanism of the sugar by the lysozyme molecule and that lysozyme is less active at pH7.0. (author)

  11. Atlas of neutron resonances

    CERN Document Server

    Mughabghab, Said

    2018-01-01

    Atlas of Neutron Resonances: Resonance Properties and Thermal Cross Sections Z= 1-60, Sixth Edition, contains an extensive list of detailed individual neutron resonance parameters for Z=1-60, as well as thermal cross sections, capture resonance integrals, average resonance parameters and a short survey of the physics of thermal and resonance neutrons. The long introduction contains: nuclear physics formulas aimed at neutron physicists; topics of special interest such as valence neutron capture, nuclear level density parameters, and s-, p-, and d-wave neutron strength functions; and various comparisons of measured quantities with the predictions of nuclear models, such as the optical model. As in the last edition, additional features have been added to appeal to a wider spectrum of users. These include: spin-dependent scattering lengths that are of interest to solid-state physicists, nuclear physicists and neutron evaluators; calculated and measured Maxwellian average 5-keV and 30-keV capture cross sections o...

  12. Broad Energy Range Neutron Spectroscopy using a Liquid Scintillator and a Proportional Counter: Application to a Neutron Spectrum Similar to that from an Improvised Nuclear Device.

    Science.gov (United States)

    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)3He and D(d,n)3He, 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 9Be(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.

  13. Neutron total cross section calculation within the framework of quasi-harmonic approximation

    DEFF Research Database (Denmark)

    Cai, Xiao Xiao; Klinkby, Esben Bryndt

    2017-01-01

    The accuracy of neutron scattering cross sections is the gauge for the realistic outcome of a neutron transport simulation. To improve the traditional harmonic physics model used in such simulations, we revisit the slow neutron transport theory in crystalline materials and aim to develop a unified...... model that has good performance for neutron transport problems in crystals in a wide range of temperatures and pressures. The quasi-harmonic approximation (QHA) correlates phonon evolution explicitly with unit cell volume. Therefore, it is capable of evaluating a variety of material properties at finite...

  14. NEUTRON DENSITY CONTROL IN A NEUTRONIC REACTOR

    Science.gov (United States)

    Young, G.J.

    1959-06-30

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

  15. Average neutron detection efficiency for DEMON detectors

    Science.gov (United States)

    Zhang, S.; Lin, W.; Rodrigues, M. R. D.; Huang, M.; Wada, R.; Liu, X.; Zhao, M.; Jin, Z.; Chen, Z.; Keutgen, T.; Kowalski, S.; Hagel, K.; Barbui, M.; Bonasera, A.; Bottosso, C.; Materna, T.; Natowitz, J. B.; Qin, L.; Sahu, P. K.; Schmidt, K. J.; Wang, J.

    2013-05-01

    The neutron detection efficiency of a DEMON detector, averaged over the whole volume, was calculated using GEANT and applied to determine neutron multiplicities in an intermediate heavy ion reaction. When a neutron source is set at a distance of about 1 m from the front surface of the detector, the average efficiency, ɛav, is found to be significantly lower (20-30%) than the efficiency measured at the center of the detector, ɛ0. In the GEANT simulation the ratio R=ɛav/ɛ0 was calculated as a function of neutron energy. The experimental central efficiency multiplied by R was then used to determine the average efficiency. The results were applied to a study of the 64Zn+112Sn reaction at 40 A MeV which employed 16 DEMON detectors. The neutron multiplicity was extracted using a moving source fit. The derived multiplicities are compared well with those determined using the neutron ball in the NIMROD detector array in a separate experiment. Both are in good agreement with multiplicities predicted by a transport model calculation using an antisymmetric molecular dynamics (AMD) model code.

  16. 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)

  17. Ionizing radiation transport in air

    Energy Technology Data Exchange (ETDEWEB)

    Klimanov, V.A.; Konovalov, S.A.; Kochanov, V.A.; Ksenofontov, A.I.; Kukhtevich, V.I.; Mashkovich, V.P.; Meshcherin, B.N.; Panchenko, A.M.; Sukhoruchkin, A.K.; Trubnikov, A.I.

    1979-01-01

    The book considers the problems of gamma-ray and neutron transport in air and near the air-ground and air-vacuum interface. New modifications of the Monte-Carlo method for solving these problems are presented. Differential and integral characteristics of gamma-ray, neutron, and secondary gamma-ray transport are given and the time characteristics of radiation sources are analyzed.

  18. Upgrades to the ultracold neutron source at the Los Alamos Neutron Science Center

    Science.gov (United States)

    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.

  19. 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.

  20. Isolated Neutron Stars

    Directory of Open Access Journals (Sweden)

    Popov S.

    2010-10-01

    Full Text Available Several aspects related to astrophysics of isolated neutron stars are discussed. We start with an introduction into the “new zoo” of young isolated neutron stars. In addition to classical radio pulsars, now we know several species (soft gamma-ray repeators, anomalous X-ray pulsars, central compact objects in supernova remnants, close-by cooling neutron stars - aka “Magnificent seven”, - RRATs, and some others. All these types are briefly discussed. In the second lecture a description of magneto-rotational evolution of neutron stars is given. Finally, in the third lecture we discuss population synthesis of isolated neutron stars. In some details we discuss population synthesis of young isolated radio pulsars and young close-by cooling neutron stars.

  1. 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 235U 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.

  2. The DIORAMA Neutron Emitter

    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.

  3. Neutron structural biology

    Energy Technology Data Exchange (ETDEWEB)

    Niimura, Nobuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-03-01

    Neutron structural biology will be one of the most important fields in the life sciences which will interest human beings in the 21st century because neutrons can provide not only the position of hydrogen atoms in biological macromolecules but also the dynamic molecular motion of hydrogen atoms and water molecules. However, there are only a few examples experimentally determined at present because of the lack of neutron source intensity. Next generation neutron source scheduled in JAERI (Performance of which is 100 times better than that of JRR-3M) opens the life science of the 21st century. (author)

  4. 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.

  5. Neutron Stars: Formation and Structure

    OpenAIRE

    Kutschera, Marek

    1998-01-01

    A short introduction is given to astrophysics of neutron stars and to physics of dense matter in neutron stars. Observed properties of astrophysical objects containing neutron stars are discussed. Current scenarios regarding formation and evolution of neutron stars in those objects are presented. Physical principles governing the internal structure of neutron stars are considered with special emphasis on the possible spin ordering in the neutron star matter.

  6. Cold neutron interferometry

    Science.gov (United States)

    Kitaguchi, Masaaki

    2009-10-01

    Neutron interferometry is a powerful technique for studying fundamental physics. A large dimensional interferometer for long wavelength neutrons is extremely important in order to investigate problems of fundamental physics, including tests of quantum measurement theories and searches for non-Newtonian effects of gravitation, since the sensitivity of interferometer depends on the wavelength and the interaction length. Neutron multilayer mirrors enable us to develop the large scale interferometer for long wavelength neutrons. The multilayer mirror is one of the most useful devices in cold neutron optics. A multilayer of two materials with different potentials is understood as a one-dimensional crystal, which is suitable for Bragg reflection of long wavelength neutrons. Cold and very cold neutrons can be utilized for the interferometer by using the multilayer mirrors with the proper lattice constants. Jamin-type interferometer by using beam splitting etalons (BSEs) has shown the feasibility of the development of large scale interferometer, which enables us to align the four independent mirrors within required precision. The BSE contains two parallel multilayer mirrors. A couple of the BSEs in the Jamin-type interferometer separates and recombines the two paths spatially. Although the path separation was small at the first test, now we have already demonstrated the interferometer with perfectly separated paths. This has confirmed that the multilayer mirrors cause no serious distortion of wave front to compose a interferometer. Arranging such mirrors, we are capable of establishing even a Mach-Zehnder type with much larger size. The interferometer using supermirrors, which reflects the wide range of the wavelength of neutrons, can increase the neutron counts for high precision measurements. We are planning the experiments using the interferometer both for the very cold neutrons and for the pulsed neutrons including J-PARC.

  7. Linear extended neutron diffusion theory for semi-in finites homogeneous means; Teoria de difusion de neutrones lineal extendida para medios homogeneos semi-infinitos

    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)

  8. 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)

  9. Neutron Multiplicity Analysis

    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.

  10. 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 MC2-3, 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.

  11. Development of a Boron Neutron Capture Enhanced Fast Neutron Therapy Beam

    Energy Technology Data Exchange (ETDEWEB)

    Sweezy, Jeremy Ed [Georgia Tech

    2002-01-01

    The combination of fast neutron therapy and boron neutron capture therapy is currently under investigation at several fast neutron therapy centers worldwide. This treatment method, termed boron neutron capture enhanced fast neutron therapy (BNCEFNT) utilizes a boron containing drug to selectively increase the dose to the target tumor. BNCEFNT may be useful in the treatment of some radioresistant brain tumors, such as glioblastoma multiforme. A neutron therapy beam for boron neutron capture enhanced fast neutron therapy has been developed for the existing Fermilab Neutron Therapy Facility. This beam produces a significant dose enhancement due to the the boron neutron capture reaction. The beam was developed by designing a filter and collimator system using the Monte Carlo radiation transport code, MCNPX. The MCNPX code was benchmarked against depth-dose measurements of the standard treatment beam. The new BNCEFNT beam is filtered with 18.3-cm of low carbon steel and is collimated with steel. Measurements of the dose enhancement of the new BNCEFNT beam were performed with paired tissue equivalent ion chambers. One of the ion chambers has boron incorporated in the wall of the chamber to measure the dose due to boron neutron capture. The measured boron dose enhancement of the BNCEFNT beam is (16.3 ± 2.6)% per 100-ppm 10B for a 20-cm diameter beam and (10.0 ± 1.6)% per 100-ppm 10B for a 10-cm diameter beam. The dose rate of the new beam is reduced to 4.4% of the dose rate of the standard treatment beam. xxi A conceptual design that overcomes the reduced dose rate is also presented. This design uses a tungsten collimator placed near the patient, with a 1.5-cm tungsten filter just upstream of the collimator. Using graphite moderation of neutrons around the patient a percent dose enhancement of 15% can be attained with good collimation, for field sizes as small as 5 × 5 cm2 , and without a reduction in dose rate.

  12. Neutron transport analysis for nuclear reactor design

    Science.gov (United States)

    Vujic, J.L.

    1993-11-30

    Replacing regular mesh-dependent ray tracing modules in a collision/transfer probability (CTP) code with a ray tracing module based upon combinatorial geometry of a modified geometrical module (GMC) provides a general geometry transfer theory code in two dimensions (2D) for analyzing nuclear reactor design and control. The primary modification of the GMC module involves generation of a fixed inner frame and a rotating outer frame, where the inner frame contains all reactor regions of interest, e.g., part of a reactor assembly, an assembly, or several assemblies, and the outer frame, with a set of parallel equidistant rays (lines) attached to it, rotates around the inner frame. The modified GMC module allows for determining for each parallel ray (line), the intersections with zone boundaries, the path length between the intersections, the total number of zones on a track, the zone and medium numbers, and the intersections with the outer surface, which parameters may be used in the CTP code to calculate collision/transfer probability and cross-section values. 28 figures.

  13. Neutronics calculation of an heterogeneous compact and thermal core by means of deterministic and stochastic transport theory. Application to the experimental reactor of the University of Strasbourg; Modelisation neutronique d`un coeur thermique compact et heterogene en theorie du transport deterministe et probabiliste. Application au reacteur experimental de l`Universite de Strasbourg

    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)

  14. Neutron Stars and Pulsars

    CERN Document Server

    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...

  15. 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.

  16. Neutron production in neutron-induced reactions at 96 MeV on 56Fe and 208Pb

    Science.gov (United States)

    Sagrado García, I. C.; Lecolley, J. F.; Lecolley, F. R.; Blideanu, V.; Ban, G.; Fontbonne, J. M.; Itis, G.; Lecouey, J. L.; Lefort, T.; Marie, N.; Steckmeyer, J. C.; Le Brun, C.; Blomgren, J.; Johansson, C.; Klug, J.; Orhn, A.; Mermod, P.; Olsson, N.; Pomp, S.; Osterlund, M.; Tippawan, U.; Prokofiev, A. V.; Nadel-Turonski, P.; Fallot, M.; Foucher, Y.; Guertin, A.; Haddad, F.; Vatre, M.

    2011-10-01

    Double-differential cross sections for neutron production were measured in 96-MeV neutron-induced reactions at The Svedberg Laboratory in Uppsala, Sweden. Measurements for Fe and Pb targets were performed using two independent setups: DECOI-DEMON, time-of-flight telescope dedicated to the detection of emitted neutrons with energies between a few and 50MeV and CLODIA-SCANDAL device devoted to measuring emitted neutrons with energies above 40MeV. Double-differential cross sections were measured for an angular range between 15 and 98 deg and with low-energy thresholds (≈2 MeV). Angular and energy distributions and total neutron emission cross sections have been obtained from those measurements. Results have been compared with predictions given by different models included in several transport codes (MCNPX, GEANT, TALYS, PHITS, and DYWAN) and with other experimental data (the EXFOR database).

  17. 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.

  18. Development of neutron detectors and neutron radiography at ...

    Indian Academy of Sciences (India)

    Apsara reactor has been used for a variety of applications in nuclear, aerospace, defense and metallurgical industries. The work done in the development of neutron detectors and neutron radiography is reported in this article. Keywords. Gas-filled neutron proportional counters; neutron radiography; hydride blis- ter.

  19. Neutron Angular Scatter Effects in 3DHZETRN: Quasi-Elastic

    Science.gov (United States)

    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.

  20. Conception d'un nouveau système de distribution de neutrons pour source à haut flux

    OpenAIRE

    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...

  1. Container Inspection Utilizing 14 MeV Neutrons

    Science.gov (United States)

    Valkovic, Vladivoj; Sudac, Davorin; Nad, Karlo; Obhodas, Jasmina

    2016-06-01

    A proposal for an autonomous and flexible ship container inspection system is presented. This could be accomplished by the incorporation of an inspection system on various container transportation devices (straddle carriers, yard gentry cranes, automated guided vehicles, trailers). The configuration is terminal specific and it should be defined by the container terminal operator. This enables that no part of the port operational area is used for inspection. The inspection scenario includes container transfer from ship to transportation device with the inspection unit mounted on it. The inspection is performed during actual container movement to the container location. A neutron generator without associated alpha particle detection is used. This allows the use of higher neutron intensities (5 × 109 - 1010 n/s in 4π). The inspected container is stationary in the “inspection position” on the transportation device while the “inspection unit” moves along its side. The following analytical methods will be used simultaneously: neutron radiography, X-ray radiography, neutron activation analysis, (n, γ) and (n,n'γ) reactions, neutron absorption. and scattering, X-ray backscattering. The neutron techniques will utilize “smart collimators” for neutrons and gamma rays, both emitted and detected. The inspected voxel is defined by the intersection of the neutron generator and the detectors solid angles. The container inspection protocol is based on identification of discrepancies between the cargo manifest, elemental “fingerprint” and radiography profiles. In addition, the information on container weight is obtained during the container transport and screening by measuring of density of material in the container.

  2. 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.

  3. 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.

  4. Convergent beam neutron crystallography

    Science.gov (United States)

    Gibson, Walter M.; Schultz, Arthur J.; Richardson, James W.; Carpenter, John M.; Mildner, David F. R.; Chen-Mayer, Heather H.; Miller, M. E.; Maxey, E.; Prask, Henry J.; Gnaeupel-Herold, Thomas H.; Youngman, Russell

    2004-01-01

    Applications of neutron diffraction for small samples (small fiducial areas are limited by the available neutron flux density. Recent demonstrations of convergent beam electron and x-ray diffraction and focusing of cold (λ>1 Å) neutrons suggest the possibility to use convergent beam neutron diffraction for small sample crystallography. We have carried out a systematic study of diffraction of both monoenergetic and broad bandwidth neutrons at the NIST Research Reactor and at the Intense Pulsed Neutron Source (IPNS) at Argonne National Laboratory. Combining convergent beams with time-of-flight Laue diffraction is particularly attractive for high efficiency small sample diffraction studies. We have studied single crystal and powder diffraction of neutrons with convergence angles as large as 15° and have observed diffracted peak intensity gains greater than 20. The convergent beam method (CBM) shows promise for crystallography on small samples of small to medium size molecules (potentially even for proteins), ultra-high pressure samples, and for mapping of strain and texture distributions in larger samples.

  5. Neutron detection technique

    CERN Document Server

    Oblath, N S

    2000-01-01

    The Sudbury Neutrino Observatory (SNO) has the ability to measure the total flux of all active flavors of neutrinos using the neutral current reaction, whose signature is a neutron. By comparing the rates of the neutral current reaction to the charged current reaction, which only detects electron neutrinos, one can test the neutrino oscillation hypothesis independent of solar models. It is necessary to understand the neutron detection efficiency of the detector to make use of the neutral current reaction. This report demonstrates a coincidence technique to identify neutrons emitted from the sup 2 sup 5 sup 2 Cf neutron calibration source. The source releases on average four neutrons when a sup 2 sup 5 sup 2 Cf nucleus spontaneously fissions. Each neutron is detected as a separate event when the neutron is captured by a deuteron, releasing a gamma ray of approximately 6.25 MeV. This gamma ray is in turn detected by the photomultiplier tube (PMT) array. By investigating the time and spatial separation between n...

  6. 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.

  7. Measurement of the Neutron Beta Decay Lifetime using Magnetically Trapped Ultracold Neutrons

    Science.gov (United States)

    Adamek, Evan Robert

    The neutron lifetime is an important parameter in the Standard Model of particle physics, with influences on the electroweak interaction and on Big Bang nucleosynthesis. Measurements of this quantity in cold beam experiments and in experiments using ultracold neutrons (UCN) disagree; this discrepancy may indicate that these measurements possess unaccounted-for systematic errors. The UCNtau experiment at Los Alamos Neutron Science Center (LANSCe) utilizes an asymmetrical magneto-gravitational storage volume with an in-situ vanadium detector. This setup is designed to either avoid or control many of the weaknesses that reduce systematic precision in other UCN lifetime experiments. Controlling for the many measurable errors requires detailed calculation and simulation, aided, for example, by the Geant4 Monte Carlo particle transport toolkit, which has been used to create a high fidelity model of the UCNtau experiment for modeling UCN transport, storage, and detection. Through the course of running the experiment, improvements in knowledge of particle measurement have led to improvements to the transport and to the detectors used in various parts of the experiment. With the experimental setup optimized to account for the subtleties of the measurement, the 2014-2015 beam period at LANSCe generated 85 measurement runs from which we could calculate the storage lifetime. Careful analysis of the effects of background on the vanadium detector assembly allowed for elimination of undesired signal and allowed for the extraction of a preliminary value for the neutron lifetime and the determination of areas to improve for the following run cycle.

  8. Neutrons in soft matter

    CERN Document Server

    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

  9. 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...

  10. Neutron scattering in liquids

    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.

  11. Hyperons in neutron stars

    Directory of Open Access Journals (Sweden)

    Tetsuya Katayama

    2015-07-01

    Full Text Available Using the Dirac–Brueckner–Hartree–Fock approach, the properties of neutron-star matter including hyperons are investigated. In the calculation, we consider both time and space components of the vector self-energies of baryons as well as the scalar ones. Furthermore, the effect of negative-energy states of baryons is partly taken into account. We obtain the maximum neutron-star mass of 2.08M⊙, which is consistent with the recently observed, massive neutron stars. We discuss a universal, repulsive three-body force for hyperons in matter.

  12. Superfluidity in neutron stars

    Energy Technology Data Exchange (ETDEWEB)

    Shaham, J.

    1980-01-01

    The possible role played by superfluid neutrons in the dynamics of neutron stars is discussed, with attention given to vortex structure and dynamics, the modes of the free vortex lattice, and the pinning of crustal vortices. Some effects associated with the interior superfluid state of neutron stars are discussed, including (1) the macroscopic post-glitch time scales, resulting from coupling between normal and superfluid components, (2) glitches due to unpinning events or to crust breaking by pinning vortices, (3) possible long-term modulation in rotation period, resulting from vortex coherent modes, and (4) gyroscopic effects of pinned vorticity.

  13. METHOD OF PRODUCING NEUTRONS

    Science.gov (United States)

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

    1964-01-14

    This patent relates to a method of producing neutrons in which there is produced a heated plasma containing heavy hydrogen isotope ions wherein heated ions are injected and confined in an elongated axially symmetric magnetic field having at least one magnetic field gradient region. In accordance with the method herein, the amplitude of the field and gradients are varied at an oscillatory periodic frequency to effect confinement by providing proper ratios of rotational to axial velocity components in the motion of said particles. The energetic neutrons may then be used as in a blanket zone containing a moderator and a source fissionable material to produce heat and thermal neutron fissionable materials. (AEC)

  14. NEUTRONIC REACTOR CONTROL

    Science.gov (United States)

    Metcalf, H.E.

    1958-10-14

    Methods of controlling reactors are presented. Specifically, a plurality of neutron absorber members are adjustably disposed in the reactor core at different distances from the center thereof. The absorber members extend into the core from opposite faces thereof and are operated by motive means coupled in a manner to simultaneously withdraw at least one of the absorber members while inserting one of the other absorber members. This feature effects fine control of the neutron reproduction ratio by varying the total volume of the reactor effective in developing the neutronic reaction.

  15. Absolute experimental and numerical calibration of the 14 MeV neutron source at the Frascati neutron generator

    Energy Technology Data Exchange (ETDEWEB)

    Angelone, M.; Pillon, M.; Batistoni, P.; Martini, M.; Martone, M.; Rado, V. [Associazione EURATOM-ENEA sulla Fusione, C.R. Frascati, C.P. 65---00044 Frascati, Rome (Italy)

    1996-06-01

    The absolute neutron yield of the 14 MeV Frascati neutron generator (FNG) is routinely measured by means of the associated alpha-particle method with a silicon surface barrier detector (SSD). This paper describes the work carried out to characterize the neutron source in terms of absolute intensity and angle-energy distribution of the emitted neutrons. The development of the measuring setup and the assessment of the measurement results are also reported. A complementary calibration procedure for validating the SSD results, based on the use of fission chambers and the activation technique, is also reported. An accurate analysis of the system has been performed via the Monte Carlo neutron and photon MCNP transport code. A detailed model of the neutron source that includes ion slowing down has been inserted into the MCNP code to permit a numerical calibration of the neutron source for comparison with the experimental results. The resulting agreement among the various methods is very good considering the uncertainties, and an accuracy of {plus_minus}2{percent} is achieved for the measurement of the 14 MeV neutron yield of the FNG. {copyright} {ital 1996 American Institute of Physics.}

  16. Absolute experimental and numerical calibration of the 14 MeV neutron source at the Frascati neutron generator

    Science.gov (United States)

    Angelone, M.; Pillon, M.; Batistoni, P.; Martini, M.; Martone, M.; Rado, V.

    1996-06-01

    The absolute neutron yield of the 14 MeV Frascati neutron generator (FNG) is routinely measured by means of the associated alpha-particle method with a silicon surface barrier detector (SSD). This paper describes the work carried out to characterize the neutron source in terms of absolute intensity and angle-energy distribution of the emitted neutrons. The development of the measuring setup and the assessment of the measurement results are also reported. A complementary calibration procedure for validating the SSD results, based on the use of fission chambers and the activation technique, is also reported. An accurate analysis of the system has been performed via the Monte Carlo neutron and photon MCNP transport code. A detailed model of the neutron source that includes ion slowing down has been inserted into the MCNP code to permit a numerical calibration of the neutron source for comparison with the experimental results. The resulting agreement among the various methods is very good considering the uncertainties, and an accuracy of ±2% is achieved for the measurement of the 14 MeV neutron yield of the FNG.

  17. On the definition of neutron lifetimes in multiplying and non-multiplying systems

    Energy Technology Data Exchange (ETDEWEB)

    Spriggs, G.D.; Adams, K.J.; Parsons, D.K. [and others

    1997-06-01

    Historically, the term neutron lifetime has been used in the literature to describe a wide variety of different time intervals associated with a neutron`s trek through a given system. This duplication of usage of the term neutron lifetime has undoubtedly resulted in some confusion concerning its physical meaning. In hopes of reducing some of this confusion, we suggest in this work that the various time intervals characterizing the life of a neutron be divided into three general categories: (1) neutron lifespans, (2) reaction rate lifetimes, and (3) neutron generation times. In this report, we define these three different time intervals and give deterministic and Monte Carlo transport expressions that can be used to calculate them.

  18. 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.

  19. Prompt fission neutron investigation in 235U(nth,f) reaction

    Science.gov (United States)

    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.

  20. Neutron signal transfer analysis

    CERN Document Server

    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...

  1. Neutrons from multifragmentation reactions

    CERN Document Server

    Trautmann, W; Brzychczyk, J; Buyukcizmeci, N; Mishustin, I N; Pawlowski, P

    2011-01-01

    The neutron emission in the fragmentation of stable and radioactive Sn and La projectiles of 600 MeV per nucleon has been studied with the Large Neutron Detector LAND coupled to the ALADIN forward spectrometer at SIS. A cluster-recognition algorithm is used to identify individual particles within the hit distributions registered with LAND. The obtained momentum distributions are extrapolated over the full phase space occupied by the neutrons from the projectile-spectator source. The mean multiplicities of spectator neutrons reach values of up to 12 and depend strongly on the isotopic composition of the projectile. An effective source temperature of T approx. 3 - 4 MeV is deduced from the transverse momentum distributions. For the interpretation of the data, calculations with the Statistical Multifragmentation Model for a properly chosen ensemble of excited sources were performed. The possible modification of the liquid-drop parameters of the fragment description in the hot environment is studied, and a signif...

  2. 3DHZETRN: Neutron leakage in finite objects

    Science.gov (United States)

    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.

  3. 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.

  4. Neutron scattering in dimers

    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...

  5. 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.

  6. Neutron Capture Gamma-Ray Libraries for Nuclear Applications

    Energy Technology Data Exchange (ETDEWEB)

    Sleaford, B W; Firestone, R B; Summers, N; Escher, J; Hurst, A; Krticka, M; Basunia, S; Molnar, G; Belgya, T; Revay, Z; Choi, H D

    2010-11-04

    The neutron capture reaction is useful in identifying and analyzing the gamma-ray spectrum from an unknown assembly as it gives unambiguous information on its composition. this can be done passively or actively where an external neutron source is used to probe an unknown assembly. There are known capture gamma-ray data gaps in the ENDF libraries used by transport codes for various nuclear applications. The Evaluated Gamma-ray Activation file (EGAF) is a new thermal neutron capture database of discrete line spectra and cross sections for over 260 isotopes that was developed as part of an IAEA Coordinated Research project. EGAF is being used to improve the capture gamma production in ENDF libraries. For medium to heavy nuclei the quasi continuum contribution to the gamma cascades is not experimentally resolved. The continuum contains up to 90% of all the decay energy and is modeled here with the statistical nuclear structure code DICEBOX. This code also provides a consistency check of the level scheme nuclear structure evaluation. The calculated continuum is of sufficient accuracy to include in the ENDF libraries. This analysis also determines new total thermal capture cross sections and provides an improved RIPL database. For higher energy neutron capture there is less experimental data available making benchmarking of the modeling codes more difficult. They are investigating the capture spectra from higher energy neutrons experimentally using surrogate reactions and modeling this with Hauser-Feshbach codes. This can then be used to benchmark CASINO, a version of DICEBOX modified for neutron capture at higher energy. This can be used to simulate spectra from neutron capture at incident neutron energies up to 20 MeV to improve the gamma-ray spectrum in neutron data libraries used for transport modeling of unknown assemblies.

  7. 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.

  8. Neutron star structure and the neutron radius of 208Pb.

    Science.gov (United States)

    Horowitz, C J; Piekarewicz, J

    2001-06-18

    We study relationships between the neutron-rich skin of a heavy nucleus and the properties of neutron-star crusts. Relativistic effective field theories with a thicker neutron skin in 208Pb have a larger electron fraction and a lower liquid-to-solid transition density for neutron-rich matter. These properties are determined by the density dependence of the symmetry energy which we vary by adding nonlinear couplings between isoscalar and isovector mesons. An accurate measurement of the neutron radius in 208Pb-via parity violating electron scattering-may have important implications for the structure of the crust of neutron stars.

  9. MAGNETIC NEUTRON SCATTERING

    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

  10. Neutron capture therapy for melanoma

    Energy Technology Data Exchange (ETDEWEB)

    Coderre, J.A.; Glass, J.D.; Micca, P.; Fairchild, R.G.

    1988-01-01

    The development of boron-containing compounds which localize selectively in tumor may require a tumor-by-tumor type of approach that exploits any metabolic pathways unique to the particular type of tumor. Melanin-producing melanomas actively transport and metabolize aromatic amino acids for use as precursors in the synthesis of the pigment melanin. It has been shown that the boron-containing amino acid analog p-borono-phenylalanine (BPA) is selectively accumulated in melanoma tissue, producing boron concentrations in tumor that are within the range estimated to be necessary for successful boron neutron capture therapy (BNCT). We report here the results of therapy experiments carried out at the Brookhaven Medical Research Reactor (BMRR). 21 refs., 5 figs., 3 tabs.

  11. 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.

  12. Optimization study for an epithermal neutron beam for boron neutron capture therapy at the University of Virginia Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Burns, Jr., Thomas Dean [Univ. of Virginia, Charlottesville, VA (United States)

    1995-05-01

    The non-surgical brain cancer treatment modality, Boron Neutron Capture Therapy (BNCT), requires the use of an epithermal neutron beam. This purpose of this thesis was to design an epithermal neutron beam at the University of Virginia Research Reactor (UVAR) suitable for BNCT applications. A suitable epithermal neutron beam for BNCT must have minimal fast neutron and gamma radiation contamination, and yet retain an appreciable intensity. The low power of the UVAR core makes reaching a balance between beam quality and intensity a very challenging design endeavor. The MCNP monte carlo neutron transport code was used to develop an equivalent core radiation source, and to perform the subsequent neutron transport calculations necessary for beam model analysis and development. The code accuracy was validated by benchmarking output against experimental criticality measurements. An epithermal beam was designed for the UVAR, with performance characteristics comparable to beams at facilities with cores of higher power. The epithermal neutron intensity of this beam is 2.2 x 108 n/cm2 • s. The fast neutron and gamma radiation KERMA factors are 10 x 10-11cGy•cm2/nepi and 20 x 10-11 cGy•cm2/nepi , respectively, and the current-to-flux ratio is 0.85. This thesis has shown that the UVAR has the capability to provide BNCT treatments, however the performance characteristics of the final beam of this study were limited by the low core power.

  13. Effects of Hot-Spot Geometry on Backscattering and Down-Scattering Neutron Spectra

    Science.gov (United States)

    Mohamed, Z. L.; Mannion, O. M.; Forrest, C. J.; Knauer, J. P.; Anderson, K. S.; Radha, P. B.

    2017-10-01

    The measured neutron spectrum produced by a fusion experiment plays a key role in inferring observable quantities. One important observable is the areal density of an implosion, which is inferred by measuring the scattering of neutrons. This project seeks to use particle-transport simulations to model the effects of hot-spot geometry on backscattering and down-scattering neutron spectra along different lines of sight. Implosions similar to those conducted at the Laboratory of Laser Energetics are modeled by neutron transport through a DT plasma and a DT ice shell using the particle transport codes MCNP and IRIS. Effects of hot-spot geometry are obtained by ``detecting'' scattered neutrons along different lines of sight. This process is repeated for various hot-spot geometries representing known shape distortions between the hot spot and the shell. This material is based upon work supported by the Department of Energy National Nuclear Security Administration under Award Number DE-NA0001944.

  14. 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.

  15. In situ calibration of neutron activation system on the large helical device

    Science.gov (United States)

    Pu, N.; Nishitani, T.; Isobe, M.; Ogawa, K.; Kawase, H.; Tanaka, T.; Li, S. Y.; Yoshihashi, S.; Uritani, A.

    2017-11-01

    In situ calibration of the neutron activation system on the Large Helical Device (LHD) was performed by using an intense 252Cf neutron source. To simulate a ring-shaped neutron source, we installed a railway inside the LHD vacuum vessel and made a train loaded with the 252Cf source run along a typical magnetic axis position. Three activation capsules loaded with thirty pieces of indium foils stacked with total mass of approximately 18 g were prepared. Each capsule was irradiated over 15 h while the train was circulating. The activation response coefficient (9.4 ± 1.2) × 10-8 of 115In(n, n')115mIn reaction obtained from the experiment is in good agreement with results from three-dimensional neutron transport calculations using the Monte Carlo neutron transport simulation code 6. The activation response coefficients of 2.45 MeV birth neutron and secondary 14.1 MeV neutron from deuterium plasma were evaluated from the activation response coefficient obtained in this calibration experiment with results from three-dimensional neutron calculations using the Monte Carlo neutron transport simulation code 6.

  16. 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

  17. 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 cm2 treatment beam. The calculated boron dose enhancement at 5.7-cm depth in a water-filled head phantom in the assembly with a 5x5 cm2 collimation was 21.9% per 100-ppm 10B 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 cm2 fast neutron treatment beam. To validate the design calculations, a simplified BNCEFNT assembly was built using four lead bricks to form a 5x5 cm2 collimator. Five 1.0-cm thick 20x20 cm2 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 10B) to measure dose due to boron neutron capture. The

  18. 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.

  19. Deficiency in Monte Carlo simulations of coupled neutron-gamma-ray fields

    NARCIS (Netherlands)

    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

  20. The intensive DT neutron generator of TU Dresden

    Directory of Open Access Journals (Sweden)

    Klix Axel

    2018-01-01

    Full Text Available TU Dresden operates an accelerator-based intensive DT neutron generator. Experimental activities comprise investigation into material activation and decay, neutron and photon transport in matter and R&D work on radiation detectors for harsh environments. The intense DT neutron generator is capable to produce a maximum of 1012 n/s. The neutron source is a solid-type water-cooled tritium target based on a titanium matrix on a copper carrier. The neutron yield at a typical deuteron beam current of 1 mA is of the order of 1011 n/s in 4Π. A pneumatic sample transport system is available for short-time irradiations and connected to wo high-purity germanium detector spectrometers for the measurement of induced activities. The overall design of the experimental hall with the neutron generator allows a flexible setup of experiments including the possibility of investigating larger structures and cooled samples or samples at high temperatures.

  1. A multitask neutron beam line for spallation neutron sources

    Science.gov (United States)

    Pietropaolo, A.; Festa, G.; Grazzi, F.; Barzagli, E.; Scherillo, A.; Schooneveld, E. M.; Civita, F.

    2011-08-01

    Here we present a new concept for a time-of-flight neutron scattering instrument allowing for simultaneous application of three different techniques: time-of-flight neutron diffraction, neutron resonance capture analysis and Bragg edge transmission analysis. The instrument can provide average resolution neutron radiography too. The potential of the proposed concept was explored by implementing the necessary equipment on INES (Italian Neutron Experimental Station) at the ISIS spallation neutron source (UK). The results obtained show the effectiveness of the proposed instrument to acquire relevant quantitative information in a non-invasive way on a historical metallurgical sample, namely a Japanese hand guard (tsuba). The aforementioned neutron techniques simultaneously exploited the extended neutron energy range available from 10 meV to 1 keV. This allowed a fully satisfactory characterization of the sample in terms of metal components and their combination in different phases, and forging and assembling methods.

  2. A collimated neutron detector for RFP plasmas in MST

    Energy Technology Data Exchange (ETDEWEB)

    Capecchi, W. J., E-mail: capecchi@wisc.edu; Anderson, J. K.; Bonofiglo, P. J.; Kim, J.; Sears, S. [University of Wisconsin- Madison, Madison, Wisconsin 53706 (United States)

    2016-11-15

    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 10{sup 4} 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 (∼10{sup 4}/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.

  3. A collimated neutron detector for RFP plasmas in MST

    Science.gov (United States)

    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.

  4. Neutron spectra and dosimetric assessment around a neutron Howitzer container

    OpenAIRE

    Barros, Silvia; Gallego Díaz, Eduardo F.; Lorente Fillol, Alfredo; Gonçalves, Isabel F.; Vaz, Pedro; Vega-Carrillo, Héctor René; Zankl, María

    2014-01-01

    The neutron Howitzer container at the Neutron Measurements Laboratory of the Nuclear Engineering Department of the Polytechnic University of Madrid (UPM), is equipped with a 241Am-Be neutron source of 74 GBq in its center. The container allows the source to be in either the irradiation or the storage position. To measure the neutron fluence rate spectra around the Howitzer container, measurements were performed using a Bonner spheres spectrometer and the spectra were unfolded using the NSDann...

  5. Neutron nuclear physics under the neutron science project

    Energy Technology Data Exchange (ETDEWEB)

    Chiba, Satoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-11-01

    The concept of fast neutron physics facility in the Neutron Science Research project is described. This facility makes use of an ultra-short proton pulse (width < 1 ns) for fast neutron time-of-flight works. The current design is based on an assumption of the maximum proton current of 100 {mu}A. Available neutron fluence and energy resolution are explained. Some of the research subjects to be performed at this facility are discussed. (author)

  6. Study of a position-sensitive scintillator neutron detector

    Science.gov (United States)

    Tang, Bin; Sun, Zhi-Jia; Zhang, Qiang; Yang, Zhen; Xu, Hong; Yang, Gui-An; Wang, Yan-Feng; Wu, Chong; Chen, Yuan-Bo; Yang, Lei

    2012-11-01

    The investigation of a novel thermal neutron detector is developed to fulfill the requirements of the high intensity power diffractometer (HIPD) at the Chinese Spallation Neutron Source (CSNS). It consists of two layers of 6LiF/ZnS(Ag) scintillators, two layers of crossed WLSF arrays, several multi-anode photo multiplier tubes (MA-PMT), and the matching readout electronics. The neutron detection efficiency of the scintilltors, the light transportation ability of the WLSF, and the spatial linearity of the readout electronics are measured and discussed in this paper. It shows that the sandwich structure and the compact readout electronics could fulfill the needs of the HIPD. A prototype with a 10 cm×10 cm sensitive area has been constructed to further study the characteristics of the neutron scintillator detector.

  7. Probing proton transition momentum in neutron-rich matter

    Science.gov (United States)

    Yong, Gao-Chan

    2018-01-01

    Around the nuclear Fermi momentum, there is a transition of nucleon momentum distribution n(k) in nuclear matter, i.e., from a constant to the 1 /k4 nucleon momentum distribution. While nowadays the transition momentum of minority in asymmetric matter is rarely studied and thus undetermined. In the framework of the IBUU transport model, proton transition momentum in nuclei is first studied. It is found that the transition momentum of proton is sensitive to the π- /π+ ratio as well as the energetic photon production in neutron-rich nuclear reaction. This result may push the study of how the proton momentum is distributed in neutron-rich matter forward and help us to better understand the dynamics of both neutron-rich nuclear reactions and neutron stars.

  8. Neutron halos in hypernuclei

    CERN Document Server

    Lue, H F; Meng, J; Zhou, S G

    2003-01-01

    Properties of single-LAMBDA and double-LAMBDA hypernuclei for even-N Ca isotopes ranging from the proton dripline to the neutron dripline are studied using the relativistic continuum Hartree-Bogolyubov theory with a zero-range pairing interaction. Compared with ordinary nuclei, the addition of one or two LAMBDA-hyperons lowers the Fermi level. The predicted neutron dripline nuclei are, respectively, sup 7 sup 5 subLAMBDA Ca and sup 7 sup 6 sub 2 subLAMBDA Ca, as the additional attractive force provided by the LAMBDA-N interaction shifts nuclei from outside to inside the dripline. Therefore, the last bound hypernuclei have two more neutrons than the corresponding ordinary nuclei. Based on the analysis of two-neutron separation energies, neutron single-particle energy levels, the contribution of continuum and nucleon density distribution, giant halo phenomena due to the pairing correlation, and the contribution from the continuum are suggested to exist in Ca hypernuclei similar to those that appear in ordinary ...

  9. Directions in Radiation Transport Modelling

    Directory of Open Access Journals (Sweden)

    P Nicholas Smith

    2016-12-01

    More exciting advances are on the horizon to increase the power of simulation tools. The advent of high performance computers is allowing bigger, higher fidelity models to be created, if the challenges of parallelization and memory management can be met. 3D whole core transport modelling is becoming possible. Uncertainty quantification is improving with large benefits to be gained from more accurate, less pessimistic estimates of uncertainty. Advanced graphical displays allow the user to assimilate and make sense of the vast amounts of data produced by modern modelling tools. Numerical solvers are being developed that use goal-based adaptivity to adjust the nodalisation of the system to provide the optimum scheme to achieve the user requested accuracy on the results, thus removing the need to perform costly convergence studies in space and angle etc. More use is being made of multi-physics methods in which radiation transport is coupled with other phenomena, such as thermal-hydraulics, structural response, fuel performance and/or chemistry in order to better understand their interplay in reactor cores.

  10. Probing neutron star physics using accreting neutron stars

    NARCIS (Netherlands)

    Patruno, A.

    2010-01-01

    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

  11. Behavior of neutrons under different thicknesses of moderation; Comportamiento de los neutrones bajo diferentes espesores de moderacion

    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)

  12. Measurement of neutron spectra in a silicon filtered neutron beam using stilbene detectors at the LVR-15 research reactor.

    Science.gov (United States)

    Košťál, Michal; Šoltés, Jaroslav; Viererbl, Ladislav; Matěj, Zdeněk; Cvachovec, František; Rypar, Vojtěch; Losa, Evžen

    2017-10-01

    A well-defined neutron spectrum is an essential tool for calibration and tests of spectrometry and dosimetry detectors, and evaluation methods for spectra processing. Many of the nowadays used neutron standards are calibrated against a fission spectrum which has a rather smooth energy dependence. In recent time, at the LVR-15 research reactor in Rez, an alternative approach was tested for the needs of fast neutron spectrometry detector calibration. This process comprises detector tests in a neutron beam, filtered by one meter of single-crystalline silicon, which contains several significant peaks in the fast neutron energy range. Tests in such neutron field can possibly reveal specific problems in the deconvolution matrix of the detection system, which may stay hidden in fields with a smooth structure and can provide a tool for a proper energy calibration. Test with several stilbene scintillator crystals in two different beam configurations supplemented by Monte-Carlo transport calculations have been carried out. The results have shown a high level of agreement between the experimental data and simulation, proving thus the accuracy of used deconvolution matrix. The chosen approach can, thus, provide a well-defined neutron reference field with a peaked structure for further tests of spectra evaluation methods and scintillation detector energy calibration. Copyright © 2017 Elsevier Ltd. All rights reserved.

  13. Virtual neutron scattering experiments

    DEFF Research Database (Denmark)

    Overgaard, Julie Hougaard; Bruun, Jesper; May, Michael

    2017-01-01

    . 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......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...... 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...

  14. 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...

  15. Hyperons and neutron stars

    Energy Technology Data Exchange (ETDEWEB)

    Vidaña, Isaac [Centro de Física Computacional, Department of Physics, University of Coimbra, PT-3004-516 Coimbra (Portugal)

    2015-02-24

    In this lecture I will briefly review some of the effects of hyperons on the properties of neutron and proto-neutron stars. In particular, I will revise the problem of the strong softening of the EoS, and the consequent reduction of the maximum mass, induced by the presence of hyperons, a puzzle which has become more intringuing and difficult to solve due the recent measurements of the unusually high masses of the millisecond pulsars PSR J1903+0327 (1.667±0.021M{sub ⊙}), PSR J1614–2230 (1.97±0.04M{sub ⊙}), and PSR J0348+0432 (2.01±0.04M{sub ⊙}). Finally, I will also examine the role of hyperons on the cooling properties of newly born neutron stars and on the so-called r-mode instability.

  16. NEUTRONIC REACTOR CORE INSTRUMENT

    Science.gov (United States)

    Mims, L.S.

    1961-08-22

    A multi-purpose instrument for measuring neutron flux, coolant flow rate, and coolant temperature in a nuclear reactor is described. The device consists essentially of a hollow thimble containing a heat conducting element protruding from the inner wall, the element containing on its innermost end an amount of fissionsble materinl to function as a heat source when subjected to neutron flux irradiation. Thermocouple type temperature sensing means are placed on the heat conducting element adjacent the fissionable material and at a point spaced therefrom, and at a point on the thimble which is in contact with the coolant fluid. The temperature differentials measured between the thermocouples are determinative of the neutron flux, coolant flow, and temperature being measured. The device may be utilized as a probe or may be incorporated in a reactor core. (AE C)

  17. Corrosion resistant neutron absorbing coatings

    Science.gov (United States)

    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.

  18. A portable shield for a neutron howitzer used for instructional and research purposes.

    Science.gov (United States)

    Ali, S; Laherty, M; Laprade, J W; Cala, P M; Lipschultz, F P; Neuhauser, B

    2015-09-01

    Neutron howitzers are routinely used in universities to activate samples for instructional laboratory experiments on radioactivity. They are also a convenient source of neutrons and gammas for research purposes, but they must be used with caution. This paper describes the modeling, design, construction, and testing of a portable, economical shield for a 1.0 Curie neutron howitzer. The Monte Carlo N Particle Transport Code (MCNP5) has been used to model the (239)PuBe source and the howitzer and to design the external neutron and gamma shield. Copyright © 2015 Elsevier Ltd. All rights reserved.

  19. Fast-neutron-induced fission of 242Pu at nELBE

    Directory of Open Access Journals (Sweden)

    Kögler Toni

    2017-01-01

    Full Text Available The fast neutron-induced fission cross section of 242Pu was determined in the range of 0.5 MeV to 10 MeV relative to 235U(n,f at the neutron time-of-flight facility nELBE. The number of target nuclei was calculated by means of measuring the spontaneous fission rate of 242Pu. Neutron transport simulations with Geant4 and MCNP6 are used to correct the relative cross section for neutron scattering. The determined results are in good agreement with current experimental and evaluated data sets.

  20. Shielding of a neutron irradiator with {sup 241}Am-Be source

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, K.A.M. de; Crispim, V.R.; Silva, A.X., E-mail: koliveira@con.ufrj.b, E-mail: verginia@con.ufrj.b, E-mail: ademir@con.ufrj.b [Universidade Federal do Rio de Janeiro (PEN/COPPE/UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-Graduacao de Engenharia. Programa de Engenharia Nuclear; Fonseca, E.S., E-mail: evaldo@ird.gov.b [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The equivalent dose rates at 1.0 cm from the outer surface of the shielding of a neutron irradiation system that uses {sup 241}Am-Be source with activity of 185 GBq (5 Ci) were determined. A theoretical-experimental approach including case studies, through computer simulations with MCNP code was employed to calculate the best shielding thickness. Following the construction of the neutron irradiator, dose measurements were conducted in order to validate data obtained from simulation. The neutron irradiator shielding was designed in such a way to allow transport of the neutron radiography system for in loco inspections ensuring workers' radiologic safety. (author)

  1. Pixelated neutron image plates

    Science.gov (United States)

    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.

  2. Interfering with the neutron spin

    Indian Academy of Sciences (India)

    Permanent link: http://www.ias.ac.in/article/fulltext/pram/063/01/0051-0056. Keywords. Neutron interferometry; spinor; Pauli anticommutation; geometic phase; non-cyclic evolution; Pancharatnam connection; neutron polarimetry. Abstract. Charge neutrality, a spin 1 2 and an associated magnetic moment of the neutron make ...

  3. GUIDE FOR POLARIZED NEUTRONS

    Science.gov (United States)

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

    1962-12-01

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

  4. Neutron recognition in the LAND detector for large neutron multiplicity

    Science.gov (United States)

    Pawłowski, P.; Brzychczyk, J.; 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.; Le Fèvre, A.; Le Gentil, E.; 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.

    2012-12-01

    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.

  5. Scission neutron emission and prompt fission neutron spectrum

    CERN Document Server

    Kornilov, N V

    2001-01-01

    The mass, energy and angular integrated spectra of prompt fission neutrons for sup 2 sup 3 sup 5 U induced fission in the energy range from thermal to 5 MeV were analyzed. It allows assume that about 0.362+-0.025 neutrons per fission are emitted due to another mechanism then neutron emission from excited fragments after full acceleration. The spectrum of scission neutrons consists of two components with average energy 0.98 MeV and 2.74 MeV. The share of scission neutrons and their spectrum shape estimated in this work does not contradict to results of differential experiments analyzed in previous papers.

  6. Particulate matter and neutron activation analysis

    Energy Technology Data Exchange (ETDEWEB)

    Otoshi, Tsunehiko [Tohoku Univ. of Community Service and Science, Sakata, Yamagata (Japan)

    2003-03-01

    In these years, economy of East Asian region is rapidly growing, and countries in this region are facing serious environmental problems. Neutron activation analysis is known as one of high-sensitive analytical method for multi elements. And it is a useful tool for environmental research, particularly for the study on atmospheric particulate matter that consists of various constituents. Elemental concentration represents status of air, such as emission of heavy metals from industries and municipal incinerators, transportation of soil derived elements more than thousands of kilometers, and so on. These monitoring data obtained by neutron activation analysis can be a cue to evaluate environment problems. Japanese government launched National Air Surveillance Network (NASN) employing neutron activation analysis in 1974, and the data has been accumulated at about twenty sampling sites. As a result of mitigation measure of air pollution sources, concentrations of elements that have anthropogenic sources decreased particularly at the beginning of the monitoring period. However, even now, concentrations of these anthropogenic elements reflect the characteristics of each sampling site, e.g. industrial/urban, rural, and remote. Soil derived elements have a seasonal variation because of the contribution of continental dust transported by strong westerly winds prevailing in winter and spring season. The health effects associated with trace elements in particulate matter have not been well characterized. However, there is increasing evidence that particulate air pollution, especially fine portion of particles in many different cities is associated with acute mortality. Neutron activation analysis is also expected to provide useful information to this new study field related to human exposures and health risk. (author)

  7. A Finite Element Projection Method for the Solution of Particle Transport Problems with Anisotropic Scattering.

    Science.gov (United States)

    1984-07-01

    Ackroyd , R. T. "A Finite Element Technique for the Even Parity Neutron Transport Equation," The Mathematics of Finite Elements and...state transport equation. In 1972 a more detailed examination of the use of finite elements to solve neutron diffusion problems was provided by Kaper...79, 269-277 (1981). Ukai, S., "Solution of the Multi-dimensional Neutron Transport Equation by Finite Element Methods,"

  8. Neutron Deep Penetration Calculations in Light Water with Monte Carlo TRIPOLI-4® Variance Reduction Techniques

    Directory of Open Access Journals (Sweden)

    Lee Yi-Kang

    2017-01-01

    Full Text Available Nuclear decommissioning takes place in several stages due to the radioactivity in the reactor structure materials. A good estimation of the neutron activation products distributed in the reactor structure materials impacts obviously on the decommissioning planning and the low-level radioactive waste management. Continuous energy Monte-Carlo radiation transport code TRIPOLI-4 has been applied on radiation protection and shielding analyses. To enhance the TRIPOLI-4 application in nuclear decommissioning activities, both experimental and computational benchmarks are being performed. To calculate the neutron activation of the shielding and structure materials of nuclear facilities, the knowledge of 3D neutron flux map and energy spectra must be first investigated. To perform this type of neutron deep penetration calculations with the Monte Carlo transport code, variance reduction techniques are necessary in order to reduce the uncertainty of the neutron activation estimation. In this study, variance reduction options of the TRIPOLI-4 code were used on the NAIADE 1 light water shielding benchmark. This benchmark document is available from the OECD/NEA SINBAD shielding benchmark database. From this benchmark database, a simplified NAIADE 1 water shielding model was first proposed in this work in order to make the code validation easier. Determination of the fission neutron transport was performed in light water for penetration up to 50 cm for fast neutrons and up to about 180 cm for thermal neutrons. Measurement and calculation results were benchmarked. Variance reduction options and their performance were discussed and compared.

  9. Neutron Deep Penetration Calculations in Light Water with Monte Carlo TRIPOLI-4® Variance Reduction Techniques

    Science.gov (United States)

    Lee, Yi-Kang

    2017-09-01

    Nuclear decommissioning takes place in several stages due to the radioactivity in the reactor structure materials. A good estimation of the neutron activation products distributed in the reactor structure materials impacts obviously on the decommissioning planning and the low-level radioactive waste management. Continuous energy Monte-Carlo radiation transport code TRIPOLI-4 has been applied on radiation protection and shielding analyses. To enhance the TRIPOLI-4 application in nuclear decommissioning activities, both experimental and computational benchmarks are being performed. To calculate the neutron activation of the shielding and structure materials of nuclear facilities, the knowledge of 3D neutron flux map and energy spectra must be first investigated. To perform this type of neutron deep penetration calculations with the Monte Carlo transport code, variance reduction techniques are necessary in order to reduce the uncertainty of the neutron activation estimation. In this study, variance reduction options of the TRIPOLI-4 code were used on the NAIADE 1 light water shielding benchmark. This benchmark document is available from the OECD/NEA SINBAD shielding benchmark database. From this benchmark database, a simplified NAIADE 1 water shielding model was first proposed in this work in order to make the code validation easier. Determination of the fission neutron transport was performed in light water for penetration up to 50 cm for fast neutrons and up to about 180 cm for thermal neutrons. Measurement and calculation results were benchmarked. Variance reduction options and their performance were discussed and compared.

  10. 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.

  11. Optimization of neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Hooper, E.B.

    1993-11-09

    I consider here the optimization of the two component neutron source, allowing beam species and energy to vary. A simple model is developed, based on the earlier publications, that permits the optimum to be obtained simply. The two component plasma, with one species of hot ion (D{sup +} or T{sup +}) and the complementary species of cold ion, is easy to analyze in the case of a spatially uniform cold plasma, as to good approximation the total number of hot ions is important but not their spatial distribution. Consequently, the optimization can ignore spatial effects. The problem of a plasma with both types of hot ions and cold ions is rather more difficult, as the neutron production by hot-hot interactions is sensitive to their spatial distributions. Consequently, consideration of this problem will be delayed to a future memorandum. The basic model is that used in the published articles on the two-component, beam-plasma mirror source. I integrate the Fokker-Planck equation analytically, obtaining good agreement with previous numerical results. This simplifies the optimization, by providing a functional form for the neutron production. The primary result is expressed in terms of the power efficiency: watts of neutrons/watts of primary power. The latter includes the positive ion neutralization efficiency. At 150 keV, the present model obtains an efficiency of 0.66%, compared with 0.53% of the earlier calculation.

  12. 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.)

  13. Neutron scattering from fractals

    DEFF Research Database (Denmark)

    Kjems, Jørgen; Freltoft, T.; Richter, D.

    1986-01-01

    -angle neutron scattering studies of the variation with aggregation rate are presented. These results allow a very detailed comparison to be made with the theoretical scattering curves. Preliminary incoherent inelastic data on the low-frequency dynamics of hydroxylated silica particle aggregates show a clear...

  14. Small Angle Neutron Scattering

    Energy Technology Data Exchange (ETDEWEB)

    Urban, Volker S [ORNL

    2012-01-01

    Small Angle Neutron Scattering (SANS) probes structural details at the nanometer scale in a non-destructive way. This article gives an introduction to scientists who have no prior small-angle scattering knowledge, but who seek a technique that allows elucidating structural information in challenging situations that thwart approaches by other methods. SANS is applicable to a wide variety of materials including metals and alloys, ceramics, concrete, glasses, polymers, composites and biological materials. Isotope and magnetic interactions provide unique methods for labeling and contrast variation to highlight specific structural features of interest. In situ studies of a material s responses to temperature, pressure, shear, magnetic and electric fields, etc., are feasible as a result of the high penetrating power of neutrons. SANS provides statistical information on significant structural features averaged over the probed sample volume, and one can use SANS to quantify with high precision the structural details that are observed, for example, in electron microscopy. Neutron scattering is non-destructive; there is no need to cut specimens into thin sections, and neutrons penetrate deeply, providing information on the bulk material, free from surface effects. The basic principles of a SANS experiment are fairly simple, but the measurement, analysis and interpretation of small angle scattering data involves theoretical concepts that are unique to the technique and that are not widely known. This article includes a concise description of the basics, as well as practical know-how that is essential for a successful SANS experiment.

  15. Cooling of Neutron Stars

    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.

  16. Neutronic Reactor Structure

    Science.gov (United States)

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

    1961-05-30

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

  17. NEUTRONIC REACTOR CONTROL ELEMENT

    Science.gov (United States)

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

    1962-04-17

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

  18. New Neutron Dosimeter

    CERN Multimedia

    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...

  19. Neutronic analysis of a high power density hybrid reactor using ...

    Indian Academy of Sciences (India)

    Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and an S8–P3 approximation. The investigated blanket using Flinabe or Li20Sn80 shows better fissile fuel breeding and fuel enrichment characteristics compared to that with natural lithium which shows that these two innovative coolants can.

  20. Coherence effects and neutrino pair bremsstrahlung in neutron stars

    NARCIS (Netherlands)

    Sedrakian, A; Dieperink, A

    1999-01-01

    We calculate the rate of energy radiation by bremsstrahlung of neutrino pairs by baryons in neutron stars employing a transport model where neutrinos couple to baryons with spectral width. The: coherence effects, which are included by computing the self energies with fully dressed propagators, lead

  1. A detector for neutron imaging

    CERN Document Server

    Britton, C L; Wintenberg, A L; Warmack, R J; McKnight, T E; Frank, S S; Cooper, R G; Dudney, N J; Veith, G M; Stephan, A C

    2004-01-01

    A bright neutron source such as the Spallation Neutron Source (SNS) places extreme requirements on detectors including excellent 2-D spatial imaging and high dynamic range. Present imaging detectors have either shown position resolutions that are less than acceptable or they exhibit excessive paralyzing dead times due to the brightness of the source. High neutron detection efficiency with good neutron- gamma discrimination is critical for applications in neutron scattering research where the usefulness of the data is highly dependent on the statistical uncertainty associated with each detector pixel.. A detector concept known as MicroMegas (MicroMEsh GAseous Structure) has been developed at CERN in Geneva for high- energy physics charged-particle tracking applications and has shown great promise for handling high data rates with a rather low-cost structure. We are attempting to optimize the MicroMegas detector concept for thermal neutrons and have designed a 1-D neutron strip detector which we have tested In ...

  2. 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.)

  3. Neutronic investigation and activation calculation for CFETR HCCB blankets

    Science.gov (United States)

    Shuling, XU; Mingzhun, LEI; Sumei, LIU; Kun, LU; Kun, XU; Kun, PEI

    2017-12-01

    The neutronic calculations and activation behavior of the proposed helium cooled ceramic breeder (HCCB) blanket were predicted for the Chinese Fusion Engineering Testing Reactor (CFETR) design model using the MCNP multi-particle transport code and its associated data library. The tritium self-sufficiency behavior of the HCCB blanket was assessed, addressing several important breeding-related arrangements inside the blankets. Two candidate first wall armor materials were considered to obtain a proper tritium breeding ratio (TBR). Presentations of other neutronic characteristics, including neutron flux, neutron-induced damages in terms of the accumulated dpa and helium production were also conducted. Activation, decay heat levels and contact dose rates of the components were calculated to estimate the neutron-induced radioactivity and personnel safety. The results indicate that neutron radiation is efficiently attenuated and slowed down by components placed between the plasma and toroidal field coil. The dominant nuclides and corresponding isotopes in the structural steel were discussed. A radioactivity comparison between pure beryllium and beryllium with specific impurities was also performed. After a millennium cooling time, the decay heat of all the concerned components and materials is less than 1 × 10‑4 kW, and most associated in-vessel components qualify for recycling by remote handling. The results demonstrate that acceptable hands-on recycling and operation still require a further long waiting period to allow the activated products to decay.

  4. The muon-induced neutron indirect detection EXperiment, MINIDEX

    Science.gov (United States)

    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.

  5. The neutrons for science facility at SPIRAL-2

    Science.gov (United States)

    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.

  6. Neutron beam design for low intensity neutron and gamma-ray radioscopy using small neutron sources

    CERN Document Server

    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.

  7. Exploring performance of neutron guide systems using pinhole beam extraction

    Energy Technology Data Exchange (ETDEWEB)

    Bertelsen, Mads, E-mail: mads.bertelsen@gmail.com [Nanoscience Center and eScience Center, Niels Bohr Institute, University of Copenhagen (Denmark); ESS Design Update Program (Denmark); Jacobsen, Henrik; Bengaard Hansen, Ursula [Nanoscience Center and eScience Center, Niels Bohr Institute, University of Copenhagen (Denmark); ESS Design Update Program (Denmark); Hoffmann Carlsen, Henrik [Nanoscience Center and eScience Center, Niels Bohr Institute, University of Copenhagen (Denmark); Lefmann, Kim [Nanoscience Center and eScience Center, Niels Bohr Institute, University of Copenhagen (Denmark); ESS Design Update Program (Denmark)

    2013-11-21

    We perform an optimization of pinhole neutron guide systems under systematically varying conditions. It is investigated how neutron guide systems consisting of a parabolic feeder inside the biological shielding followed by a pinhole and an elliptical guide perform with different pinhole sizes and divergence requirements. We have clarified in which situations such a guide system is a viable choice and when the parabolic feeder is necessary in terms of neutron transport. The advantage of this design is the reduction of background from fast thermal neutrons compared to a system without a pinhole, hence the smallest possible pinhole is of interest. It is found that instruments with divergence requirements of ±1.0° will have excellent neutron transport with a 3×3 cm{sup 2} pinhole, while lower divergence requirements of ±0.5° can do with a smaller pinhole of 2×2 cm{sup 2}. The feeder effectively reduces the necessary pinhole size, and is especially beneficial for short instruments. In addition to these qualities, a feeder will often smoothen the divergence profile, mostly for longer instruments.

  8. Neutron imaging integrated circuit and method for detecting neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Nagarkar, Vivek V.; More, Mitali J.

    2017-12-05

    The present disclosure provides a neutron imaging detector and a method for detecting neutrons. In one example, a method includes providing a neutron imaging detector including plurality of memory cells and a conversion layer on the memory cells, setting one or more of the memory cells to a first charge state, positioning the neutron imaging detector in a neutron environment for a predetermined time period, and reading a state change at one of the memory cells, and measuring a charge state change at one of the plurality of memory cells from the first charge state to a second charge state less than the first charge state, where the charge state change indicates detection of neutrons at said one of the memory cells.

  9. Fast neutron imaging device and method

    Science.gov (United States)

    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.

  10. Neutron Activation Analysis: Techniques and Applications

    Science.gov (United States)

    MacLellan, Ryan

    2011-04-01

    The role of neutron activation analysis in low-energy low-background experimentsis discussed in terms of comparible methods. Radiochemical neutron activation analysis is introduce. The procedure of instrumental neutron activation analysis is detailed especially with respect to the measurement of trace amounts of natural radioactivity. The determination of reactor neutron spectrum parameters required for neutron activation analysis is also presented.

  11. Neutron electric polarizability

    CERN Document Server

    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.

  12. 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.

  13. Are there good probes for the di-neutron correlation in light neutron-rich nuclei?

    OpenAIRE

    Hagino, K.; Sagawa, H.

    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.

  14. High-resolution neutron microtomography with noiseless neutron counting detector

    Science.gov (United States)

    Tremsin, A. S.; McPhate, J. B.; Vallerga, J. V.; Siegmund, O. H. W.; Feller, W. B.; Lehmann, E.; Butler, L. G.; Dawson, M.

    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 (˜70% for cold neutrons), spatial resolutions ranging from 15 to 55 μm and a temporal resolution of ˜1 μs—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 ˜400 μm grains in an organic powder encapsulated in a ˜700 μm thick metal casing.

  15. Neutron polarizability. Possibilities of its determination in neutron experiments

    CERN Document Server

    Aleksandrov, Y A

    2001-01-01

    The history of question of neutron polarizability is discussed. Most of the neutron physical experiments conducted at neutron energies below 14 MeV to discover the electric polarizability of the neutron are reviewed.The existence of additional scattering after all known long range-related phenomena are taken into account are emphasized. In the keV neutron energy region, the effect of neutron polarizability on the angular distribution of scattering (over a wide range of angles) and the energy behavior of the total cross section of neutron interaction is studied. Finally, in the region of low energies (below 1 keV) the focus is on the influence of polarizability on the energy dependence of total neutron cross sections. It is emphasized that measurements at energies below several hundreds keV have not given any positive results yet due to the smallness of the experimental effect. Possible existence of an additional potential of neutron scattering on nuclei with a longer range than that of the usual nuclear poten...

  16. Neutron Activation Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Corliss, William R.

    1968-01-01

    In activation analysis, a sample of an unknown material is first irradiated (activated) with nuclear particles. In practice these nuclear particles are almost always neutrons. The success of activation analysis depends upon nuclear reactions which are completely independent of an atom's chemical associations. The value of activation analysis as a research tool was recognized almost immediately upon the discovery of artificial radioactivity. This book discusses activation analysis experiments, applications and technical considerations.

  17. Neutron structural biology

    Energy Technology Data Exchange (ETDEWEB)

    Schoenborn, B.

    1997-10-01

    This is the final report of a one-year, Laboratory Directed Research and Development (LDRD) project at Los Alamos National Laboratory (LANL). We investigated design concepts of neutron scattering capabilities for structural biology at spallation sources. This included the analysis of design parameters for protein crystallography as well as membrane diffraction instruments. These instruments are designed to be general user facilities and will be used by scientists from industry, universities, and other national laboratories.

  18. THERMAL NEUTRONIC REACTOR

    Science.gov (United States)

    Spinrad, B.I.

    1960-01-12

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

  19. 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.

  20. Neutron cameras for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, L.C.; Barnes, C.W.; Batistoni, P. [ITER San Diego Joint Work Site, La Jolla, CA (United States)] [and others

    1998-12-31

    Neutron cameras with horizontal and vertical views have been designed for ITER, based on systems used on JET and TFTR. The cameras consist of fan-shaped arrays of collimated flight tubes, with suitably chosen detectors situated outside the biological shield. The sight lines view the ITER plasma through slots in the shield blanket and penetrate the vacuum vessel, cryostat, and biological shield through stainless steel windows. This paper analyzes the expected performance of several neutron camera arrangements for ITER. In addition to the reference designs, the authors examine proposed compact cameras, in which neutron fluxes are inferred from {sup 16}N decay gammas in dedicated flowing water loops, and conventional cameras with fewer sight lines and more limited fields of view than in the reference designs. It is shown that the spatial sampling provided by the reference designs is sufficient to satisfy target measurement requirements and that some reduction in field of view may be permissible. The accuracy of measurements with {sup 16}N-based compact cameras is not yet established, and they fail to satisfy requirements for parameter range and time resolution by large margins.

  1. Neutrons are flying

    CERN Multimedia

    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...

  2. Compactness of Neutron Stars.

    Science.gov (United States)

    Chen, Wei-Chia; Piekarewicz, J

    2015-10-16

    Recent progress in the determination of both masses and radii of neutron stars is starting to place stringent constraints on the dense matter equation of state. In particular, new theoretical developments together with improved statistical tools seem to favor stellar radii that are significantly smaller than those predicted by models using purely nucleonic equations of state. Given that the underlying equation of state must also account for the observation of 2M⊙ neutron stars, theoretical approaches to the study of the dense matter equation of state are facing serious challenges. In response to this challenge, we compute the underlying equation of state associated with an assumed mass-radius template similar to the "common radius" assumption used in recent studies. Once such a mass-radius template is adopted, the equation of state follows directly from the implementation of Lindblom's algorithm; assumptions on the nature or composition of the dense stellar core are not required. By analyzing mass-radius profiles with a maximum mass consistent with observation and common radii in the 8-11 km range, a lower limit on the stellar radius of a 1.4M⊙ neutron star of RNS≳10.7  km is required to prevent the equation of state from violating causality.

  3. 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.

  4. Shielding for neutrons produced by medical linear accelerators

    Energy Technology Data Exchange (ETDEWEB)

    Rebello, Wilson F.; Silva, Ademir X. [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE). Programa de Engenharia Nuclear]. E-mail: rebello@ime.eb.br; ademir@con.ufrj.br

    2007-07-01

    The shielding system called Multileaf Shielding (MLS) was designed in Brazil to be used for protection patients, who undergo radiotherapy treatment, against undesired neutrons produced in the medical linear accelerator heads. During the conceiving of the MLS it was necessary to evaluate its efficiency. For that purpose, several simulations using the Monte Carlo N-particle radiation transport code, MCNP5, were made, in order to evaluate the response of the new shielding system. The results showed a significant neutron dose reduction after the inclusion of the MLS. This work aims to presenting these simulation results. (author)

  5. Bulk shielding benchmark experiment at Frascati neutron generator (FNG)

    Energy Technology Data Exchange (ETDEWEB)

    Batistoni, P.; Angelone, M.; Martone, M.; Pillon, M.; Rado, V. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Santamarina, A.; Abidi, I.; Gastaldi, B.; Martini, M.; Marquette, J.P. [CEA Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France)

    1994-11-01

    In the framework of the European Fusion Technology Program, ENEA (Italian Agency for New Technologies, Energy and the Environment) - Frascati and CEA (Commissariat a` l`Energie Atomique) - Cadarache, in collaboration performed a bulk shielding benchmark experiment, using the 14-MeV Frascati neutron generator (FNG), aimed at obtaining accurate experimental data for improving the nuclear data base and methods used in shielding designs. The experiment consisted of the irradiation of a stainless steel block by 14-MeV neutrons. The experimental results have been compared with numerical results calculated using both Sn and Monte Carlo transport codes and the cross section library EFF.1 (european fusion file).

  6. Solar Neutrons and Related Phenomena

    CERN Document Server

    Dorman, Lev

    2010-01-01

    This book presents the first comprehensive compilation and review of the extensive body of experimental and theoretical material on solar neutrons and related phenomena published in the scientific literature over the last sixty years. Phenomena related to solar neutrons are more specifically: the decay products of solar neutrons solar gamma rays generated in processes like nuclear reactions between solar energetic charged particles and matter of the solar atmosphere, as well as by the capture of solar neutrons by hydrogen atoms in the solar atmosphere the propagation of solar neutrons, solar gamma rays and other secondary particles through the solar photosphere, chromosphere and corona, as well as through interplanetary space and through the Earth's atmosphere. Models and simulations of particle acceleration, interactions, and propagation processes show that observations of solar neutrons and gamma rays in space and in the Earth's atmosphere yield essential and unique information on the source function of ene...

  7. 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....

  8. The physics of neutron stars.

    Science.gov (United States)

    Lattimer, J M; Prakash, M

    2004-04-23

    Neutron stars are some of the densest manifestations of massive objects in the universe. They are ideal astrophysical laboratories for testing theories of dense matter physics and provide connections among nuclear physics, particle physics, and astrophysics. Neutron stars may exhibit conditions and phenomena not observed elsewhere, such as hyperon-dominated matter, deconfined quark matter, superfluidity and superconductivity with critical temperatures near 10(10) kelvin, opaqueness to neutrinos, and magnetic fields in excess of 10(13) Gauss. Here, we describe the formation, structure, internal composition, and evolution of neutron stars. Observations that include studies of pulsars in binary systems, thermal emission from isolated neutron stars, glitches from pulsars, and quasi-periodic oscillations from accreting neutron stars provide information about neutron star masses, radii, temperatures, ages, and internal compositions.

  9. 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.

  10. Microcantilever-Enabled Neutron Detection

    Directory of Open Access Journals (Sweden)

    Kevin R. Kyle

    2014-04-01

    Full Text Available A new concept for neutron radiation detection was demonstrated using piezoresistive microcantilevers as the active sensing element. Microcantilevers were used to measure the tiny volumetric changes in a sensing material upon absorption of neutron radiation and transmutation into a new element. Highly ordered inorganic crystalline lattices of boron-rich materials dispersed in polymeric rubber matrices were shown to act as volumetric neutron transducers.

  11. APPARATUS FOR CONTROLLING NEUTRONIC REACTORS

    Science.gov (United States)

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

    1958-09-16

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

  12. On the neutron bursts origin.

    CERN Document Server

    Stenkin, Yu V

    2002-01-01

    The origin of the neutron bursts in Extensive Air Showers (EAS) is explained using results of the experiments and CORSIKA based Monte-Carlo simulations. It is shown that events with very high neutron multiplicity observed last years in neutron monitors as well as in surrounding detectors, are caused by usual EAS core with primary energies > 1 PeV. No exotic processes were needed for the explanation.

  13. Reactor neutrons in nuclear astrophysics

    Science.gov (United States)

    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.

  14. 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.

  15. Ultra-Cold Neutrons (UCN)

    Data.gov (United States)

    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...

  16. 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)

  17. T-violation in neutron optics

    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)

  18. MPACT Standard Input User s Manual, Version 2.2.0

    Energy Technology Data Exchange (ETDEWEB)

    Collins, Benjamin S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Fitzgerald, Andrew [Univ. of Michigan, Ann Arbor, MI (United States); Gehin, Jess C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Godfrey, Andrew T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Graham, Aaron [Univ. of Michigan, Ann Arbor, MI (United States); Jabaay, Daniel [Univ. of Michigan, Ann Arbor, MI (United States); Kelley, Blake W. [Univ. of Michigan, Ann Arbor, MI (United States); Kim, Kang [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kochunas, Brendan [Univ. of Michigan, Ann Arbor, MI (United States); Kulesza, Joel A. [Univ. of Michigan, Ann Arbor, MI (United States); Larsen, Edward W. [Univ. of Michigan, Ann Arbor, MI (United States); Liu, Yuxuan [Univ. of Michigan, Ann Arbor, MI (United States); Liu, Zhouyu [Univ. of Michigan, Ann Arbor, MI (United States); Martin, William R. [Univ. of Michigan, Ann Arbor, MI (United States); Nelson, Adam G. [Univ. of Michigan, Ann Arbor, MI (United States); Palmtag, Scott [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Rose, Michael [Univ. of Michigan, Ann Arbor, MI (United States); Saller, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Stimpson, Shane G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Trahan, Travis [Univ. of Michigan, Ann Arbor, MI (United States); Wang, J. W. [Univ. of Michigan, Ann Arbor, MI (United States); Wieselquist, William A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Young, Mitchell [Univ. of Michigan, Ann Arbor, MI (United States); Zhu, Ang [Univ. of Michigan, Ann Arbor, MI (United States)

    2016-08-01

    The MPACT (Michigan PArallel Charactistics based Transport) code is designed to perform high-fidelity light water reactor (LWR) analysis using whole-core pin-resolved neutron transport calculations on modern parallel-computing hardware. The code consists of several libraries which provide the functionality necessary to solve steady-state eigenvalue problems. Several transport capabilities are available within MPACT including both 2-D and 3-D Method of Characteristics (MOC). A three-dimensional whole core solution based on the 2D-1D solution method provides the capability for full core depletion calculations.

  19. MPACT VERA Input User s Manual, Version 2.2.0

    Energy Technology Data Exchange (ETDEWEB)

    Collins, Benjamin S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Fitzgerald, Andrew [Univ. of Michigan, Ann Arbor, MI (United States); Gehin, Jess C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Godfrey, Andrew T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Graham, Aaron [Univ. of Michigan, Ann Arbor, MI (United States); Jabaay, Daniel [Univ. of Michigan, Ann Arbor, MI (United States); Kelley, Blake W. [Univ. of Michigan, Ann Arbor, MI (United States); Kim, Kang [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kochunas, Brendan [Univ. of Michigan, Ann Arbor, MI (United States); Kulesza, Joel A. [Univ. of Michigan, Ann Arbor, MI (United States); Larsen, Edward W. [Univ. of Michigan, Ann Arbor, MI (United States); Liu, Yuxuan [Univ. of Michigan, Ann Arbor, MI (United States); Liu, Zhouyu [Univ. of Michigan, Ann Arbor, MI (United States); Martin, William R. [Univ. of Michigan, Ann Arbor, MI (United States); Nelson, Adam G. [Univ. of Michigan, Ann Arbor, MI (United States); Palmtag, Scott [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Rose, Michael [Univ. of Michigan, Ann Arbor, MI (United States); Saller, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Stimpson, Shane G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Trahan, Travis [Univ. of Michigan, Ann Arbor, MI (United States); Wang, J. W. [Univ. of Michigan, Ann Arbor, MI (United States); Wieselquist, William A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Young, Mitchell [Univ. of Michigan, Ann Arbor, MI (United States); Zhu, Ang [Univ. of Michigan, Ann Arbor, MI (United States)

    2016-06-09

    The MPACT (Michigan PArallel Charactistics based Transport) code is designed to perform high-fidelity light water reactor (LWR) analysis using whole-core pin-resolved neutron transport calculations on modern parallel-computing hardware. The code consists of several libraries which provide the functionality necessary to solve steady-state eigenvalue problems. Several transport capabilities are available within MPACT including both 2-D and 3-D Method of Characteristics (MOC). A three-dimensional whole core solution based on the 2D-1D solution method provides the capability for full core depletion calculations.

  20. Benchmarking PARTISN with Analog Monte Carlo: Moments of the Neutron Number and the Cumulative Fission Number Probability Distributions

    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.

  1. Calculation of neutron fluence to dose equivalent conversion coefficients using GEANT4; Calculo de coeficientes de fluencia de neutrons para equivalente de dose individual utilizando o GEANT4

    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)

  2. 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.

  3. A theoretical model for predicting neutron fluxes for cyclic Neutron ...

    African Journals Online (AJOL)

    A theoretical model has been developed for prediction of thermal neutron fluxes required for cyclic irradiations of a sample to obtain the same activity previously used for the detection of any radionuclide of interest. The model is suitable for radiotracer production or for long-lived neutron activation products where the ...

  4. 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)

  5. Local transport method for hybrid diffusion-transport calculations in 2-D cylindrical (R, THETA) geometry

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Dingkang; Rahnema, Farzad, E-mail: Dingkang.zhang@gatech.edu, E-mail: farzad@gatech.edu [Nuclear and Radiological Engineering & Medical Physics Programs, Georgia Institute of Technology, Atlanta, GA (United States); Ougouag, Abderrfi M., E-mail: Abderrafi.Ougouag@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States)

    2011-07-01

    A response-based local transport method has been developed in 2-D (r, θ) geometry for coupling to any coarse-mesh (nodal) diffusion method/code. Monte Carlo method is first used to generate a (pre-computed) the response function library for each unique coarse mesh in the transport domain (e.g., the outer reflector region of the Pebble Bed Reactor). The scalar flux and net current at the diffusion/transport interface provided by the diffusion method are used as an incoming surface source to the transport domain. A deterministic iterative sweeping method together with the response function library is utilized to compute the local transport solution within all transport coarse meshes. After the partial angular currents crossing the coarse mesh surfaces are converged, albedo coefficients are computed as boundary conditions for the diffusion methods. The iteration on the albedo boundary condition (for the diffusion method via transport) and the incoming angular flux boundary condition (for the transport via diffusion) is continued until convergence is achieved. The method was tested for in a simplified 2-D (r, θ) pebble bed reactor problem consisting of an inner reflector, an annular fuel region and a controlled outer reflector. The comparisons have shown that the results of the response-function-based transport method agree very well with a direct MCNP whole core solution. The agreement in coarse mesh averaged flux was found to be excellent: relative difference of about 0.18% and a maximum difference of about 0.55%. Note that the MCNP uncertainty was less than 0.1%. (author)

  6. 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)

  7. Electromagnetic Chirps from Neutron Star–Black Hole Mergers

    Science.gov (United States)

    Schnittman, Jeremy D.; Dal Canton, Tito; Camp, Jordan; Tsang, David; Kelly, Bernard J.

    2018-02-01

    We calculate the electromagnetic signal of a gamma-ray flare coming from the surface of a neutron star shortly before merger with a black hole companion. Using a new version of the Monte Carlo radiation transport code Pandurata that incorporates dynamic spacetimes, we integrate photon geodesics from the neutron star surface until they reach a distant observer or are captured by the black hole. The gamma-ray light curve is modulated by a number of relativistic effects, including Doppler beaming and gravitational lensing. Because the photons originate from the inspiraling neutron star, the light curve closely resembles the corresponding gravitational waveform: a chirp signal characterized by a steadily increasing frequency and amplitude. We propose to search for these electromagnetic chirps using matched filtering algorithms similar to those used in LIGO data analysis.

  8. Neutron fluctuations a treatise on the physics of branching processes

    CERN Document Server

    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...

  9. Quantized Ultracold Neutrons in Rough Waveguides: GRANIT Experiments and Beyond

    Directory of Open Access Journals (Sweden)

    M. Escobar

    2014-01-01

    Full Text Available We apply our general theory of transport in systems with random rough boundaries to gravitationally quantized ultracold neutrons in rough waveguides as in GRANIT experiments (ILL, Grenoble. We consider waveguides with roughness in both two and one dimensions (2D and 1D. In the biased diffusion approximation the depletion times for the gravitational quantum states can be easily expressed via each other irrespective of the system parameters. The calculation of the exit neutron count reduces to evaluation of a single constant which contains a complicated integral of the correlation function of surface roughness. In the case of 1D roughness (random grating this constant is calculated analytically for common types of the correlation functions. The results obey simple scaling relations which are slightly different in 1D and 2D. We predict the exit neutron count for the new GRANIT cell.

  10. The MCLIB library: Monte Carlo simulation of neutron scattering instruments

    Energy Technology Data Exchange (ETDEWEB)

    Seeger, P.A.

    1995-09-01

    Monte Carlo is a method to integrate over a large number of variables. Random numbers are used to select a value for each variable, and the integrand is evaluated. The process is repeated a large number of times and the resulting values are averaged. For a neutron transport problem, first select a neutron from the source distribution, and project it through the instrument using either deterministic or probabilistic algorithms to describe its interaction whenever it hits something, and then (if it hits the detector) tally it in a histogram representing where and when it was detected. This is intended to simulate the process of running an actual experiment (but it is much slower). This report describes the philosophy and structure of MCLIB, a Fortran library of Monte Carlo subroutines which has been developed for design of neutron scattering instruments. A pair of programs (LQDGEOM and MC{_}RUN) which use the library are shown as an example.

  11. The neutron instrument Monte Carlo library MCLIB: Recent developments

    Energy Technology Data Exchange (ETDEWEB)

    Seeger, P.A.; Daemen, L.L.; Hjelm, R.P. Jr.; Thelliez, T.G.

    1998-12-31

    A brief review is given of the developments since the ICANS-XIII meeting made in the neutron instrument design codes using the Monte Carlo library MCLIB. Much of the effort has been to assure that the library and the executing code MC{_}RUN connect efficiently with the World Wide Web application MC-WEB as part of the Los Alamos Neutron Instrument Simulation Package (NISP). Since one of the most important features of MCLIB is its open structure and capability to incorporate any possible neutron transport or scattering algorithm, this document describes the current procedure that would be used by an outside user to add a feature to MCLIB. Details of the calling sequence of the core subroutine OPERATE are discussed, and questions of style are considered and additional guidelines given. Suggestions for standardization are solicited, as well as code for new algorithms.

  12. On the design of a cold neutron irradiator (CNI) for quantitative materials characterization

    Energy Technology Data Exchange (ETDEWEB)

    Atwood, Alexander Grover [Cornell Univ., Ithaca, NY (United States)

    1997-08-01

    A design study of a cold neutron irradiator (CNI) for materials characterization using prompt gamma-ray neutron activation analysis (PGNAA) is presented. Using 252Cf neutron sources in a block of moderator, a portion of which is maintained at a cryogenic temperature, the CNI employs cold neutrons instead of thermal neutrons to enhance the neutron capture reaction rate in a sample. Capture gamma rays are detected in an HPGe photon detector. Optimization of the CNI with respect to elemental sensitivity (counts per mg) is the primary goal of this design study. Monte Carlo simulation of radiation transport, by means of the MCNP code and the ENDF/B cross-section libraries, is used to model the CNI. A combination of solid methane at 22 K, room-temperature polyethylene, and room-temperature beryllium has been chosen for the neutron delivery subsystem of the CNI. Using four 250-microgram 252Cf neutron sources, with a total neutron emission rate of 2.3 x 109 neutrons/s, a thermal-equivalent neutron flux of 1.7 x 107 neutrons/cm2-s in an internally located cylindrical sample space of diameter 6.5 cm and height 6.0 cm is predicted by MCNP calculations. A cylindrical port with an integral annular collimator composed of bismuth, lead, polyethylene, and lithium carbonate, is located between the sample and the detector. Calculations have been performed of gamma-ray and neutron transport in the port and integral collimator with the objective of optimizing the statistical precision with which one can measure elemental masses in the sample while also limiting the fast neutron flux incident upon the HPGe detector to a reasonable level. The statistical precision with which one can measure elemental masses can be enhanced by a factor of between 2.3 and 5.3 (depending on the origin of the background gamma rays) compared with a neutron irradiator identical to the CNI except for the replacement of the cryogenic solid methane by room

  13. 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

  14. Neutron quality factor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-06-01

    Both the International Commission on Radiological Protection (ICRP) and the National Council on Radiation Protection and Measurements (NCRP) have recommended that the radiation quality weighting factor for neutrons (Q{sub n}, or the corresponding new modifying factor, w{sub R}) be increased by a value of two for most radiation protection practices. This means an increase in the recommended value for Q{sub n} from a nominal value of 10 to a nominal value of 20. This increase may be interpreted to mean that the biological effectiveness of neutrons is two times greater than previously thought. A decision to increase the value of Q{sub n} will have a major impact on the regulations and radiation protection programs of Federal agencies responsible for the protection of radiation workers. Therefore, the purposes of this report are: (1) to examine the general concept of {open_quotes}quality factor{close_quotes} (Q) in radiation protection and the rationale for the selection of specific values of Q{sub n}; and (2) to make such recommendations to the Federal agencies, as appropriate. This report is not intended to be an exhaustive review of the scientific literature on the biological effects of neutrons, with the aim of defending a particular value for Q{sub n}. Rather, the working group examined the technical issues surrounding the current recommendations of scientific advisory bodies on this matter, with the aim of determining if these recommendations should be adopted by the Federal agencies. Ultimately, the group concluded that there was no compelling basis for a change in Q{sub n}. The report was prepared by Federal scientists working under the auspices of the Science Panel of the Committee on Interagency Radiation Research and Policy Coordination (CIRRPC).

  15. Handbook of neutron optics

    CERN Document Server

    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

  16. Advanced Neutron Spectrometer

    Science.gov (United States)

    Christl, Mark; Dobson, Chris; Norwood, Joseph; Kayatin, Matthew; Apple, Jeff; Gibson, Brian; Dietz, Kurt; Benson, Carl; Smith, Dennis; Howard, David; hide

    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.

  17. NEUTRONIC REACTOR CONSTRUCTION

    Science.gov (United States)

    Vernon, H.C.; Goett, J.J.

    1958-09-01

    A cover device is described for the fuel element receiving tube of a neutronic reactor of the heterogeneous, water cooled type wherein said tubes are arranged in a moderator with their longitudinal axes vertical. The cover is provided with means to support a rod-type fuel element from the bottom thereof and means to lock the cover in place, the latter being adapted for remote operation. This cover device is easily removable and seals the opening in the upper end of the fuel tube against leakage of coolant.

  18. Magnetic Fields of Neutron Stars

    Science.gov (United States)

    Konar, Sushan

    2017-09-01

    This article briefly reviews our current understanding of the evolution of magnetic fields in neutron stars, which basically defines the evolutionary pathways between different observational classes of neutron stars. The emphasis here is on the evolution in binary systems and the newly emergent classes of millisecond pulsars.

  19. Physics of Neutron Star Crusts

    Directory of Open Access Journals (Sweden)

    Chamel Nicolas

    2008-12-01

    Full Text Available The physics of neutron star crusts is vast, involving many different research fields, from nuclear and condensed matter physics to general relativity. This review summarizes the progress, which has been achieved over the last few years, in modeling neutron star crusts, both at the microscopic and macroscopic levels. The confrontation of these theoretical models with observations is also briefly discussed.

  20. Neutron and P, T symmetry

    Energy Technology Data Exchange (ETDEWEB)

    Masuda, Y. [National Lab. for High Energy Physics, Tsukuba, Ibaraki (Japan)

    1996-05-01

    New ideas for experiments to improve the T-violation limit by a factor of 10 to 100 is discussed for a intensive spallation neutron source. The methods to improve the limit of the right-handed current and the neutron lifetime are also discussed. (author)

  1. High power neutron production targets

    Energy Technology Data Exchange (ETDEWEB)

    Wender, S. [Los Alamos National Lab., NM (United States)

    1996-06-01

    The author describes issues of concern in the design of targets and associated systems for high power neutron production facilities. The facilities include uses for neutron scattering, accelerator driven transmutation, accelerator production of tritium, short pulse spallation sources, and long pulse spallation sources. Each of these applications requires a source with different design needs and consequently different implementation in practise.

  2. Neutron protein crystallography in JAERI

    Indian Academy of Sciences (India)

    crystals and a database of hydrogen and hydration have also been developed. In the near future, a pulsed neutron diffractometer for biological macromolecules has been proposed at J-PARC in JAERI. Keywords. Crystallization macromolecules; crystallographic databases; crystal struc- ture neutron diffraction and scattering ...

  3. Physics of Neutron Star Crusts.

    Science.gov (United States)

    Chamel, Nicolas; Haensel, Pawel

    2008-01-01

    The physics of neutron star crusts is vast, involving many different research fields, from nuclear and condensed matter physics to general relativity. This review summarizes the progress, which has been achieved over the last few years, in modeling neutron star crusts, both at the microscopic and macroscopic levels. The confrontation of these theoretical models with observations is also briefly discussed.

  4. Magnetic fields in Neutron Stars

    NARCIS (Netherlands)

    Viganò, D.; Pons, J.A.; Miralles, J.A.; Rea, N.; Cenarro, A.J.; Figueras, F.; Hernández-Monteagudo, J.; Bueno, T.; Valdivielso, L.

    2015-01-01

    Isolated neutron stars show a diversity in timing and spectral properties, which has historically led to a classification in different sub-classes. The magnetic field plays a key role in many aspects of the neutron star phenomenology: it regulates the braking torque responsible for their timing

  5. Neutrons for technology and science

    Energy Technology Data Exchange (ETDEWEB)

    Aeppli, G.

    1995-10-01

    We reviewed recent work using neutrons generated at nuclear reactors an accelerator-based spallation sources. Provided that large new sources become available, neutron beams will continue to have as great an impact on technology and science as in the past.

  6. KAON CONDENSATION IN NEUTRON STARS.

    Energy Technology Data Exchange (ETDEWEB)

    RAMOS,A.; SCHAFFNER-BIELICH,J.; WAMBACH,J.

    2001-04-24

    We discuss the kaon-nucleon interaction and its consequences for the change of the properties of the kaon in the medium. The onset of kaon condensation in neutron stars under various scenarios as well its effects for neutron star properties are reviewed.

  7. High-Resolution Neutron Radiography Analysis of Proton Exchange Membrane Fuel Cells

    Science.gov (United States)

    Hussey, D. S.; Jacobson, D. L.

    Neutron radiography enables direct visualization and quantification of many water transport phenomena in proton exchange membrane fuel cells (PEMFCs). The advantage of the technique is that neutrons have a long penetration length through most common PEMFC materials of construction (with a 1/e length of about 11 cm for aluminum), while having a relatively short 1/e length for water (of order 3 mm). This sensitivity to water enables precise measurements via neutron radiography of the water content in an operating PEMFC that are primarily limited by systematic measurement uncertainties. Recent advances in the spatial resolution of neutron detector technology enable direct measurement of the through-plane water content. This new data provides gas diffusion layer water profiles that can serve as input or comparison data for a large class of one-dimensional PEMFC models. In this article, the technique of neutron radiography is discussed, with an emphasis on the quantitative image analysis of the through-plane water content.

  8. 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.

  9. A Monte Carlo simulation of monoenergetic neutrons traversing rectangular and spherical polyethylene moderators.

    Science.gov (United States)

    Monk, S D; Selwood, M

    2011-03-01

    A Monte Carlo-based simulation of the transport of a series of monoenergetic neutron sources through first a rectangular block of 0.93 g cm(-3) density polyethylene and secondly through a sphere made of the same substance is presented here. In both instances, the neutron fields are monitored at closely spread intervals through the moderator mass, producing a lot of data in the process. To reduce the amount of data presented, a figure of merit is created by estimating the cross section for each discrete neutron energy and by applying this to the number of neutrons present of each energy giving an arbitrary response figure. This work was undertaken in order to aid the design and development of a novel neutron spectrometer.

  10. Measurement of neutron dose on a fusion reactor shield using TLD-300 phosphors

    Energy Technology Data Exchange (ETDEWEB)

    Angelone, M.; Batistoni, P.; Martone, M.; Pillon, M.; Rado, V. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Esposito, A.; Pelliccioni, M. [Comitato Nazionale per l`Energia Nucleare, Frascati (Italy). Lab. Nazionali di Frascati

    1997-09-01

    A benchmark experiment was carried out at the 14 MeV Frascati Neutron Generator (FNG) to measure the nuclear heating in an experimental assembly simulating a fusion reactor shield and its superconducting coils. TLD-300 detectors were used allowing the neutron and gamma contribution to the total nuclear heating to be separated by using the two-peak method. The experiment aimed at validating the EFF and FENDL neutron transport cross section files presently available for nuclear heating calculations of fusion reactor shields. The calculations were performed using the Monte Carlo code MCNP. The measured and calculated gamma absorbed dose agree well while differences up to 30% were found for the neutron absorbed dose when the EFF file is used. The discrepancy increases up to 50% when the FENDL file is used to calculate the neutron absorbed dose. (author).

  11. Measurements of elastic scattering cross sections of carbon, iron and lead for 75 MeV Neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Ibaraki, Masanobu; Baba, Mamoru; Nauchi, Yasushi; Miura, Takako; Hirasawa, Yoshitaka; Hirakawa, Naohiro [Department of Quantum Science and Energy Engineering, Tohoku University, Sendai, Miyagi (Japan); Nakashima, Hiroshi; Meigo, Shin-ichiro; Tanaka, Susumu

    1999-03-01

    We have performed the measurements of elastic scattering cross sections of carbon, iron and lead for 75 MeV neutrons using a {sup 7}Li(p, n) quasi-monoenergetic neutron source. Elastically scattered neutrons were measured with a time of flight method (TOF) using five liquid scintillation detectors. The data were obtained at 25 laboratory angles between 2.6deg and 53.0deg. The experimental data were compared with the neutron cross section libraries, systematics used in cascade/transport codes and optical model calculations. (author)

  12. 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

  13. Measuring neutron spectra in radiotherapy using the nested neutron spectrometer.

    Science.gov (United States)

    Maglieri, Robert; Licea, Angel; Evans, Michael; Seuntjens, Jan; Kildea, John

    2015-11-01

    Out-of-field neutron doses resulting from photonuclear interactions in the head of a linear accelerator pose an iatrogenic risk to patients and an occupational risk to personnel during radiotherapy. To quantify neutron production, in-room measurements have traditionally been carried out using Bonner sphere systems (BSS) with activation foils and TLDs. In this work, a recently developed active detector, the nested neutron spectrometer (NNS), was tested in radiotherapy bunkers. The NNS is designed for easy handling and is more practical than the traditional BSS. Operated in current-mode, the problem of pulse pileup due to high dose-rates is overcome by measuring current, similar to an ionization chamber. In a bunker housing a Varian Clinac 21EX, the performance of the NNS was evaluated in terms of reproducibility, linearity, and dose-rate effects. Using a custom maximum-likelihood expectation-maximization algorithm, measured neutron spectra at various locations inside the bunker were then compared to Monte Carlo simulations of an identical setup. In terms of dose, neutron ambient dose equivalents were calculated from the measured spectra and compared to bubble detector neutron dose equivalent measurements. The NNS-measured spectra for neutrons at various locations in a treatment room were found to be consistent with expectations for both relative shape and absolute magnitude. Neutron fluence-rate decreased with distance from the source and the shape of the spectrum changed from a dominant fast neutron peak near the Linac head to a dominant thermal neutron peak in the moderating conditions of the maze. Monte Carlo data and NNS-measured spectra agreed within 30% at all locations except in the maze where the deviation was a maximum of 40%. Neutron ambient dose equivalents calculated from the authors' measured spectra were consistent (one standard deviation) with bubble detector measurements in the treatment room. The NNS may be used to reliably measure the neutron

  14. Grand unification of neutron stars.

    Science.gov (United States)

    Kaspi, Victoria M

    2010-04-20

    The last decade has shown us that the observational properties of neutron stars are remarkably diverse. From magnetars to rotating radio transients, from radio pulsars to isolated neutron stars, from central compact objects to millisecond pulsars, observational manifestations of neutron stars are surprisingly varied, with most properties totally unpredicted. The challenge is to establish an overarching physical theory of neutron stars and their birth properties that can explain this great diversity. Here I survey the disparate neutron stars classes, describe their properties, and highlight results made possible by the Chandra X-Ray Observatory, in celebration of its 10th anniversary. Finally, I describe the current status of efforts at physical "grand unification" of this wealth of observational phenomena, and comment on possibilities for Chandra's next decade in this field.

  15. Neutron matter, neutron pairing, and neutron drops based on chiral effective field theory interactions

    Energy Technology Data Exchange (ETDEWEB)

    Krueger, Thomas

    2016-10-19

    The physics of neutron-rich systems is of great interest in nuclear and astrophysics. Precise knowledge of the properties of neutron-rich nuclei is crucial for understanding the synthesis of heavy elements. Infinite neutron matter determines properties of neutron stars, a final stage of heavy stars after a core-collapse supernova. It also provides a unique theoretical laboratory for nuclear forces. Strong interactions are determined by quantum chromodynamics (QCD). However, QCD is non-perturbative at low energies and one presently cannot directly calculate nuclear forces from it. Chiral effective field theory circumvents these problems and connects the symmetries of QCD to nuclear interactions. It naturally and systematically includes many-nucleon forces and gives access to uncertainty estimates. We use chiral interactions throughout all calculation in this thesis. Neutron stars are very extreme objects. The densities in their interior greatly exceed those in nuclei. The exact composition and properties of neutron stars is still unclear but they consist mainly of neutrons. One can explore neutron stars theoretically with calculations of neutron matter. In the inner core of neutron stars exist very high densities and thus maybe exotic phases of matter. To investigate whether there exists a phase transition to such phases even at moderate densities we study the chiral condensate in neutron matter, the order parameter of chiral symmetry breaking, and find no evidence for a phase transition at nuclear densities. We also calculate the more extreme system of spin-polarised neutron matter. With this we address the question whether there exists such a polarised phase in neutron stars and also provide a benchmark system for lattice QCD. We find spin-polarised neutron matter to be an almost non-interacting Fermi gas. To understand the cooling of neutron stars neutron pairing is of great importance. Due to the high densities especially triplet pairing is of interest. We

  16. Dark neutron stars

    Science.gov (United States)

    Jones, P. B.

    2017-06-01

    There is good evidence that electron-positron pair formation is not present in that section of the pulsar open magnetosphere, which is the source of coherent radio emission, but the possibility of two-photon pair creation in an outer gap remains. Calculation of transition rates for this process based on measured whole-surface temperatures, combined with a survey of γ-ray, X-ray and optical luminosities, expressed per primary beam lepton, shows that few Fermi-LAT pulsars have significant outer-gap pair creation. For radio-loud pulsars with positive polar-cap corotational charge density and an ion-proton plasma, there must be an outward flow of electrons from some other part of the magnetosphere to maintain a constant net charge on the star. In the absence of pair creation, it is likely that this current is the source of GeV γ-emission observed by the Fermi-LAT and its origin is in the region of the outer gap. With negative polar-cap corotational charge density, the compensating current in the absence of pair creation can consist only of ions or protons. These neutron stars are likely to be radio-quiet, have no observable γ-emission, and hence can be described as dark neutron stars.

  17. Fast neutron environments.

    Energy Technology Data Exchange (ETDEWEB)

    Buchheit, Thomas Edward; Kotula, Paul Gabriel; Lu, Ping; Brewer, Luke N. (Naval Postgraduate School, Monterey, CA); Goods, Steven Howard (Sandia National Laboratories, Livermore, CA); Foiles, Stephen Martin; Puskar, Joseph David; Hattar, Khalid Mikhiel; Doyle, Barney Lee; Boyce, Brad Lee; Clark, Blythe G.

    2011-10-01

    The goal of this LDRD project is to develop a rapid first-order experimental procedure for the testing of advanced cladding materials that may be considered for generation IV nuclear reactors. In order to investigate this, a technique was developed to expose the coupons of potential materials to high displacement damage at elevated temperatures to simulate the neutron environment expected in Generation IV reactors. This was completed through a high temperature high-energy heavy-ion implantation. The mechanical properties of the ion irradiated region were tested by either micropillar compression or nanoindentation to determine the local properties, as a function of the implantation dose and exposure temperature. In order to directly compare the microstructural evolution and property degradation from the accelerated testing and classical neutron testing, 316L, 409, and 420 stainless steels were tested. In addition, two sets of diffusion couples from 316L and HT9 stainless steels with various refractory metals. This study has shown that if the ion irradiation size scale is taken into consideration when developing and analyzing the mechanical property data, significant insight into the structural properties of the potential cladding materials can be gained in about a week.

  18. Multiple-wavelength neutron holography with pulsed neutrons.

    Science.gov (United States)

    Hayashi, Kouichi; Ohoyama, Kenji; Happo, Naohisa; Matsushita, Tomohiro; Hosokawa, Shinya; Harada, Masahide; Inamura, Yasuhiro; Nitani, Hiroaki; Shishido, Toetsu; Yubuta, Kunio

    2017-08-01

    Local structures around impurities in solids provide important information for understanding the mechanisms of material functions, because most of them are controlled by dopants. For this purpose, the x-ray absorption fine structure method, which provides radial distribution functions around specific elements, is most widely used. However, a similar method using neutron techniques has not yet been developed. If one can establish a method of local structural analysis with neutrons, then a new frontier of materials science can be explored owing to the specific nature of neutron scattering-that is, its high sensitivity to light elements and magnetic moments. Multiple-wavelength neutron holography using the time-of-flight technique with pulsed neutrons has great potential to realize this. We demonstrated multiple-wavelength neutron holography using a Eu-doped CaF2 single crystal and obtained a clear three-dimensional atomic image around trivalent Eu substituted for divalent Ca, revealing an interesting feature of the local structure that allows it to maintain charge neutrality. The new holography technique is expected to provide new information on local structures using the neutron technique.

  19. Neutron/muon correlation functions to improve neutron detection capabilities outside nuclear facilities

    Science.gov (United States)

    Ordinario, Donald Thomas

    The natural neutron background rate is largely due to cosmic ray interactions in the atmosphere and the subsequent neutron emission from the interaction products. The neutron background is part of a larger cosmic radiation shower that also includes electrons, gamma rays, and muons. Since neutrons interact much differently than muons in building materials, the muon and neutron fluence rates in the natural background can be compared to the measured muon and neutron fluence rate when shielded by common building materials. The simultaneous measurement of muon and neutron fluence rates might allow for an earlier identification of man-made neutron sources, such as hidden nuclear materials. This study compares natural background neutron rates to computer simulated neutron rates shielded by common structural and building materials. The characteristic differences between neutrons and muons resulted in different attenuation properties under the same shielded conditions. Correlation functions between cosmic ray generated neutrons and muons are then used to predict neutron fluence rates in different urban environments.

  20. Characterization of source neutrons produced by Frascati neutron generator

    Energy Technology Data Exchange (ETDEWEB)

    Pillon, M.; Angelone, M.; Martone, M.; Rado, V. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy)

    1994-11-01

    The characterization of the 14 MeV neutron source of the Frascati neutron generator (FNG) has been carried out using an absolutely calibrated U{sup Nat} micro-fission chamber with plane electrodes and {sup 93}Nb(n,2n){sup 92m}Nb activation reaction. The reaction rates have been measured at different angles with respect to the beam direction. The experimental results are compared with the corresponding quantities calculated by using a Monte Carlo code with an ad hoc source model. The calculated neutron emission reproduces satisfactory well the measured one. Improving of the used source model is discussed.

  1. Spectroscopic and neutron detection properties of rare earth and titanium doped LiAlO 2 single crystals

    Energy Technology Data Exchange (ETDEWEB)

    Dickens, Peter T.; Marcial, José; McCloy, John; McDonald, Benjamin S.; Lynn, Kelvin G.

    2017-10-01

    In this study, LiAlO2 crystals doped with rare-earth elements and Ti were produced by the CZ method and spectroscopic and neutron detection properties were investigated. Photoluminescence revealed no clear luminescent activation of LiAlO2 by the rare-earth dopants though some interesting luminescence was observed from secondary phases within the crystal. Gamma-ray pulse height spectra collected using a 137Cs source exhibited only a Compton edge for the crystals. Neutron modeling using Monte Carlo N-Particle Transport Code revealed most neutrons used in the detection setup are thermalized, and while using natural lithium in the crystal growth, which contains 7.6 % 6Li, a 10 mm Ø by 10 mm sample of LiAlO2 has a 70.7 % intrinsic thermal neutron capture efficiency. Furthermore, the pulse height spectra collected using a 241Am-Be neutron source demonstrated a distinct neutron peak.

  2. Power Burst Facility/Boron Neutron Capture Therapy Program for cancer treatment

    Energy Technology Data Exchange (ETDEWEB)

    Ackermann, A.L. (ed.); Dorn, R.V. III.

    1990-08-01

    This report discusses monthly progress in the Power Boron Facility/Boron Neutron Capture Therapy (PBF/BNCT) Program for Cancer Treatment. Highlights of the PBF/BNCT Program during August 1990 include progress within the areas of: Gross Boron Analysis in Tissue, Blood, and Urine, boron microscopic (subcellular) analytical development, noninvasive boron quantitative determination, analytical radiation transport and interaction modeling for BNCT, large animal model studies, neutron source and facility preparation, administration and common support and PBF operations.

  3. Power Burst Facility/Boron Neutron Capture Therapy program for cancer treatment, Volume 4, No. 7

    Energy Technology Data Exchange (ETDEWEB)

    Ackermann, A.L. (ed.)

    1990-07-01

    This report discusses the monthly progress of the Power Burst Facility/Boron Neutron Capture Therapy (PBF/BNLT) program for cancer treatment. Highlights of the PBF/BNCT Program during July 1990 include progress within the areas of: Gross boron analysis in tissue, blood, and urine; noninvasive boron quantitative determination; analytical radiation transport and interaction modeling for BNCT; large animal model studies; neutron source and facility preparation; administration and common support and PBF operations.

  4. Progress toward the development and testing of source reconstruction methods for NIF neutron imaging.

    Science.gov (United States)

    Loomis, E N; Grim, G P; Wilde, C; Wilson, D C; Morgan, G; Wilke, M; Tregillis, I; Merrill, F; Clark, D; Finch, J; Fittinghoff, D; Bower, D

    2010-10-01

    Development of analysis techniques for neutron imaging at the National Ignition Facility is an important and difficult task for the detailed understanding of high-neutron yield inertial confinement fusion implosions. Once developed, these methods must provide accurate images of the hot and cold fuels so that information about the implosion, such as symmetry and areal density, can be extracted. One method under development involves the numerical inversion of the pinhole image using knowledge of neutron transport through the pinhole aperture from Monte Carlo simulations. In this article we present results of source reconstructions based on simulated images that test the methods effectiveness with regard to pinhole misalignment.

  5. Constraining gravity with hadron physics: neutron stars, modified gravity and gravitational waves

    Directory of Open Access Journals (Sweden)

    Llanes-Estrada Felipe J.

    2017-01-01

    Full Text Available The finding of Gravitational Waves (GW by the aLIGO scientific and VIRGO collaborations opens opportunities to better test and understand strong interactions, both nuclear-hadronic and gravitational. Assuming General Relativity holds, one can constrain hadron physics at a neutron star. But precise knowledge of the Equation of State and transport properties in hadron matter can also be used to constrain the theory of gravity itself. I review a couple of these opportunities in the context of modified f (R gravity, the maximum mass of neutron stars, and progress in the Equation of State of neutron matter from the chiral effective field theory of QCD.

  6. Developing an interface between MCNP and McStas for simulation of neutron moderators

    DEFF Research Database (Denmark)

    Klinkby, Esben Bryndt; Lauritzen, Bent; Nonbøl, Erik

    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 typic...... moderator geometries, interference between beamlines as well as shielding requirements along the neutron guides. In this paper different possible interfaces between McStas and MCNP/X are discussed and first preliminary performance results are shown....

  7. The MCNP code in planning and interpretation of thermal neutron pulsed experiments

    Energy Technology Data Exchange (ETDEWEB)

    Dabrowska, J.; Drozdowicz, K.; Woznicka, U. [The Henryk Niewodniczanski Inst. of Nuclear Physics, Krakow (Poland)

    2001-07-01

    A possibility to use the MCNP code to support planning and interpretation of the neutron pulsed experiments is discussed. Example of the simulated experiments for polyethylene are shown and compared to the real experimental results. A usefulness of the MCNP code for consideration of the time-dependent thermal neutron fields is stated. There are indicated some properties of the code and cross-section libraries which create problems or make impossible its using for many hydrogenous materials when the thermal neutron transport has to be considered with a high accuracy. (orig.)

  8. THERMAL NEUTRON INTENSITIES IN SOILS IRRADIATED BY FAST NEUTRONS FROM POINT SOURCES. (R825549C054)

    Science.gov (United States)

    Thermal-neutron fluences in soil are reported for selected fast-neutron sources, selected soil types, and selected irradiation geometries. Sources include 14 MeV neutrons from accelerators, neutrons from spontaneously fissioning 252Cf, and neutrons produced from alp...

  9. Performance verification of an epithermal neutron flux monitor using accelerator-based BNCT neutron sources

    Science.gov (United States)

    Guan, X.; Murata, I.; Wang, T.

    2017-09-01

    The performance of an epithermal neutron flux monitor developed for boron neutron capture therapy (BNCT) is verified by Monte Carlo simulations using accelerator-based neutron sources (ABNSs). The results indicate that the developed epithermal neutron flux monitor works well and it can be efficiently used in practical applications to measure the epithermal neutron fluxes of ABNSs in a high accuracy.

  10. Determination of neutron buildup factor using analytical solution of one-dimensional neutron diffusion equation in cylindrical geometry

    Energy Technology Data Exchange (ETDEWEB)

    Fernandes, Julio Cesar L.; Vilhena, Marco Tullio, E-mail: julio.lombaldo@ufrgs.b, E-mail: vilhena@pq.cnpq.b [Universidade Federal do Rio Grande do Sul (DMPA/UFRGS), Porto Alegre, RS (Brazil). Dept. de Matematica Pura e Aplicada. Programa de Pos Graduacao em Matematica Aplicada; Borges, Volnei; Bodmann, Bardo Ernest, E-mail: bardo.bodmann@ufrgs.b, E-mail: borges@ufrgs.b [Universidade Federal do Rio Grande do Sul (PROMEC/UFRGS), Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica

    2011-07-01

    The principal idea of this work, consist on formulate an analytical method to solved problems for diffusion of neutrons with isotropic scattering in one-dimensional cylindrical geometry. In this area were develop many works that study the same problem in different system of coordinates as well as cartesian system, nevertheless using numerical methods to solve the shielding problem. In view of good results in this works, we starting with the idea that we can represent a source in the origin of the cylindrical system by a Delta Dirac distribution, we describe the physical modeling and solved the neutron diffusion equation inside of cylinder of radius R. For the case of transport equation, the formulation of discrete ordinates S{sub N} consists in discretize the angular variables in N directions and in using a quadrature angular set for approximate the sources of scattering, where the Diffusion equation consist on S{sub 2} approximated transport equation in discrete ordinates. We solved the neutron diffusion equation with an analytical form by the finite Hankel transform. Was presented also the build-up factor for the case that we have neutron flux inside the cylinder. (author)

  11. Studies and modeling of cold neutron sources; Etude et modelisation des sources froides de neutron

    Energy Technology Data Exchange (ETDEWEB)

    Campioni, G

    2004-11-15

    With the purpose of updating knowledge in the fields of cold neutron sources, the work of this thesis has been run according to the 3 following axes. First, the gathering of specific information forming the materials of this work. This set of knowledge covers the following fields: cold neutron, cross-sections for the different cold moderators, flux slowing down, different measurements of the cold flux and finally, issues in the thermal analysis of the problem. Secondly, the study and development of suitable computation tools. After an analysis of the problem, several tools have been planed, implemented and tested in the 3-dimensional radiation transport code Tripoli-4. In particular, a module of uncoupling, integrated in the official version of Tripoli-4, can perform Monte-Carlo parametric studies with a spare factor of Cpu time fetching 50 times. A module of coupling, simulating neutron guides, has also been developed and implemented in the Monte-Carlo code McStas. Thirdly, achieving a complete study for the validation of the installed calculation chain. These studies focus on 3 cold sources currently functioning: SP1 from Orphee reactor and 2 other sources (SFH and SFV) from the HFR at the Laue Langevin Institute. These studies give examples of problems and methods for the design of future cold sources.

  12. Neutron imaging and tomography with MCPS

    NARCIS (Netherlands)

    Duarte Pinto, S.C.; Ortega, R; Ritzau, S.; Pasquale, D; Laprade, B.; Mrotek, S.; Gardell, S.; Zhou, Z.; Plomp, J.; van Eijck, L.; Bilheux, H.; Dhiman, I.

    2017-01-01

    A neutron imaging detector based on neutron-sensitive microchannel plates (mcps) was constructed and tested at beamlines of thermal and cold neutrons. The mcps are made of a glass mixture containing 10B and natural Gd, which makes the bulk of the mcp an efficient neutron converter. Contrary to the

  13. Neutron detectors for the ESS diffractometers

    Czech Academy of Sciences Publication Activity Database

    Stefanescu, I.; Christensen, M.; Fenske, J.; Hall-Wilton, R.; Henry, P. F.; Kirstein, O.; Muller, M.; Nowak, G.; Pooley, D.; Raspino, D.; Rhodes, N.; Šaroun, Jan; Schefer, J.; Schooneveld, E.; Sykora, J.; Schweika, W.

    2017-01-01

    Roč. 12, JAN (2017), č. článku P01019. ISSN 1748-0221 R&D Projects: GA MŠk LM2015048 Institutional support: RVO:61389005 Keywords : instrumentation for neutron sources * neutron diffraction detectors * neutron detectors (cold, thermal, fast neutrons) Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 1.220, year: 2016

  14. Composite boron nitride neutron detectors

    Science.gov (United States)

    Roth, M.; Mojaev, E.; Khakhan, O.; Fleider, A.; Dul`kin, E.; Schieber, M.

    2014-09-01

    Single phase polycrystalline hexagonal boron nitride (BN) or mixed with boron carbide (BxC) embedded in an insulating polymeric matrix acting as a binder and forming a composite material as well as pure submicron size polycrystalline BN has been tested as a thermal neutron converter in a multilayer thermal neutron detector design. Metal sheet electrodes were covered with 20-50 μm thick layers of composite materials and assembled in a multi-layer sandwich configuration. High voltage was applied to the metal electrodes to create an interspacing electric field. The spacing volume could be filled with air, nitrogen or argon. Thermal neutrons were captured in converter layers due to the presence of the 10B isotope. The resulting nuclear reaction produced α-particles and 7Li ions which ionized the gas in the spacing volume. Electron-ion pairs were collected by the field to create an electrical signal proportional to the intensity of the neutron source. The detection efficiency of the multilayer neutron detectors is found to increase with the number of active converter layers. Pixel structures of such neutron detectors necessary for imaging applications and incorporation of internal moderator materials for field measurements of fast neutron flux intensities are discussed as well.

  15. Nuclear Masses and Neutron Stars

    CERN Document Server

    Kreim, Susanne; Lunney, David; Schaffner-Bielich, Jürgen

    2013-01-01

    Precision mass spectrometry of neutron-rich nuclei is of great relevance for astrophysics. Masses of exotic nuclides impose constraints on models for the nuclear interaction and thus affect the description of the equation of state of nuclear matter, which can be extended to describe neutron-star matter. With knowledge of the masses of nuclides near shell closures, one can also derive the neutron-star crustal composition. The Penning-trap mass spectrometer ISOLTRAP at CERN-ISOLDE has recently achieved a breakthrough measuring the mass of 82Zn, which allowed constraining neutron-star crust composition to deeper layers (Wolf et al., PRL 110, 2013). We perform a more detailed study on the sequence of nuclei in the outer crust of neutron stars with input from different nuclear models to illustrate the sensitivity to masses and the robustness of neutron-star models. The dominant role of the N=50 and N=82 closed neutron shells for the crustal composition is confirmed.

  16. Overview of Neutron Science Project

    Energy Technology Data Exchange (ETDEWEB)

    Mukaiyama, Takehiko [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-11-01

    JAERI has launched the Neutron Science Project which aims at bringing scientific and technological innovation for the 21st century in the fields of basic science and nuclear technology using a high power spallation neutron source. The Project is preparing the design for a high intensity pulsed and cw spallation neutron sources for such basic science as neutron structural biology, material science, and for accelerator-driven transmutation of long-lived radio-nuclides which are associated with nuclear power generation. The major facilities to be constructed under the Project are, (1) a super-conducting proton linac with the proton energy of 1.5 GeV and the maximum beam power of 8 MW, (2) a spallation target station with input beam power of 5 MW allowing high intensity pulsed neutron beams for neutron scattering, and (3) research facility complex for accelerator-driven transmutation experiments, neutron physics, material irradiation, isotopes production, spallation produced RI beam experiments for exotic nuclei investigation. (author)

  17. Neutronic calculation of fast reactors by the EUCLID/V1 integrated code

    Science.gov (United States)

    Koltashev, D. A.; Stakhanova, A. A.

    2017-01-01

    This article considers neutronic calculation of a fast-neutron lead-cooled reactor BREST-OD-300 by the EUCLID/V1 integrated code. The main goal of development and application of integrated codes is a nuclear power plant safety justification. EUCLID/V1 is integrated code designed for coupled neutronics, thermomechanical and thermohydraulic fast reactor calculations under normal and abnormal operating conditions. EUCLID/V1 code is being developed in the Nuclear Safety Institute of the Russian Academy of Sciences. The integrated code has a modular structure and consists of three main modules: thermohydraulic module HYDRA-IBRAE/LM/V1, thermomechanical module BERKUT and neutronic module DN3D. In addition, the integrated code includes databases with fuel, coolant and structural materials properties. Neutronic module DN3D provides full-scale simulation of neutronic processes in fast reactors. Heat sources distribution, control rods movement, reactivity level changes and other processes can be simulated. Neutron transport equation in multigroup diffusion approximation is solved. This paper contains some calculations implemented as a part of EUCLID/V1 code validation. A fast-neutron lead-cooled reactor BREST-OD-300 transient simulation (fuel assembly floating, decompression of passive feedback system channel) and cross-validation with MCU-FR code results are presented in this paper. The calculations demonstrate EUCLID/V1 code application for BREST-OD-300 simulating and safety justification.

  18. Iodine neutron capture therapy

    Science.gov (United States)

    Ahmed, Kazi Fariduddin

    A new technique, Iodine Neutron Capture Therapy (INCT) is proposed to treat hyperthyroidism in people. Present thyroid therapies, surgical removal and 131I treatment, result in hypothyroidism and, for 131I, involve protracted treatment times and excessive whole-body radiation doses. The new technique involves using a low energy neutron beam to convert a fraction of the natural iodine stored in the thyroid to radioactive 128I, which has a 24-minute half-life and decays by emitting 2.12-MeV beta particles. The beta particles are absorbed in and damage some thyroid tissue cells and consequently reduce the production and release of thyroid hormones to the blood stream. Treatment times and whole-body radiation doses are thus reduced substantially. This dissertation addresses the first of the several steps needed to obtain medical profession acceptance and regulatory approval to implement this therapy. As with other such programs, initial feasibility is established by performing experiments on suitable small mammals. Laboratory rats were used and their thyroids were exposed to the beta particles coming from small encapsulated amounts of 128I. Masses of 89.0 mg reagent-grade elemental iodine crystals have been activated in the ISU AGN-201 reactor to provide 0.033 mBq of 128I. This activity delivers 0.2 Gy to the thyroid gland of 300-g male rats having fresh thyroid tissue masses of ˜20 mg. Larger iodine masses are used to provide greater doses. The activated iodine is encapsulated to form a thin (0.16 cm 2/mg) patch that is then applied directly to the surgically exposed thyroid of an anesthetized rat. Direct neutron irradiation of a rat's thyroid was not possible due to its small size. Direct in-vivo exposure of the thyroid of the rat to the emitted radiation from 128I is allowed to continue for 2.5 hours (6 half-lives). Pre- and post-exposure blood samples are taken to quantify thyroid hormone levels. The serum T4 concentration is measured by radioimmunoassay at

  19. Calibration and Monte Carlo modelling of neutron long counters

    CERN Document Server

    Tagziria, H

    2000-01-01

    The Monte Carlo technique has become a very powerful tool in radiation transport as full advantage is taken of enhanced cross-section data, more powerful computers and statistical techniques, together with better characterisation of neutron and photon source spectra. At the National Physical Laboratory, calculations using the Monte Carlo radiation transport code MCNP-4B have been combined with accurate measurements to characterise two long counters routinely used to standardise monoenergetic neutron fields. New and more accurate response function curves have been produced for both long counters. A novel approach using Monte Carlo methods has been developed, validated and used to model the response function of the counters and determine more accurately their effective centres, which have always been difficult to establish experimentally. Calculations and measurements agree well, especially for the De Pangher long counter for which details of the design and constructional material are well known. The sensitivit...

  20. Development of the V4.2m5 and V5.0m0 Multigroup Cross Section Libraries for MPACT for PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Seog [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Clarno, Kevin T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gentry, Cole [Univ. of Tennessee, Knoxville, TN (United States); Wiarda, Dorothea [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Williams, Mark L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kochunas, Brendan [Univ. of Michigan, Ann Arbor, MI (United States); Liu, Yuxuan [Univ. of Michigan, Ann Arbor, MI (United States); Palmtag, Scott [Core Physics, Inc., Wilmington, NC (United States); Godfrey, Andrew T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-03-01

    The MPACT neutronics module of the Consortium for Advanced Simulation of Light Water Reactors (CASL) core simulator is a 3-D whole core transport code being developed for the CASL toolset, Virtual Environment for Reactor Analysis (VERA). Key characteristics of the MPACT code include (1) a subgroup method for resonance selfshielding and (2) a whole-core transport solver with a 2-D/1-D synthesis method. The MPACT code requires a cross section library to support all the MPACT core simulation capabilities which would be the most influencing component for simulation accuracy.

  1. Polyethylene-reflected plutonium metal sphere : subcritical neutron and gamma measurements.

    Energy Technology Data Exchange (ETDEWEB)

    Mattingly, John K.

    2009-11-01

    Numerous benchmark measurements have been performed to enable developers of neutron transport models and codes to evaluate the accuracy of their calculations. In particular, for criticality safety applications, the International Criticality Safety Benchmark Experiment Program (ICSBEP) annually publishes a handbook of critical and subcritical benchmarks. Relatively fewer benchmark measurements have been performed to validate photon transport models and codes, and unlike the ICSBEP, there is no program dedicated to the evaluation and publication of photon benchmarks. Even fewer coupled neutron-photon benchmarks have been performed. This report documents a coupled neutron-photon benchmark for plutonium metal reflected by polyethylene. A 4.5-kg sphere of ?-phase, weapons-grade plutonium metal was measured in six reflected configurations: (1) Bare; (2) Reflected by 0.5 inch of high density polyethylene (HDPE); (3) Reflected by 1.0 inch of HDPE; (4) Reflected by 1.5 inches of HDPE; (5) Reflected by 3.0 inches of HDPE; and (6) Reflected by 6.0 inches of HDPE. Neutron and photon emissions from the plutonium sphere were measured using three instruments: (1) A gross neutron counter; (2) A neutron multiplicity counter; and (3) A high-resolution gamma spectrometer. This report documents the experimental conditions and results in detail sufficient to permit developers of radiation transport models and codes to construct models of the experiments and to compare their calculations to the measurements. All of the data acquired during this series of experiments are available upon request.

  2. QPO Constraints on Neutron Stars

    Science.gov (United States)

    Miller, M. Coleman

    2005-01-01

    The kilohertz frequencies of QPOs from accreting neutron star systems imply that they are generated in regions of strong gravity, close to the star. This suggests that observations of the QPOs can be used to constrain the properties of neutron stars themselves, and in particular to inform us about the properties of cold matter beyond nuclear densities. Here we discuss some relatively model-insensitive constraints that emerge from the kilohertz QPOs, as well as recent developments that may hint at phenomena related to unstable circular orbits outside neutron stars.

  3. Neutron Star Physics and EOS

    Directory of Open Access Journals (Sweden)

    Lattimer James M.

    2016-01-01

    Full Text Available Neutron stars are important because measurement of their masses and radii will determine the dense matter equation of state. They will constrain the nuclear matter symmetry energy, which controls the neutron star matter pressure and the interior composition, and will influence the interpretation of nuclear experiments. Astrophysical observations include pulsar timing, X-ray bursts, quiescent low-mass X-ray binaries, pulse profiles from millisecond pulsars, neutrino observations from gravitational collapse supernovae,and gravitational radiation from compact object mergers. These observations will also constrain the neutron star interior, including the properties of superfluidity there, and determine the existence of a possible QCD phase transition.

  4. Neutron star structure from QCD

    CERN Document Server

    Fraga, Eduardo S; Vuorinen, Aleksi

    2016-01-01

    In this review article, we argue that our current understanding of the thermodynamic properties of cold QCD matter, originating from first principles calculations at high and low densities, can be used to efficiently constrain the macroscopic properties of neutron stars. In particular, we demonstrate that combining state-of-the-art results from Chiral Effective Theory and perturbative QCD with the current bounds on neutron star masses, the Equation of State of neutron star matter can be obtained to an accuracy better than 30% at all densities.

  5. POLI: Polarised hot neutron diffractometer

    Directory of Open Access Journals (Sweden)

    Vladimir Hutanu

    2015-08-01

    Full Text Available POLI, which is operated by the Institute of Crystallography, RWTH Aachen University in cooperation with JCNS, Forschungszentrum Jülich, is a versatile two axes single crystal diffractometer with broad field of applications. Mostly dedicated to the investigation of magnetic structures in single crystals using neutron spin polarisation, POLI is also used for classical structural investigations under extreme conditions. High intensity hot neutrons flux makes it attractive also for the other applications like study of parity violations phenomena in nuclear physics or BNCT (boron neutron-capture therapy in medicine.

  6. Neutrons for probing matter; Des neutrons pour explorer la matiere

    Energy Technology Data Exchange (ETDEWEB)

    Torres, F. Ed.; Mazzucchetti, D

    2008-07-01

    The authors tell the story of the French Orphee reactor located in Saclay from the decision to build it in the seventies, to its commissioning in 1980, to its upgrading in the nineties and to its today's operating life. As early as its feasibility studies Orphee has been designed as a dual-purpose reactor: scientific research for instance in crystallography and magnetism, and industrial uses like neutron radiography, silicon doping or radionuclide production. This book is divided into 4 parts: 1) the neutron: an explorer of the matter, 2) the Orphee reactor: a neutron source, 3) the adventurers of the matter: Leon Brillouin laboratory's staff, and 4) the perspectives for neutrons.

  7. INFLUENCE OF SCATTERED NEUTRON RADIATION ON METROLOGICAL CHARACTERISTICS OF АТ140 NEUTRON CALIBRATION FACILITY

    Directory of Open Access Journals (Sweden)

    D. I. Komar

    2017-01-01

    Full Text Available Today facilities with collimated radiation field are widely used as reference in metrological support of devices for neutron radiation measurement. Neutron fields formed by radionuclide neutron sources. The aim of this research was to study characteristics of experimentally realized neutron fields geometries on АТ140 Neutron Calibration Facility using Monte Carlo method.For calibration, we put a device into neutron field with known flux density or ambient equivalent dose rate. We can form neutron beam from radionuclide fast-neutron source in different geometries. In containercollimator of АТ140 Neutron Calibration Facility we can install special inserts to gather fast-neutron geometry or thermal-neutron geometry. We need to consider neutron scattering from air and room’s walls. We can conduct measurements of neutron field characteristics in several points and get the other using Monte Carlo method.Thermal neutron collimator forms a beam from radionuclide source with a significant amount of neutrons with thermal energies. From found relationship between full neutron flux and distance to neutron source we see that inverse square law is violated. Scattered radiation contribution into total flux increases when we are moving away from neutron source and significantly influences neutron fields characteristics. While source is exposed in shadow-cone geometry neutron specter has pronounced thermal component from wall scattering.In this work, we examined main geometry types used to acquire reference neutron radiation using radionuclide sources. We developed Monte Carlo model for 238Pu-Be neutron source and АТ140 Neutron Calibration Facility’s container-collimator. We have shown the most significant neutron energy distribution factor to be scattered radiation from room’s walls. It leads to significant changes of neutron radiation specter at a distance from the source. When planning location, and installing the facility we should consider

  8. Oscillations in neutron stars

    Energy Technology Data Exchange (ETDEWEB)

    Hoeye, Gudrun Kristine

    1999-07-01

    We have studied radial and nonradial oscillations in neutron stars, both in a general relativistic and non-relativistic frame, for several different equilibrium models. Different equations of state were combined, and our results show that it is possible to distinguish between the models based on their oscillation periods. We have particularly focused on the p-, f-, and g-modes. We find oscillation periods of II approx. 0.1 ms for the p-modes, II approx. 0.1 - 0.8 ms for the f-modes and II approx. 10 - 400 ms for the g-modes. For high-order (l (>{sub )} 4) f-modes we were also able to derive a formula that determines II{sub l+1} from II{sub l} and II{sub l-1} to an accuracy of 0.1%. Further, for the radial f-mode we find that the oscillation period goes to infinity as the maximum mass of the star is approached. Both p-, f-, and g-modes are sensitive to changes in the central baryon number density n{sub c}, while the g-modes are also sensitive to variations in the surface temperature. The g-modes are concentrated in the surface layer, while p- and f-modes can be found in all parts of the star. The effects of general relativity were studied, and we find that these are important at high central baryon number densities, especially for the p- and f-modes. General relativistic effects can therefore not be neglected when studying oscillations in neutron stars. We have further developed an improved Cowling approximation in the non-relativistic frame, which eliminates about half of the gap in the oscillation periods that results from use of the ordinary Cowling approximation. We suggest to develop an improved Cowling approximation also in the general relativistic frame. (Author)

  9. Shielding evaluation of neutron generator hall by Monte Carlo simulations

    Energy Technology Data Exchange (ETDEWEB)

    Pujala, U.; Selvakumaran, T.S.; Baskaran, R.; Venkatraman, B. [Radiological Safety Division, Indira Gandhi Center for Atomic Research, Kalpakkam (India); Thilagam, L.; Mohapatra, D.K., E-mail: swathythila2@yahoo.com [Safety Research Institute, Atomic Energy Regulatory Board, Kalpakkam (India)

    2017-04-01

    A shielded hall was constructed for accommodating a D-D, D-T or D-Be based pulsed neutron generator (NG) with 4π yield of 10{sup 9} n/s. The neutron shield design of the facility was optimized using NCRP-51 methodology such that the total dose rates outside the hall areas are well below the regulatory limit for full occupancy criterion (1 μSv/h). However, the total dose rates at roof top, cooling room trench exit and labyrinth exit were found to be above this limit for the optimized design. Hence, additional neutron shielding arrangements were proposed for cooling room trench and labyrinth exits. The roof top was made inaccessible. The present study is an attempt to evaluate the neutron and associated capture gamma transport through the bulk shields for the complete geometry and materials of the NG-Hall using Monte Carlo (MC) codes MCNP and FLUKA. The neutron source terms of D-D, D-T and D-Be reactions are considered in the simulations. The effect of additional shielding proposed has been demonstrated through the simulations carried out with the consideration of the additional shielding for D-Be neutron source term. The results MC simulations using two different codes are found to be consistent with each other for neutron dose rate estimates. However, deviation up to 28% is noted between these two codes at few locations for capture gamma dose rate estimates. Overall, the dose rates estimated by MC simulations including additional shields shows that all the locations surrounding the hall satisfy the full occupancy criteria for all three types of sources. Additionally, the dose rates due to direct transmission of primary neutrons estimated by FLUKA are compared with the values calculated using the formula given in NCRP-51 which shows deviations up to 50% with each other. The details of MC simulations and NCRP-51 methodology for the estimation of primary neutron dose rate along with the results are presented in this paper. (author)

  10. Effective mass of free neutrons in neutron star crust

    Energy Technology Data Exchange (ETDEWEB)

    Chamel, Nicolas [Copernicus Astronomical Center (CAMK), Polish Academy of Sciences, ul. Bartycka 18, 00-716 Warsaw (Poland) and LUTH, Paris Observatory, 5 place Jules Janssen, 92195 Meudon (France)]. E-mail: nchamel@camk.edu.pl

    2006-07-24

    The inner layers of a neutron star crust, composed of a Coulomb lattice of neutron rich nuclear clusters immersed in a sea of 'free' superfluid neutrons, are closely analogous to periodic condensed matter systems such as electronic, photonic or phononic crystals. Applying methods from solid state physics to the neutron star context, it has been recently shown that Bragg scattering leads to a strong renormalization of the neutron mass in the bottom layers of the crust. Knowledge of this effective mass throughout the crust is essential in order to understand the dynamical properties of the neutron superfluid and the origin of pulsar glitches. The purpose of the present work is to evaluate this effective mass in the outermost layers of the inner crust, near the drip point {rho}{sub drip}{approx}4x10{sup 11} g-bar cm{sup -3}. The results are compared with the case of electrons in ordinary solids and as an example the corresponding effective electron mass in copper is calculated.

  11. Gamma/neutron competition above the neutron separation energy in delayed neutron emitters

    Directory of Open Access Journals (Sweden)

    Valencia E.

    2014-03-01

    Full Text Available To study the β-decay properties of some well known delayed neutron emitters an experiment was performed in 2009 at the IGISOL facility (University of Jyväskylä in Finland using Total Absorption γ-ray Spectroscopy (TAGS technique. The aim of these measurements is to obtain the full β-strength distribution below the neutron separation energy (Sn and the γ/neutron competition above. This information is a key parameter in nuclear technology applications as well as in nuclear astrophysics and nuclear structure. Preliminary results of the analysis show a significant γ-branching ratio above Sn.

  12. Magnetic waveguides for neutron reflectometry

    Science.gov (United States)

    Khaydukov, Yu.; Petrzhik, A. M.; Borisenko, I. V.; Kalabukhov, A.; Winkler, D.; Keller, T.; Ovsyannikov, G. A.; Keimer, B.

    2017-10-01

    We show that the sensitivity and depth selectivity of neutron reflectometry can be greatly enhanced through a waveguide design that takes advantage of the spin-dependent magnetic neutron scattering potential to steer spin-up and spin-down neutrons into waveguide modes with different depth profiles. Using a bilayer of manganate and ruthenate ferromagnets, we demonstrate that a magnetic waveguide structure with sharp spin-up and spin-down modes centered in the two different layers can be generated by adding a magnetically inactive capping layer. The resulting reflectometric data allow accurate and reliable determination of a small in-plane magnetization in the ruthenate layer, despite its immediate proximity to the manganate layer with much larger magnetization. Magnetic neutron waveguides thus enable depth-sensitive measurements of small electronic spin polarizations in a large variety of magnetic multilayers and devices.

  13. Helium Detonations on Neutron Stars

    Science.gov (United States)

    Zingale, M.; Timmes, F. X.; Fryxell, B.; Lamb, D. Q.; Olson, K.; Calder, A. C.; Dursi, L. J.; Ricker, P.; Rosner, R.; Truran, J. W.; MacNeice, P.; Tufo, H.

    2000-05-01

    We present the results of a numerical study of helium detonations on the surfaces of neutron stars. We analyze the evolution of a detonation as it breaks through the envelope of the neutron star and propagates across its surface. A series of surface waves propagate across the pool of hot ash with a speed of 1.3 x 109 \\ cm \\ s-1, matching the speed expected from shallow water wave theory. The entire envelope bounces in the gravitational potential well of the neutron star with a period of 50 μ s. The photosphere reaches a height of 15 km above the surface of the neutron star. The sensitivity of the results to the spatial resolution and assumed initial conditions are analyzed, and the relevance of this model to Type I X-ray bursts is discussed. This work is supported by the Department of Energy under Grant No. B341495 to the Center for Astrophysical Thermonuclear Flashes at the University of Chicago.

  14. Theory of neutron star magnetospheres

    CERN Document Server

    Curtis Michel, F

    1990-01-01

    An incomparable reference for astrophysicists studying pulsars and other kinds of neutron stars, "Theory of Neutron Star Magnetospheres" sums up two decades of astrophysical research. It provides in one volume the most important findings to date on this topic, essential to astrophysicists faced with a huge and widely scattered literature. F. Curtis Michel, who was among the first theorists to propose a neutron star model for radio pulsars, analyzes competing models of pulsars, radio emission models, winds and jets from pulsars, pulsating X-ray sources, gamma-ray burst sources, and other neutron-star driven phenomena. Although the book places primary emphasis on theoretical essentials, it also provides a considerable introduction to the observational data and its organization. Michel emphasizes the problems and uncertainties that have arisen in the research as well as the considerable progress that has been made to date.

  15. Neutron radiography in field use

    Energy Technology Data Exchange (ETDEWEB)

    Warman, E. A.

    1962-10-15

    Some of the basic aspects of neutron radiography are reviewed with particular emphasis on the physics of this nondestructive testing method. The results of an experimental evaluation program are presented which show conclusively that isotopic antimony-beryllium photoneutron sources can be used for neutron radiographic applications. This program provided valuable new information regarding neution moderation and beam collimation requirements. Based on this experimental evidence, a portable device was designed and is described. The device employs a portable radiographic camera, with im antimony- beryllium photoneutron source, in conjunction with a neutron moderator and beam collimator configuration. This development makes available, for the first time, a means of obtaining economical inspection neutron radiographs in field application usage. (auth)

  16. Neutron protein crystallography in JAERI

    Indian Academy of Sciences (India)

    Author Affiliations. I Tanaka1 2. Neutron Science Research Center, Japan Atomic Energy Research Institute (JAERI), Tokai, Ibaraki 319-1195, Japan; Faculty of Engineering, Ibaraki University, Hitachi, Ibaraki 316-8511, Japan ...

  17. Prototype Neutron Portal Monitor Detector

    Science.gov (United States)

    Schier, W.

    2014-05-01

    A very large drum-shaped neutron detector which could replace the 3He neutron portal monitor detector is under development. Detection is based on the 6Li(n,3H)4He reaction. 6Li metal is evaporated onto aluminum plates then covered with 22-cm x 27-cm ZnS(Ag) scintillation sheets and sealed about the edges. The equivalent of 40 detector plates will be arrayed in the 80-cm diameter drum housing and viewed by a single 20-cm diameter hemispherical photomultiplier tube without the use of light guides. Presently 25 detector plates are installed. Light collection tests are performed with a bare 210Po alpha source on a ZnS(Ag) disk. Neutron detection studies include neutrons from a 2-curie PuBe source and from a 0.255-gram 240Pu source.

  18. Materials for spallation neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    Sommer, W.F.; Daemen, L.L. [comps.

    1996-03-01

    The Workshop on Materials for Spallation Neutron Sources at the Los Alamos Neutron Science Center, February 6 to 10, 1995, gathered scientists from Department of Energy national laboratories, other federal institutions, universities, and industry to discuss areas in which work is needed, successful designs and use of materials, and opportunities for further studies. During the first day of the workshop, speakers presented overviews of current spallation neutron sources. During the next 3 days, seven panels allowed speakers to present information on a variety of topics ranging from experimental and theoretical considerations on radiation damage to materials safety issues. An attempt was made to identify specific problems that require attention within the context of spallation neutron sources. This proceedings is a collection of summaries from the overview sessions and the panel presentations.

  19. The Spallation Neutron Source accelerator system design

    Energy Technology Data Exchange (ETDEWEB)

    Henderson, S., E-mail: stuarth@fnal.gov [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Abraham, W. [Lawrence Berkeley National Laboratory, 1 Cyclotron Road, Berkeley, CA 94720 (United States); Aleksandrov, A. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Allen, C. [Techsource, Inc., 1475 Central Avenue, Suite 250, Los Alamos, NM 87544-3291 (United States); Alonso, J. [Lawrence Berkeley National Laboratory, 1 Cyclotron Road, Berkeley, CA 94720 (United States); Anderson, D. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Arenius, D. [Thomas Jefferson National Accelerator Facility, 12000 Jefferson Avenue, Newport News, VA 23606 (United States); Arthur, T. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Assadi, S. [Techsource, Inc., 1475 Central Avenue, Suite 250, Los Alamos, NM 87544-3291 (United States); Ayers, J.; Bach, P. [Lawrence Berkeley National Laboratory, 1 Cyclotron Road, Berkeley, CA 94720 (United States); Badea, V. [Brookhaven National Laboratory, P.O. Box 5000, Upton, NY 11973-5000 (United States); Battle, R. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Beebe-Wang, J. [Brookhaven National Laboratory, P.O. Box 5000, Upton, NY 11973-5000 (United States); Bergmann, B.; Bernardin, J.; Bhatia, T.; Billen, J.; Birke, T.; Bjorklund, E. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); and others

    2014-11-01

    The Spallation Neutron Source (SNS) was designed and constructed by a collaboration of six U.S. Department of Energy national laboratories. The SNS accelerator system consists of a 1 GeV linear accelerator and an accumulator ring providing 1.4 MW of proton beam power in microsecond-long beam pulses to a liquid mercury target for neutron production. The accelerator complex consists of a front-end negative hydrogen-ion injector system, an 87 MeV drift tube linear accelerator, a 186 MeV side-coupled linear accelerator, a 1 GeV superconducting linear accelerator, a 248-m circumference accumulator ring and associated beam transport lines. The accelerator complex is supported by ∼100 high-power RF power systems, a 2 K cryogenic plant, ∼400 DC and pulsed power supply systems, ∼400 beam diagnostic devices and a distributed control system handling ∼100,000 I/O signals. The beam dynamics design of the SNS accelerator is presented, as is the engineering design of the major accelerator subsystems.

  20. The Spallation Neutron Source accelerator system design

    Science.gov (United States)

    Henderson, S.; Abraham, W.; Aleksandrov, A.; Allen, C.; Alonso, J.; Anderson, D.; Arenius, D.; Arthur, T.; Assadi, S.; Ayers, J.; Bach, P.; Badea, V.; Battle, R.; Beebe-Wang, J.; Bergmann, B.; Bernardin, J.; Bhatia, T.; Billen, J.; Birke, T.; Bjorklund, E.; Blaskiewicz, M.; Blind, B.; Blokland, W.; Bookwalter, V.; Borovina, D.; Bowling, S.; Bradley, J.; Brantley, C.; Brennan, J.; Brodowski, J.; Brown, S.; Brown, R.; Bruce, D.; Bultman, N.; Cameron, P.; Campisi, I.; Casagrande, F.; Catalan-Lasheras, N.; Champion, M.; Champion, M.; Chen, Z.; Cheng, D.; Cho, Y.; Christensen, K.; Chu, C.; Cleaves, J.; Connolly, R.; Cote, T.; Cousineau, S.; Crandall, K.; Creel, J.; Crofford, M.; Cull, P.; Cutler, R.; Dabney, R.; Dalesio, L.; Daly, E.; Damm, R.; Danilov, V.; Davino, D.; Davis, K.; Dawson, C.; Day, L.; Deibele, C.; Delayen, J.; DeLong, J.; Demello, A.; DeVan, W.; Digennaro, R.; Dixon, K.; Dodson, G.; Doleans, M.; Doolittle, L.; Doss, J.; Drury, M.; Elliot, T.; Ellis, S.; Error, J.; Fazekas, J.; Fedotov, A.; Feng, P.; Fischer, J.; Fox, W.; Fuja, R.; Funk, W.; Galambos, J.; Ganni, V.; Garnett, R.; Geng, X.; Gentzlinger, R.; Giannella, M.; Gibson, P.; Gillis, R.; Gioia, J.; Gordon, J.; Gough, R.; Greer, J.; Gregory, W.; Gribble, R.; Grice, W.; Gurd, D.; Gurd, P.; Guthrie, A.; Hahn, H.; Hardek, T.; Hardekopf, R.; Harrison, J.; Hatfield, D.; He, P.; Hechler, M.; Heistermann, F.; Helus, S.; Hiatt, T.; Hicks, S.; Hill, J.; Hill, J.; Hoff, L.; Hoff, M.; Hogan, J.; Holding, M.; Holik, P.; Holmes, J.; Holtkamp, N.; Hovater, C.; Howell, M.; Hseuh, H.; Huhn, A.; Hunter, T.; Ilg, T.; Jackson, J.; Jain, A.; Jason, A.; Jeon, D.; Johnson, G.; Jones, A.; Joseph, S.; Justice, A.; Kang, Y.; Kasemir, K.; Keller, R.; Kersevan, R.; Kerstiens, D.; Kesselman, M.; Kim, S.; Kneisel, P.; Kravchuk, L.; Kuneli, T.; Kurennoy, S.; Kustom, R.; Kwon, S.; Ladd, P.; Lambiase, R.; Lee, Y. Y.; Leitner, M.; Leung, K.-N.; Lewis, S.; Liaw, C.; Lionberger, C.; Lo, C. C.; Long, C.; Ludewig, H.; Ludvig, J.; Luft, P.; Lynch, M.; Ma, H.; MacGill, R.; Macha, K.; Madre, B.; Mahler, G.; Mahoney, K.; Maines, J.; Mammosser, J.; Mann, T.; Marneris, I.; Marroquin, P.; Martineau, R.; Matsumoto, K.; McCarthy, M.; McChesney, C.; McGahern, W.; McGehee, P.; Meng, W.; Merz, B.; Meyer, R.; Meyer, R.; Miller, B.; Mitchell, R.; Mize, J.; Monroy, M.; Munro, J.; Murdoch, G.; Musson, J.; Nath, S.; Nelson, R.; Nelson, R.; O`Hara, J.; Olsen, D.; Oren, W.; Oshatz, D.; Owens, T.; Pai, C.; Papaphilippou, I.; Patterson, N.; Patterson, J.; Pearson, C.; Pelaia, T.; Pieck, M.; Piller, C.; Plawski, T.; Plum, M.; Pogge, J.; Power, J.; Powers, T.; Preble, J.; Prokop, M.; Pruyn, J.; Purcell, D.; Rank, J.; Raparia, D.; Ratti, A.; Reass, W.; Reece, K.; Rees, D.; Regan, A.; Regis, M.; Reijonen, J.; Rej, D.; Richards, D.; Richied, D.; Rode, C.; Rodriguez, W.; Rodriguez, M.; Rohlev, A.; Rose, C.; Roseberry, T.; Rowton, L.; Roybal, W.; Rust, K.; Salazer, G.; Sandberg, J.; Saunders, J.; Schenkel, T.; Schneider, W.; Schrage, D.; Schubert, J.; Severino, F.; Shafer, R.; Shea, T.; Shishlo, A.; Shoaee, H.; Sibley, C.; Sims, J.; Smee, S.; Smith, J.; Smith, K.; Spitz, R.; Staples, J.; Stein, P.; Stettler, M.; Stirbet, M.; Stockli, M.; Stone, W.; Stout, D.; Stovall, J.; Strelo, W.; Strong, H.; Sundelin, R.; Syversrud, D.; Szajbler, M.; Takeda, H.; Tallerico, P.; Tang, J.; Tanke, E.; Tepikian, S.; Thomae, R.; Thompson, D.; Thomson, D.; Thuot, M.; Treml, C.; Tsoupas, N.; Tuozzolo, J.; Tuzel, W.; Vassioutchenko, A.; Virostek, S.; Wallig, J.; Wanderer, P.; Wang, Y.; Wang, J. G.; Wangler, T.; Warren, D.; Wei, J.; Weiss, D.; Welton, R.; Weng, J.; Weng, W.-T.; Wezensky, M.; White, M.; Whitlatch, T.; Williams, D.; Williams, E.; Wilson, K.; Wiseman, M.; Wood, R.; Wright, P.; Wu, A.; Ybarrolaza, N.; Young, K.; Young, L.; Yourd, R.; Zachoszcz, A.; Zaltsman, A.; Zhang, S.; Zhang, W.; Zhang, Y.; Zhukov, A.

    2014-11-01

    The Spallation Neutron Source (SNS) was designed and constructed by a collaboration of six U.S. Department of Energy national laboratories. The SNS accelerator system consists of a 1 GeV linear accelerator and an accumulator ring providing 1.4 MW of proton beam power in microsecond-long beam pulses to a liquid mercury target for neutron production. The accelerator complex consists of a front-end negative hydrogen-ion injector system, an 87 MeV drift tube linear accelerator, a 186 MeV side-coupled linear accelerator, a 1 GeV superconducting linear accelerator, a 248-m circumference accumulator ring and associated beam transport lines. The accelerator complex is supported by ~100 high-power RF power systems, a 2 K cryogenic plant, ~400 DC and pulsed power supply systems, ~400 beam diagnostic devices and a distributed control system handling ~100,000 I/O signals. The beam dynamics design of the SNS accelerator is presented, as is the engineering design of the major accelerator subsystems.

  1. Digital fast neutron radiography of steel reinforcing bar in concrete

    Science.gov (United States)

    Mitton, K.; Jones, A.; Joyce, M. J.

    2014-12-01

    Neutron imaging has previously been used in order to test for cracks, degradation and water content in concrete. However, these techniques often fall short of alternative non-destructive testing methods, such as γ-ray and X-ray imaging, particularly in terms of resolution. Further, thermal neutron techniques can be compromised by the significant expense associated with thermal neutron sources of sufficient intensity to yield satisfactory results that can often precipitate the need for a reactor. Such embodiments are clearly not portable in the context of the needs of field applications. This paper summarises the results of a study to investigate the potential for transmission radiography based on fast neutrons. The objective of this study was to determine whether the presence of heterogeneities in concrete, such as reinforcement structures, could be identified on the basis of variation in transmitted fast-neutron flux. Monte-Carlo simulations have been performed and the results from these are compared to those arising from practical tests using a 252Cf source. The experimental data have been acquired using a digital pulse-shape discrimination system that enables fast neutron transmission to be studied across an array of liquid scintillators placed in close proximity to samples under test, and read out in real time. Whilst this study does not yield sufficient spatial resolution, a comparison of overall flux ratios does provide a basis for the discrimination between samples with contrasting rebar content. This approach offers the potential for non-destructive testing that gives less dose, better transportability and better accessibility than competing approaches. It is also suitable for thick samples where γ-ray and X-ray methods can be limited.

  2. Outline of spallation neutron source engineering

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Noboru [Center for Neutron Science, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan)

    2001-01-01

    Slow neutrons such as cold and thermal neutrons are unique probes which can determine structures and dynamics of condensed matter in atomic scale. The neutron scattering technique is indispensable not only for basic sciences such as condensed matter research and life science, but also for basic industrial technology in 21 century. It is believed that to survive in the science-technology competition in 21 century would be almost impossible without neutron scattering. However, the intensity of neutrons presently available is much lower than synchrotron radiation sources, etc. Thus, R and D of intense neutron sources become most important. The High-Intensity Proton Accelerator Project is now being promoted jointly by Japan Atomic Energy Research Institute and High Energy Accelerator Research Organization, but there has so far been no good text which covers all the aspects of pulsed spallation neutron sources. The present review was prepare aiming at giving a better understanding on pulsed spallation neutron sources not only to neutron source researchers but also more widely to neutron scattering researchers and accelerator scientists in this field. The contents involve, starting from what is neutron scattering and what neutrons are necessary for neutron scattering, what is the spallation reaction, how to produce neutrons required for neutron scattering more efficiently, target-moderator-reflector neutronics and its engineering, shielding, target station, material issues, etc. The author have engaged in R and D of pulsed apallation neutron sources and neutron scattering research using them over 30 years. The present review is prepared based on the author's experiences with useful information obtained through ICANS collaboration and recent data from the JSNS (Japanese Spallation Neutron Source) design team. (author)

  3. Systematics in delayed neutron yields

    Energy Technology Data Exchange (ETDEWEB)

    Ohsawa, Takaaki [Kinki Univ., Higashi-Osaka, Osaka (Japan). Atomic Energy Research Inst.

    1998-03-01

    An attempt was made to reproduce the systematic trend observed in the delayed neutron yields for actinides on the basis of the five-Gaussian representation of the fission yield together with available data sets for delayed neutron emission probability. It was found that systematic decrease in DNY for heavier actinides is mainly due to decrease of fission yields of precursors in the lighter side of the light fragment region. (author)

  4. Bunching of continuous neutron beams

    Energy Technology Data Exchange (ETDEWEB)

    Golub, R. [Hahn Meitner Institut, Glienickerstr. 100, 14109 Berlin (Germany); Gaehler, R. [Institut Laue Langevin, 6 Rue Jules Horowitz, 38042 Grenoble (France)]. E-mail: gahler@ill.fr; Habicht, K. [Hahn Meitner Institut, Glienickerstr. 100, 14109 Berlin (Germany); Klimko, S. [Institut Laue Langevin, 6 Rue Jules Horowitz, 38042 Grenoble (France)

    2006-01-09

    We present a system which can shape the intensity of a continuous, polychromatic polarized neutron beam into a high frequency sinusoidal signal of high contrast. It is based on a pure quantum-mechanical effect (longitudinal Stern-Gerlach). The system relies on a sequence of three RF flippers, and it works virtually loss-free. The signal occurs far downstream of the last flipper within a well defined region. As an application we discuss its use in high resolution neutron spectroscopy.

  5. CONTROL MEANS FOR NEUTRONIC REACTORS

    Science.gov (United States)

    Tonks, L.

    1962-08-01

    A control device surrounding the active portion of a nuclear reactor is described. The control device consists of a plurality of contiguous cylinders partly filled with a neutron absorbing material and partly filled with a neutron reflecting material, each cylinder having a longitudinal reentrant surface into which a portion of an adjacent cylinder extends, one of the cylinders having two re-entrant surfaces, and means for rotating the cylinders one at a time. (AEC)

  6. Alternative Neutron Detection Testing Summary

    Energy Technology Data Exchange (ETDEWEB)

    Kouzes, Richard T.; Ely, James H.; Erikson, Luke E.; Kernan, Warnick J.; Lintereur, Azaree T.; Siciliano, Edward R.; Stromswold, David C.; Woodring, Mitchell L.

    2010-04-08

    Radiation portal monitors used for interdiction of illicit materials at borders include highly sensitive neutron detection systems. The main reason for having neutron detection capability is to detect fission neutrons from plutonium. Most currently deployed radiation portal monitors (RPMs) use neutron detectors based upon 3He-filled gas proportional counters, which are the most common large area neutron detector. This type of neutron detector is used in the TSA and other RPMs installed in international locations and in the Ludlum and Science Applications International Corporation RPMs deployed primarily for domestic applications. There is a declining supply of 3He in the world and, thus, methods to reduce the use of this gas in RPMs with minimal changes to the current system designs and sensitivity to cargo-borne neutrons are being investigated. Four technologies have been identified as being currently commercially available, potential alternative neutron detectors to replace the use of 3He in RPMs. These technologies are: 1) Boron trifluoride-filled proportional counters, 2) Boron-lined proportional counters, 3) Lithium-loaded glass fibers, and 4) Coated wavelength-shifting plastic fibers. Reported here is a summary of the testing carried out at Pacific Northwest National Laboratory on these technologies to date, as well as measurements on 3He tubes at various pressures. Details on these measurements are available in the referenced reports. Sponsors of these tests include the Department of Energy (DOE), Department of Homeland Security (DHS), and the Department of Defense (DoD), as well as internal Pacific Northwest National Laboratory funds.

  7. Sensitivity Analysis of Cf-252 (sf) Neutron and Gamma Observables in CGMF

    Energy Technology Data Exchange (ETDEWEB)

    Carter, Austin Lewis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Talou, Patrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stetcu, Ionel [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kiedrowski, Brian Christopher [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Jaffke, Patrick John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-12-06

    CGMF is a Monte Carlo code that simulates the decay of primary fission fragments by emission of neutrons and gamma rays, according to the Hauser-Feshbach equations. As the CGMF code was recently integrated into the MCNP6.2 transport code, great emphasis has been placed on providing optimal parameters to CGMF such that many different observables are accurately represented. Of these observables, the prompt neutron spectrum, prompt neutron multiplicity, prompt gamma spectrum, and prompt gamma multiplicity are crucial for accurate transport simulations of criticality and nonproliferation applications. This contribution to the ongoing efforts to improve CGMF presents a study of the sensitivity of various neutron and gamma observables to several input parameters for Californium-252 spontaneous fission. Among the most influential parameters are those that affect the input yield distributions in fragment mass and total kinetic energy (TKE). A new scheme for representing Y(A,TKE) was implemented in CGMF using three fission modes, S1, S2 and SL. The sensitivity profiles were calculated for 17 total parameters, which show that the neutron multiplicity distribution is strongly affected by the TKE distribution of the fragments. The total excitation energy (TXE) of the fragments is shared according to a parameter RT, which is defined as the ratio of the light to heavy initial temperatures. The sensitivity profile of the neutron multiplicity shows a second order effect of RT on the mean neutron multiplicity. A final sensitivity profile was produced for the parameter alpha, which affects the spin of the fragments. Higher values of alpha lead to higher fragment spins, which inhibit the emission of neutrons. Understanding the sensitivity of the prompt neutron and gamma observables to the many CGMF input parameters provides a platform for the optimization of these parameters.

  8. High efficiency focus neutron generator

    Science.gov (United States)

    Sadeghi, H.; Amrollahi, R.; Zare, M.; Fazelpour, S.

    2017-12-01

    In the present paper, the new idea to increase the neutron yield of plasma focus devices is investigated and the results are presented. Based on many studies, more than 90% of neutrons in plasma focus devices were produced by beam target interactions and only 10% of them were due to thermonuclear reactions. While propounding the new idea, the number of collisions between deuteron ions and deuterium gas atoms were increased remarkably well. The COMSOL Multiphysics 5.2 was used to study the given idea in the known 28 plasma focus devices. In this circumstance, the neutron yield of this system was also obtained and reported. Finally, it was found that in the ENEA device with 1 Hz working frequency, 1.1 × 109 and 1.1 × 1011 neutrons per second were produced by D–D and D–T reactions, respectively. In addition, in the NX2 device with 16 Hz working frequency, 1.34 × 1010 and 1.34 × 1012 neutrons per second were produced by D–D and D–T reactions, respectively. The results show that with regards to the sizes and energy of these devices, they can be used as the efficient neutron generators.

  9. Neutron spectrum for neutron capture therapy in boron; Espectro de neutrones para terapia por captura de neutrones en boro

    Energy Technology Data Exchange (ETDEWEB)

    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); Baltazar R, A. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica, Programa de Doctorado en Ingenieria y Tecnologia Aplicada, 98068 Zacatecas, Zac. (Mexico); Vega C, H. R., E-mail: dmedina_c@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico)

    2016-10-15

    Glioblastoma multiforme is the most common and aggressive of brain tumors and is difficult to treat by surgery, chemotherapy or conventional radiation therapy. One treatment alternative is the Neutron Capture Therapy in Boron, which requires a beam modulated in neutron energy and a drug with {sup 10}B able to be fixed in the tumor. When the patients head is exposed to the neutron beam, they are captured by the {sup 10}B and produce a nucleus of {sup 7}Li and an alpha particle whose energy is deposited in the cancer cells causing it to be destroyed without damaging the normal tissue. One of the problems associated with this therapy is to have an epithermal neutrons flux of the order of 10{sup 9} n/cm{sup 2}-sec, whereby irradiation channels of a nuclear research reactor are used. In this work using Monte Carlo methods, the neutron spectra obtained in the radial irradiation channel of the TRIGA Mark III reactor are calculated when inserting filters whose position and thickness have been modified. From the arrangements studied, we found that the Fe-Cd-Al-Cd polyethylene filter yielded a ratio between thermal and epithermal neutron fluxes of 0.006 that exceeded the recommended value (<0.05), and the dose due to the capture gamma rays is lower than the dose obtained with the other arrangements studied. (Author)

  10. The Neutron Star Zoo

    Science.gov (United States)

    Harding, Alice K.

    2014-01-01

    Neutron stars are a very diverse population, both in their observational and their physical properties. They prefer to radiate most of their energy at X-ray and gamma-ray wavelengths. But whether their emission is powered by rotation, accretion, heat, magnetic fields or nuclear reactions, they are all different species of the same animal whose magnetic field evolution and interior composition remain a mystery. This article will broadly review the properties of inhabitants of the neutron star zoo, with emphasis on their high-energy emission. XXX Neutron stars are found in a wide variety of sources, displaying an amazing array of behavior. They can be isolated or in binary systems, accreting, heating, cooling, spinning down, spinning up, pulsing, flaring and bursting. The one property that seems to determine their behavior most strongly is their magnetic field strength, structure and evolution. The hot polar caps, bursts and flares of magnetars are likely due to the rapid decay and twisting of their superstrong magnetic fields, whose very existence requires some kind of early dynamo activity. The intermediate-strength magnetic fields of RPPs determines their spin-down behavior and radiation properties. However, the overlap of the magnetar and RPP populations is not understood at present. Why don't high-field RPPs burst or flare? Why don't lower-field magnetars sometimes behave more like RPPs? INS may be old magnetars whose high fields have decayed, but they do not account for the existence of younger RPPs with magnetar-strength fields. Not only the strength of the magnetic field but also its configuration may be important in making a NS a magnetar or a RPP. Magnetic field decay is a critical link between other NS populations as well. "Decay" of the magnetic field is necessary for normal RPPs to evolve into MSPs through accretion and spin up in LMXBs. Some kind of accretion-driven field reduction is the most likely mechanism, but it is controversial since it is not

  11. Compilation of Existing Neutron Screen Technology

    Directory of Open Access Journals (Sweden)

    N. Chrysanthopoulou

    2014-01-01

    Full Text Available The presence of fast neutron spectra in new reactors is expected to induce a strong impact on the contained materials, including structural materials, nuclear fuels, neutron reflecting materials, and tritium breeding materials. Therefore, introduction of these reactors into operation will require extensive testing of their components, which must be performed under neutronic conditions representative of those expected to prevail inside the reactor cores when in operation. Due to limited availability of fast reactors, testing of future reactor materials will mostly take place in water cooled material test reactors (MTRs by tailoring the neutron spectrum via neutron screens. The latter rely on the utilization of materials capable of absorbing neutrons at specific energy. A large but fragmented experience is available on that topic. In this work a comprehensive compilation of the existing neutron screen technology is attempted, focusing on neutron screens developed in order to locally enhance the fast over thermal neutron flux ratio in a reactor core.

  12. The crystal acceleration effect for cold neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Braginetz, Yu. P., E-mail: aiver@pnpi.spb.ru [Petersburg Nuclear Physics Institute NRC KI (Russian Federation); Berdnikov, Ya. A. [Peter the Great St. Petersburg Polytechnic University (Russian Federation); Fedorov, V. V., E-mail: vfedorov@pnpi.spb.ru; Kuznetsov, I. A.; Lasitsa, M. V.; Semenikhin, S. Yu., E-mail: ssy@pnpi.spb.ru; Vezhlev, E. O.; Voronin, V. V., E-mail: vvv@pnpi.spb.ru [Petersburg Nuclear Physics Institute NRC KI (Russian Federation)

    2017-01-15

    A new mechanism of neutron acceleration is discussed and studied experimentally in detail for cold neutrons passing through the accelerated perfect crystal with the energies close to the Bragg one. The effect arises due to the following reason. The crystal refraction index (neutron-crystal interaction potential) for neutron in the vicinity of the Bragg resonance sharply depends on the parameter of deviation from the exact Bragg condition, i.e. on the crystal-neutron relative velocity. Therefore the neutrons enter into accelerated crystal with one neutron-crystal interaction potential and exit with the other. Neutron kinetic energy cannot vary inside the crystal due to its homogeneity. So after passage through such a crystal neutrons will be accelerated or decelerated because of the different energy change at the entrance and exit crystal boundaries.

  13. ATW neutronics design studies.

    Energy Technology Data Exchange (ETDEWEB)

    Wade, D. C.; Yang, W. S.; Khalil, H.

    2000-11-10

    The Accelerator Transmutation of Waste (ATW) concept has been proposed as a transuranics (TRU) (and long-lived fission product) incinerator for processing the 87,000 metric tonnes of Light Water Reactor used fuel which will have been generated by the time the currently deployed fleet of commercial reactors in the US reach the end of their licensed lifetime. The ATW is proposed to separate the uranium from the transuranics and fission products in the LWR used fuel, to fission the transuranics, to send the LWR and ATW generated fission products to the geologic repository and to send the uranium to either a low level waste disposal site or to save it for future use. The heat liberated in fissioning the transuranics would be converted to electricity and sold to partially offset the cost of ATW construction and operations. Options for incineration of long-lived fission products are under evaluation. A six-year science-based program of ATW trade and system studies was initiated in the US FY 2000 to achieve two main purposes: (1) ''to evaluate ATW within the framework of nonproliferation, waste management, and economic considerations,'' and (2) ''to evaluate the efficacy of the numerous technical options for ATW system configuration.'' This paper summarizes the results from neutronics and thermal/hydraulics trade studies which were completed at Argonne National Laboratory during the first year of the program. Core designs were developed for Pb-Bi cooled and Na cooled 840 MW{sub th} fast spectrum transmuter designs employing recycle. Additionally, neutronics analyses were performed at Argonne for a He cooled 600 MW{sub th} hybrid thermal and fast core design proposed by General Atomics Co. which runs critical for 3/4 and subcritical for 1/4 of its four year once-thin burn cycle. The mass flows and the ultimate loss of transuranic isotopes to the waste stream per unit of heat generated during transmutation have been calculated on a

  14. Optimization aspects of the new nELBE photo-neutron source

    Directory of Open Access Journals (Sweden)

    Schwengner R.

    2010-10-01

    Full Text Available The nELBE beamline at Forschungszentrum Dresden-Rossendorf (FZD provides intense neutron beams by stopping primary electrons in a liquid lead target, where neutrons are produced by bremsstrahlung photons via (γ,n reactions. With the aim to increase the neutron yield through the enhancement of the electron beam energy (from the current 40 MeV limit up to 50 MeV, as well as to minimize several sources of background that are presently affecting the measurements, a new neutron beam-line and a new, larger neutron experimental room have been designed. The optimization of the neutron/photon ratio, the minimization of the backscattered radiation from the walls and the possibility to have better experimental conditions are the main advantages of the new design. To optimize the beamline, extensive simulations with the particle interaction and transport code FLUKA have been performed. Starting from the primary electron beam, both the photon and neutron radiation fields have been fully characterized. To have a cross-check of the results, the calculated values of the neutron yields at different energies of the primary beam have been compared both with an independent simulation with the MCNP code and with analytical calculations, obtaining a very satisfactory agreement at the level of few percent. The evaluated radiation fields have been used to optimize the direction of the new neutron beamline, in order to minimize the photon flash contribution. A general overview of the new photo-neutron source, together with all the steps of the optimization study, is here presented and discussed.

  15. Geant4 Analysis of a Thermal Neutron Real-Time Imaging System

    Science.gov (United States)

    Datta, Arka; Hawari, Ayman I.

    2017-07-01

    Thermal neutron imaging is a technique for nondestructive testing providing complementary information to X-ray imaging for a wide range of applications in science and engineering. Advancement of electronic imaging systems makes it possible to obtain neutron radiographs in real time. This method requires a scintillator to convert neutrons to optical photons and a charge-coupled device (CCD) camera to detect those photons. Alongside, a well collimated beam which reduces geometrical blurriness, the use of a thin scintillator can improve the spatial resolution significantly. A representative scintillator that has been applied widely for thermal neutron imaging is 6LiF:ZnS (Ag). In this paper, a multiphysics simulation approach for designing thermal neutron imaging system is investigated. The Geant4 code is used to investigate the performance of a thermal neutron imaging system starting with a neutron source and including the production of charged particles and optical photons in the scintillator and their transport for image formation in the detector. The simulation geometry includes the neutron beam collimator and sapphire filter. The 6LiF:ZnS (Ag) scintillator is modeled along with a pixelated detector for image recording. The spatial resolution of the system was obtained as the thickness of the scintillator screen was varied between 50 and 400 μm. The results of the simulation were compared to experimental results, including measurements performed using the PULSTAR nuclear reactor imaging beam, showing good agreement. Using the established model, further examination showed that the resolution contribution of the scintillator screen is correlated with its thickness and the range of the neutron absorption reaction products (i.e., the alpha and triton particles). Consequently, thinner screens exhibit improved spatial resolution. However, this will compromise detection efficiency due to the reduced probability of neutron absorption.

  16. Neutron and X-ray Detectors

    Energy Technology Data Exchange (ETDEWEB)

    Carini, Gabriella [SLAC National Accelerator Lab., Menlo Park, CA (United States); Denes, Peter [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Gruener, Sol [Cornell Univ., Ithaca, NY (United States); Lessner, Elianne [Dept. of Energy (DOE), Washington DC (United States). Office of Science Office of Basic Energy Sciences

    2012-08-01

    : Improvements in the readout speed and energy resolution of X-ray detectors are essential to enable chemically sensitive microscopies. Advances would make it possible to take images with simultaneous spatial and chemical information. Very high-energy-resolution X-ray detectors: The energy resolution of semiconductor detectors, while suitable for a wide range of applications, is far less than what can be achieved with X-ray optics. A direct detector that could rival the energy resolution of optics could dramatically improve the efficiency of a multitude of experiments, as experiments are often repeated at a number of different energies. Very high-energy-resolution detectors could make these experiments parallel, rather than serial. Low-background, high-spatial-resolution neutron detectors: Low-background detectors would significantly improve experiments that probe excitations (phonons, spin excitations, rotation, and diffusion in polymers and molecular substances, etc.) in condensed matter. Improved spatial resolution would greatly benefit radiography, tomography, phase-contrast imaging, and holography. Improved acquisition and visualization tools: In the past, with the limited variety of slow detectors, it was straightforward to visualize data as it was being acquired (and adjust experimental conditions accordingly) to create a compact data set that the user could easily transport. As detector complexity and data rates explode, this becomes much more challenging. Three goals were identified as important for coping with the growing data volume from high-speed detectors: Facilitate better algorithm development. In particular, algorithms that can minimize the quantity of data stored. Improve community-driven mechanisms to reduce data protocols and enhance quantitative, interactive visualization tools. Develop and distribute community-developed, detector-specific simulation tools. Aim for parallelization to take advantage of high-performance analysis platforms. Improved analysis

  17. Neutronic effects on tungsten-186 double neutron capture

    Science.gov (United States)

    Garland, Marc Alan

    Rhenium-188, a daughter product of tungsten-188, is an isotope of great interest in therapeutic nuclear medicine, being used in dozens of laboratory and clinical investigations worldwide. Applications include various cancer therapy strategies, treatment of rheumatoid arthritis, prevention of restenosis following coronary artery angioplasty, and palliation of bone pain associated with cancer metastases. With its half-life of 17 hours, 2.12 MeV (maximum) beta-particle emission, chemical similarity to technetium-99m (the most widely used diagnostic radioisotope), and its availability in a convenient tungsten-188/rhenium-188 generator system, rhenium-188 is a superb candidate for a broad range of applications. Production of 188W is typically via double neutron capture by 186W in a high flux nuclear reactor, predominantly the High Flux Isotope Reactor at the Oak Ridge National Laboratory in Tennessee. Experience at HFIR has shown that production yields (measured in Ci of 188W produced per g of 186W target) decrease considerably as target size increases. While the phenomenon of neutron resonance self-shielding would be expected to produce such an effect, temperature effects on neutron flux distribution and neutron capture rates may also be involved. Experimental investigations of these phenomena have not been previously performed. The work presented in this thesis evaluates the factors that contribute to the decrease in 188W yield from both theoretical and experimental standpoints. Neutron self-shielding and temperature effects were characterized to develop a strategy for target design that would optimize production yield, an important factor in minimizing health care costs. It was determined that decrease in yield due to neutron self-shielding can be attributed to depletion of epithermal neutrons at resonant energies, most significantly within the initial 0.4 mm depth of the target. The results from these studies further show that 188W yield in the interior of the

  18. Neutrons and Fundamental Symmetries

    Energy Technology Data Exchange (ETDEWEB)

    Plaster, Bradley [Univ. of Kentucky, Lexington, KY (United States). Dept. of Physics and Astronomy

    2016-01-11

    The research supported by this project addressed fundamental open physics questions via experiments with subatomic particles. In particular, neutrons constitute an especially ideal “laboratory” for fundamental physics tests, as their sensitivities to the four known forces of nature permit a broad range of tests of the so-called “Standard Model”, our current best physics model for the interactions of subatomic particles. Although the Standard Model has been a triumphant success for physics, it does not provide satisfactory answers to some of the most fundamental open questions in physics, such as: are there additional forces of nature beyond the gravitational, electromagnetic, weak nuclear, and strong nuclear forces?, or why does our universe consist of more matter than anti-matter? This project also contributed significantly to the training of the next generation of scientists, of considerable value to the public. Young scientists, ranging from undergraduate students to graduate students to post-doctoral researchers, made significant contributions to the work carried out under this project.

  19. Binary Neutron Star Mergers.

    Science.gov (United States)

    Faber, Joshua A; Rasio, Frederic A

    2012-01-01

    We review the current status of studies of the coalescence of binary neutron star systems. We begin with a discussion of the formation channels of merging binaries and we discuss the most recent theoretical predictions for merger rates. Next, we turn to the quasi-equilibrium formalisms that are used to study binaries prior to the merger phase and to generate initial data for fully dynamical simulations. The quasi-equilibrium approximation has played a key role in developing our understanding of the physics of binary coalescence and, in particular, of the orbital instability processes that can drive binaries to merger at the end of their lifetimes. We then turn to the numerical techniques used in dynamical simulations, including relativistic formalisms, (magneto-)hydrodynamics, gravitational-wave extraction techniques, and nuclear microphysics treatments. This is followed by a summary of the simulations performed across the field to date, including the most recent results from both fully relativistic and microphysically detailed simulations. Finally, we discuss the likely directions for the field as we transition from the first to the second generation of gravitational-wave interferometers and while supercomputers reach the petascale frontier.

  20. Binary Neutron Star Mergers

    Directory of Open Access Journals (Sweden)

    Joshua A. Faber

    2012-07-01

    Full Text Available We review the current status of studies of the coalescence of binary neutron star systems. We begin with a discussion of the formation channels of merging binaries and we discuss the most recent theoretical predictions for merger rates. Next, we turn to the quasi-equilibrium formalisms that are used to study binaries prior to the merger phase and to generate initial data for fully dynamical simulations. The quasi-equilibrium approximation has played a key role in developing our understanding of the physics of binary coalescence and, in particular, of the orbital instability processes that can drive binaries to merger at the end of their lifetimes. We then turn to the numerical techniques used in dynamical simulations, including relativistic formalisms, (magneto-hydrodynamics, gravitational-wave extraction techniques, and nuclear microphysics treatments. This is followed by a summary of the simulations performed across the field to date, including the most recent results from both fully relativistic and microphysically detailed simulations. Finally, we discuss the likely directions for the field as we transition from the first to the second generation of gravitational-wave interferometers and while supercomputers reach the petascale frontier.

  1. Passive neutron assay of irradiated nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hsue, S.T.; Stewart, J.E.; Kaieda, K.; Halbig, J.K.; Phillips, J.R.; Lee, D.M.; Hatcher, C.R.

    1979-02-01

    Passive neutron assay of irradiated nuclear fuel has been investigated by calculations and experiments as a simple, complementary technique to the gamma assay. From the calculations it was found that the neutron emission arises mainly from the curium isotopes, the neutrons exhibit very good penetrability of the assemblies, and the neutron multiplication is not affected by the burnup. From the experiments on BWR and PWR assemblies, the neutron emission rate is proportional to burnup raised to 3.4 power. The investigations indicate that the passive neutron assay is a simple and useful technique to determine the consistency of burnups between assemblies.

  2. Production of epithermal neutron beams for BNCT

    CERN Document Server

    Bisceglie, E; Colonna, N; Paticchio, V; Santorelli, P; Variale, V

    2002-01-01

    The use of boron neutron capture therapy (BNCT) for the treatment of deep-seated tumors requires neutron beams of suitable energy and intensity. Simulations indicate the optimal energy to reside in the epithermal region, in particular between 1 and 10 keV. Therapeutic neutron beams with high spectral purity in this energy range could be produced with accelerator-based neutron sources through a suitable neutron-producing reaction. Herein, we report on different solutions that have been investigated as possible sources of epithermal neutron beams for BNCT. The potential use of such sources for a hospital-based therapeutic facility is discussed.

  3. Overview. Department of Environmental and Radiation Transport Physics. Section 6

    Energy Technology Data Exchange (ETDEWEB)

    Loskiewicz, J. [Institute of Nuclear Physics, Cracow (Poland)

    1995-12-31

    Research activities in the Department of Environmental and Radiation Transport Physics are carried out by three Laboratories: Laboratory of Environmental Physics, Laboratory of Neutron Transport Physics and Laboratory of Physics and Modeling of Radiation Transport. The researches provided in 1994 cover: tracer transport and flows in porous media, studies on pollution in atmospheric air, physics of molecular phenomena in chromatographic detectors, studies on neutron transport in heterogenous media, studies on evaluation of neutron cross-section in the thermal region, studies on theory and utilization of neural network in data evaluation, numerical modelling of particle cascades for particle accelerator shielding purpose. In this section the description of mentioned activities as well as the information about personnel employed in the Department, papers and reports published in 1994, contribution to conferences and grants is also given.

  4. Precision Neutron Scattering Length Measurements with Neutron Interferometry

    Science.gov (United States)

    Huber, M. G.; Arif, M.; Jacobson, D. L.; Pushin, D. A.; Abutaleb, M. O.; Shahi, C. B.; Wietfeldt, F. E.; Black, T. C.

    2011-10-01

    Since its inception, single-crystal neutron interferometry has often been utilized for precise neutron scattering length, b, measurements. Scattering length data of light nuclei is particularly important in the study of few nucleon interactions as b can be predicted by two + three nucleon interaction (NI) models. As such they provide a critical test of the accuracy 2+3 NI models. Nuclear effective field theories also make use of light nuclei b in parameterizing mean-field behavior. The NIST neutron interferometer and optics facility has measured b to less than 0.8% relative uncertainty in polarized 3He and to less than 0.1% relative uncertainty in H, D, and unpolarized 3He. A neutron interferometer consists of a perfect silicon crystal machined such that there are three separate blades on a common base. Neutrons are Bragg diffracted in the blades to produce two spatially separate (yet coherent) beam paths much like an optical Mach-Zehnder interferometer. A gas sample placed in one of the beam paths of the interferometer causes a phase difference between the two paths which is proportional to b. This talk will focus on the latest scattering length measurement for n-4He which ran at NIST in Fall/Winter 2010 and is currently being analyzed.

  5. Application of adjoint Monte Carlo to accelerate simulations of mono-directional beams in treatment planning for Boron Neutron Capture Therapy

    NARCIS (Netherlands)

    Nievaart, V.A.; Legrady, D.; Moss, R.L.; Kloosterman, J.L.; Van der Hagen, T.H.; Van Dam, H.

    2007-01-01

    This paper deals with the application of the adjoint transport theory in order to optimize Monte Carlo based radiotherapy treatment planning. The technique is applied to Boron Neutron Capture Therapy where most often mixed beams of neutrons and gammas are involved. In normal forward Monte Carlo

  6. Uncertainty in the delayed neutron fraction in fuel assembly depletion calculations

    Directory of Open Access Journals (Sweden)

    Aures Alexander

    2017-01-01

    Full Text Available This study presents uncertainty and sensitivity analyses of the delayed neutron fraction of light water reactor and sodium-cooled fast reactor fuel assemblies. For these analyses, the sampling-based XSUSA methodology is used to propagate cross section uncertainties in neutron transport and depletion calculations. Cross section data is varied according to the SCALE 6.1 covariance library. Since this library includes nu-bar uncertainties only for the total values, it has been supplemented by delayed nu-bar uncertainties from the covariance data of the JENDL-4.0 nuclear data library. The neutron transport and depletion calculations are performed with the TRITON/NEWT sequence of the SCALE 6.1 package. The evolution of the delayed neutron fraction uncertainty over burn-up is analysed without and with the consideration of delayed nu-bar uncertainties. Moreover, the main contributors to the result uncertainty are determined. In all cases, the delayed nu-bar uncertainties increase the delayed neutron fraction uncertainty. Depending on the fuel composition, the delayed nu-bar values of uranium and plutonium in fact give the main contributions to the delayed neutron fraction uncertainty for the LWR fuel assemblies. For the SFR case, the uncertainty of the scattering cross section of U-238 is the main contributor.

  7. Accuracy of neutron dose evaluation in the area monitoring for LHD experiments

    CERN Document Server

    Yamanishi, H; Uda, T; Tanahashi, S; Saitou, M; Handa, H

    2000-01-01

    The error in the evaluation of neutron dose during calculation of the neutron field around the large helical device (LHD) in D-D operation is discussed. The expected neutron dose at each monitoring point was derived from the dose conversion factor and neutron fluence data, which was calculated with the radiation transport code DOT-3.5. In contrast, the detected dose at the neutron counter was obtained from the fluence data and the detector response given by calculation with MCNP-4b. The neutron counter used in these calculations consisted of a helium-3 proportional counter with a cylindrical polyethylene moderator. According to the results of the calculations, the ratio of the detected dose to the expected dose was found to lie in the range 1.0-3.0 on the outdoor monitoring points. Since the response of a single neutron counter may lead to inconsistencies in the dose conversion factor, we attempted to minimize these inconsistencies by using a pair of counters with moderators of different thickness. The ratio ...

  8. Scientific Design of the New Neutron Radiography Facility (SANRAD) at SAFARI-1 for South Africa

    Science.gov (United States)

    de Beer, F. C.; Gruenauer, F.; Radebe, J. M.; Modise, T.; Schillinger, B.

    The final scientific design for an upgraded neutron radiography/tomography facility at beam port no.2 of the SAFARI-1 nuclear research reactor has been performed through expert advice from Physics Consulting, FRMII in Germany and IPEN, Brazil. A need to upgrade the facility became apparent due to the identification of various deficiencies of the current SANRAD facility during an IAEA-sponsored expert mission of international scientists to Necsa, South Africa. A lack of adequate shielding that results in high neutron background on the beam port floor, a mismatch in the collimator aperture to the core that results in a high gradient in neutron flux on the imaging plane and due to a relative low L/D the quality of the radiographs are poor, are a number of deficiencies to name a few.The new design, based on results of Monte Carlo (MCNP-X) simulations of neutron- and gamma transport from the reactor core and through the new facility, is being outlined. The scientific design philosophy, neutron optics and imaging capabilities that include the utilization of fission neutrons, thermal neutrons, and gamma-rays emerging from the core of SAFARI-1 are discussed.

  9. Elimination of ghosting artifacts from wavelength-shifting fiber neutron detectors

    Energy Technology Data Exchange (ETDEWEB)

    Wang, C. L.; Diawara, Y.; Hannan, B. W.; Hodges, J. P. [Instrument and Source Design Division, Neutron Sciences Directorate, Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Clonts, L. G. [Research Accelerator Division, Neutron Sciences Directorate, Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States)

    2013-01-15

    Misassignment of neutron position (ghosting) produces artifacts which have been observed in wavelength-shifting (WLS) fiber detectors developed for time-of-flight (TOF) neutron powder diffraction. In position-sensitive detectors (PSDs) with WLS fiber encoding, thermal and cold neutrons interact with a monolithic {sup 6}LiF/ZnS:Ag scintillator screen, and scintillation photons are generated and transported through the crossed fibers to photomultipliers (PMTs). The neutron position is determined by photon counts in the PMTs within a preset time window. Ghosting occurs when neutrons hit the group boundaries of two neighboring PMTs for x-position multiplexing, which is modeled as resulting from a long travel length (about 3-5 mm) of a small number of scintillation photons. This model is supported by the change observed in aperture images when the threshold number for photon-pulses was adjusted for neutron event determination. When the threshold number of photon-pulses was set above 10 for each PMT, the ghost peaks in the aperture images and TOF spectra of powder diffraction were strongly suppressed or completely eliminated, and the intrinsic background levels of the WLS detectors were significantly reduced. Our result indicates that WLS fiber detector is a promising alternative for {sup 3}He PSDs for neutron scattering.

  10. Elimination of ghosting artifacts from wavelength-shifting fiber neutron detectors.

    Science.gov (United States)

    Wang, C L; Clonts, L G; Diawara, Y; Hannan, B W; Hodges, J P

    2013-01-01

    Misassignment of neutron position (ghosting) produces artifacts which have been observed in wavelength-shifting (WLS) fiber detectors developed for time-of-flight (TOF) neutron powder diffraction. In position-sensitive detectors (PSDs) with WLS fiber encoding, thermal and cold neutrons interact with a monolithic (6)LiF/ZnS:Ag scintillator screen, and scintillation photons are generated and transported through the crossed fibers to photomultipliers (PMTs). The neutron position is determined by photon counts in the PMTs within a preset time window. Ghosting occurs when neutrons hit the group boundaries of two neighboring PMTs for x-position multiplexing, which is modeled as resulting from a long travel length (about 3-5 mm) of a small number of scintillation photons. This model is supported by the change observed in aperture images when the threshold number for photon-pulses was adjusted for neutron event determination. When the threshold number of photon-pulses was set above 10 for each PMT, the ghost peaks in the aperture images and TOF spectra of powder diffraction were strongly suppressed or completely eliminated, and the intrinsic background levels of the WLS detectors were significantly reduced. Our result indicates that WLS fiber detector is a promising alternative for (3)He PSDs for neutron scattering.

  11. Evaluation of energy response of neutron rem monitor applied to high-energy accelerator facilities

    CERN Document Server

    Nakane, Y; Sakamoto, Y

    2003-01-01

    A neutron rem monitor was newly developed for applying to the high-intensity proton accelerator facility (J-PARC) that is under construction as a joint project between the Japan Atomic Energy Research Institute and the High Energy Accelerator Research Organization. To measure the dose rate accurately for wide energy range of neutrons from thermal to high-energy region, the neutron rem monitor was fabricated by adding a lead breeder layer to a conventional neutron rem monitor. The energy response of the monitor was evaluated by using neutron transport calculations for the energy range from thermal to 150 MeV. For verifying the results, the response was measured at neutron fields for the energy range from thermal to 65 MeV. The comparisons between the energy response and dose conversion coefficients show that the newly developed neutron rem monitor has a good performance in energy response up to 150 MeV, suggesting that the present study offered prospects of a practical fabrication of the rem monitor applicable...

  12. Evaluation of energy response of neutron rem monitor applied to high-energy accelerator facilities

    Energy Technology Data Exchange (ETDEWEB)

    Nakane, Yoshihiro; Harada, Yasunori; Sakamoto, Yukio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    2003-03-01

    A neutron rem monitor was newly developed for applying to the high-intensity proton accelerator facility (J-PARC) that is under construction as a joint project between the Japan Atomic Energy Research Institute and the High Energy Accelerator Research Organization. To measure the dose rate accurately for wide energy range of neutrons from thermal to high-energy region, the neutron rem monitor was fabricated by adding a lead breeder layer to a conventional neutron rem monitor. The energy response of the monitor was evaluated by using neutron transport calculations for the energy range from thermal to 150 MeV. For verifying the results, the response was measured at neutron fields for the energy range from thermal to 65 MeV. The comparisons between the energy response and dose conversion coefficients show that the newly developed neutron rem monitor has a good performance in energy response up to 150 MeV, suggesting that the present study offered prospects of a practical fabrication of the rem monitor applicable to the high intensity proton accelerator facility. (author)

  13. Nodal collocation approximation for the multidimensional PL equations applied to transport source problems

    Energy Technology Data Exchange (ETDEWEB)

    Verdu, G. [Departamento de Ingenieria Quimica Y Nuclear, Universitat Politecnica de Valencia, Cami de Vera, 14, 46022. Valencia (Spain); Capilla, M.; Talavera, C. F.; Ginestar, D. [Dept. of Nuclear Engineering, Departamento de Matematica Aplicada, Universitat Politecnica de Valencia, Cami de Vera, 14, 46022. Valencia (Spain)

    2012-07-01

    PL equations are classical high order approximations to the transport equations which are based on the expansion of the angular dependence of the angular neutron flux and the nuclear cross sections in terms of spherical harmonics. A nodal collocation method is used to discretize the PL equations associated with a neutron source transport problem. The performance of the method is tested solving two 1D problems with analytical solution for the transport equation and a classical 2D problem. (authors)

  14. Neutron Tomography at the Los Alamos Neutron Science Center

    Energy Technology Data Exchange (ETDEWEB)

    Myers, William Riley [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-08-07

    Neutron imaging is an incredibly powerful tool for non-destructive sample characterization and materials science. Neutron tomography is one technique that results in a three-dimensional model of the sample, representing the interaction of the neutrons with the sample. This relies both on reliable data acquisition and on image processing after acquisition. Over the course of the project, the focus has changed from the former to the latter, culminating in a large-scale reconstruction of a meter-long fossilized skull. The full reconstruction is not yet complete, though tools have been developed to improve the speed and accuracy of the reconstruction. This project helps to improve the capabilities of LANSCE and LANL with regards to imaging large or unwieldy objects.

  15. Fission-neutrons source with fast neutron-emission timing

    Energy Technology Data Exchange (ETDEWEB)

    Rusev, G., E-mail: rusev@lanl.gov; Baramsai, B.; Bond, E.M.; Jandel, M.

    2016-05-01

    A neutron source with fast timing has been built to help with detector-response measurements. The source is based on the neutron emission from the spontaneous fission of {sup 252}Cf. The time is provided by registering the fission fragments in a layer of a thin scintillation film with a signal rise time of 1 ns. The scintillation light output is measured by two silicon photomultipliers with rise time of 0.5 ns. Overall time resolution of the source is 0.3 ns. Design of the source and test measurements using it are described. An example application of the source for determining the neutron/gamma pulse-shape discrimination by a stilbene crystal is given.

  16. The Fate of Merging Neutron Stars

    Science.gov (United States)

    Kohler, Susanna

    2017-08-01

    A rapidly spinning, highly magnetized neutron star is one possible outcome when two smaller neutron stars merge. [Casey Reed/Penn State University]When two neutron stars collide, the new object that they make can reveal information about the interior physics of neutron stars. New theoretical work explores what we should be seeing, and what it can teach us.Neutron Star or Black Hole?So far, the only systems from which weve detected gravitational waves are merging black holes. But other compact-object binaries exist and are expected to merge on observable timescales in particular, binary neutron stars. When two neutron stars merge, the resulting object falls into one of three categories:a stable neutron star,a black hole, ora supramassive neutron star, a large neutron star thats supported by its rotation but will eventually collapse to a black hole after it loses angular momentum.Histograms of the initial (left) and final (right) distributions of objects in the authors simulations, for five different equations of state. Most cases resulted primarily in the formation of neutron stars (NSs) or supramassive neutron stars (sNSs), not black holes (BHs). [Piro et al. 2017]Whether a binary-neutron-star merger results in another neutron star, a black hole, or a supramassive neutron star depends on the final mass of the remnant and what the correct equation of state is that describes the interiors of neutron stars a longstanding astrophysical puzzle.In a recent study, a team of scientists led by Anthony Piro (Carnegie Observatories) estimated which of these outcomes we should expect for mergers of binary neutron stars. The teams results along with future observations of binary neutron stars may help us to eventually pin down the equation of state for neutron stars.Merger OutcomesPiro and collaborators used relativistic calculations of spinning and non-spinning neutron stars to estimate the mass range that neutron stars would have for several different realistic equations of

  17. Neutron scattering instrumentation for biology at spallation neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    Pynn, R. [Los Alamos National Laboratory, NM (United States)

    1994-12-31

    Conventional wisdom holds that since biological entities are large, they must be studied with cold neutrons, a domain in which reactor sources of neutrons are often supposed to be pre-eminent. In fact, the current generation of pulsed spallation neutron sources, such as LANSCE at Los Alamos and ISIS in the United Kingdom, has demonstrated a capability for small angle scattering (SANS) - a typical cold- neutron application - that was not anticipated five years ago. Although no one has yet built a Laue diffractometer at a pulsed spallation source, calculations show that such an instrument would provide an exceptional capability for protein crystallography at one of the existing high-power spoliation sources. Even more exciting is the prospect of installing such spectrometers either at a next-generation, short-pulse spallation source or at a long-pulse spallation source. A recent Los Alamos study has shown that a one-megawatt, short-pulse source, which is an order of magnitude more powerful than LANSCE, could be built with today`s technology. In Europe, a preconceptual design study for a five-megawatt source is under way. Although such short-pulse sources are likely to be the wave of the future, they may not be necessary for some applications - such as Laue diffraction - which can be performed very well at a long-pulse spoliation source. Recently, it has been argued by Mezei that a facility that combines a short-pulse spallation source similar to LANSCE, with a one-megawatt, long-pulse spallation source would provide a cost-effective solution to the global shortage of neutrons for research. The basis for this assertion as well as the performance of some existing neutron spectrometers at short-pulse sources will be examined in this presentation.

  18. Particle Filter-Based Recursive Data Fusion With Sensor Indexing for Large Core Neutron Flux Estimation

    Science.gov (United States)

    Tamboli, Prakash Kumar; Duttagupta, Siddhartha P.; Roy, Kallol

    2017-06-01

    We introduce a sequential importance sampling particle filter (PF)-based multisensor multivariate nonlinear estimator for estimating the in-core neutron flux distribution for pressurized heavy water reactor core. Many critical applications such as reactor protection and control rely upon neutron flux information, and thus their reliability is of utmost importance. The point kinetic model based on neutron transport conveniently explains the dynamics of nuclear reactor. The neutron flux in the large core loosely coupled reactor is sensed by multiple sensors measuring point fluxes located at various locations inside the reactor core. The flux values are coupled to each other through diffusion equation. The coupling facilitates redundancy in the information. It is shown that multiple independent data about the localized flux can be fused together to enhance the estimation accuracy to a great extent. We also propose the sensor anomaly handling feature in multisensor PF to maintain the estimation process even when the sensor is faulty or generates data anomaly.

  19. Probing the nuclear equation of state by heavy-ion reactions and neutron star properties

    Energy Technology Data Exchange (ETDEWEB)

    Sahu, P.K.; Cassing, W.; Thoma, M.H. [Inst. fuer Theoretische Physik, Univ. Giessen (Germany)

    1998-06-01

    We discuss the nuclear equation of state (EOS) using a non-linear relativistic transport model. From the baryon flow for Ni + Ni as well as Au + Au systems we find that the strength of the vector potential has to be reduced at high density or at high relative momenta to describe the experimental flow data at 1-2 A GeV. We use the same dynamical model to calculate the nuclear EOS and then employ this EOS to neutron star structure calculations. We consider the core of the neutron star to be composed of neutrons with an admixture of protons, electrons, muons, sigmas and lambdas at zero temperature. We find that the nuclear equation of state is softer at high densities and hence the maximum mass and the radius of the neutron star are in the observable range of M {proportional_to} 1.7 M{sub s}un and R = 8 km, respectively. (orig.)

  20. Neutron-induced damage evolution under Beam Raster Scanner conditions for IFMIF

    Energy Technology Data Exchange (ETDEWEB)

    Mota, Fernando, E-mail: fernando.mota@ciemat.es [Laboratorio Nacional de Fusion por Confinamiento Magnetico - CIEMAT, 28040 Madrid (Spain); Ortiz, Christophe J., E-mail: christophe.ortiz@ciemat.es [Laboratorio Nacional de Fusion por Confinamiento Magnetico - CIEMAT, 28040 Madrid (Spain); Ibarra, Angel, E-mail: Angel.ibarra@ciemat.es [Laboratorio Nacional de Fusion por Confinamiento Magnetico - CIEMAT, 28040 Madrid (Spain); Vila, Rafael, E-mail: Rafael.vila@ciemat.es [Laboratorio Nacional de Fusion por Confinamiento Magnetico - CIEMAT, 28040 Madrid (Spain)

    2011-10-01

    The formation and evolution of defects in materials irradiated with a homogeneous neutron source and with the Beam Raster Scanner (BRS) solution was investigated. The intensity neutron source fluctuations inherent to the BRS system were determined using the neutron transport McDeLicious code. Defects generated during irradiation were calculated using the binary collision approximation MARLOWE code, using the primary knock-on atom (PKA) energy spectrum resulting from neutron interactions with the material. In order to predict the evolution of defects during irradiation, a Rate Theory model based on ab initio parameters was developed. Our model accounts for the migration of mobile defects, the formation of clusters and their recombination. As an example, we investigated defect evolution in Fe irradiated at room temperature in both beam configurations. Simulation results clearly indicate that the defect evolution expected in the BRS configuration is nearly the same as the one expected in a homogeneous irradiation system.