Science.gov

Sample records for neutronic codes coupled

  1. Interface requirements to couple thermal-hydraulic codes to 3D neutronic codes

    SciTech Connect

    Langenbuch, S.; Austregesilo, H.; Velkov, K.

    1997-07-01

    The present situation of thermalhydraulics codes and 3D neutronics codes is briefly described and general considerations for coupling of these codes are discussed. Two different basic approaches of coupling are identified and their relative advantages and disadvantages are discussed. The implementation of the coupling for 3D neutronics codes in the system ATHLET is presented. Meanwhile, this interface is used for coupling three different 3D neutronics codes.

  2. 3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

    SciTech Connect

    Langenbuch, S.; Velkov, K.; Lizorkin, M.

    1997-07-01

    This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

  3. TART97 a coupled neutron-photon 3-D, combinatorial geometry Monte Carlo transport code

    SciTech Connect

    Cullen, D.E.

    1997-11-22

    TART97 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo transport code. This code can on any modern computer. It is a complete system to assist you with input preparation, running Monte Carlo calculations, and analysis of output results. TART97 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART97 is distributed on CD. This CD contains on- line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART97 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART97 and its data riles.

  4. Coupled neutronic and thermal-hydraulic code benchmark activities at the International Nuclear Safety Center.

    SciTech Connect

    Podlazov, L. N.

    1998-07-29

    Two realistic benchmark problems are defined and used to assess the performance of coupled thermal-hydraulic and neutronic codes used in simulating dynamic processes in VVER-1000 and RBMK reactor systems. One of the problems simulates a design basis accident involving the ejection of three control and protection system rods from a VVER-1000 reactor. The other is based on a postulated rod withdrawal from an operating RBMK reactor. Preliminary results calculated by various codes are compared. While these results show significant differences, the intercomparisons performed so far provide a basis for further evaluation of code limitations and modeling assumptions.

  5. Parameterization of nuclear cross-sections for coupled neutronic- thermalhydraulic codes

    SciTech Connect

    Miro, R.; Verdu, G.; Barrachina, T.; Rosello, O.

    2006-07-01

    The present work consists of developing an in-house methodology, called SIMTAB, to characterize, in a simplified way, the reactor core of LWR Nuclear Power Plants. Specifically, a cross-sections and kinetic parameters set are obtained as a function of the prompt and control variables. So that, the core can be modeled using a limited number of neutronic regions, in such a way that the reactor kinetic behavior is properly characterized. This simplification of the reactor core permits, from an operative point of view, the use of few cross sections data sets in coupled 3D neutronic-thermalhydraulic codes. (authors)

  6. Coupled Neutron-Photon, 3-D, Combinatorial Geometry, Time Dependent, Monte Carlo Transport Code System.

    Energy Science and Technology Software Center (ESTSC)

    2013-06-24

    Version 07 TART2012 is a coupled neutron-photon Monte Carlo transport code designed to use three-dimensional (3-D) combinatorial geometry. Neutron and/or photon sources as well as neutron induced photon production can be tracked. It is a complete system to assist you with input preparation, running Monte Carlo calculations, and analysis of output results. TART2012 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared tomore » other similar codes. Use of the entire system can save you a great deal of time and energy. TART2012 extends the general utility of the code to even more areas of application than available in previous releases by concentrating on improving the physics, particularly with regard to improved treatment of neutron fission, resonance self-shielding, molecular binding, and extending input options used by the code. Several utilities are included for creating input files and displaying TART results and data. TART2012 uses the latest ENDF/B-VI, Release 8, data. New for TART2012 is the use of continuous energy neutron cross sections, in addition to its traditional multigroup cross sections. For neutron interaction, the data are derived using ENDF-ENDL2005 and include both continuous energy cross sections and 700 group neutron data derived using a combination of ENDF/B-VI, Release 8, and ENDL data. The 700 group structure extends from 10-5 eV up to 1 GeV. Presently nuclear data are only available up to 20 MeV, so that only 616 of the groups are currently used. For photon interaction, 701 point photon data were derived using the Livermore EPDL97 file. The new 701 point structure extends from 100 eV up to 1 GeV, and is currently used over this entire energy range. TART2012 completely supersedes all older versions of TART, and it is strongly recommended that one use only the most recent version of TART2012 and its data files. Check author’s homepage for related information: http

  7. A Coupled Neutron-Photon 3-D Combinatorial Geometry Monte Carlo Transport Code

    Energy Science and Technology Software Center (ESTSC)

    1998-06-12

    TART97 is a coupled neutron-photon, 3 dimensional, combinatorial geometry, time dependent Monte Carlo 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. TART97 is also incredibly fast: if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system canmore » save you a great deal of time and energy. TART 97 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART97 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART97 and ist data files.« less

  8. Cynod: A Neutronics Code for Pebble Bed Modular Reactor Coupled Transient Analysis

    SciTech Connect

    Hikaru Hiruta; Abderrafi M. Ougouag; Hans D. Gougar; Javier Ortensi

    2008-09-01

    The Pebble Bed Reactor (PBR) is one of the two concepts currently considered for development into the Next Generation Nuclear Plant (NGNP). This interest is due, in particular, to the concept’s inherent safety characteristics. In order to verify and confirm the design safety characteristics of the PBR computational tools must be developed that treat the range of phenomena that are expected to be important for this type of reactors. This paper presents a recently developed 2D R-Z cylindrical nodal kinetics code and shows some of its capabilities by applying it to a set of known and relevant benchmarks. The new code has been coupled to the thermal hydraulics code THERMIX/KONVEK[1] for application to the simulation of very fast transients in PBRs. The new code, CYNOD, has been written starting with a fixed source solver extracted from the nodal cylindrical geometry solver contained within the PEBBED code. The fixed source solver was then incorporated into a kinetic solver.. The new code inherits the spatial solver characteristics of the nodal solver within PEBBED. Thus, the time-dependent neutron diffusion equation expressed analytically in each node of the R-Z cylindrical geometry sub-domain (or node) is transformed into one-dimensional equations by means of the usual transverse integration procedure. The one-dimensional diffusion equations in each of the directions are then solved using the analytic Green’s function method. The resulting equations for the entire domain are then re-cast in the form of the Direct Coarse Mesh Finite Difference (D-CMFD) for convenience of solution. The implicit Euler method is used for the time variable discretization. In order to correctly treat the cusping effect for nodes that contain a partially inserted control rod a method is used that takes advantage of the Green’s function solution available in the intrinsic method. In this corrected treatment, the nodes are re-homogenized using axial flux shapes reconstructed based on the

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

    SciTech Connect

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

    2005-02-15

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

  10. TART98 a coupled neutron-photon 3-D, combinatorial geometry time dependent Monte Carlo Transport code

    SciTech Connect

    Cullen, D E

    1998-11-22

    TART98 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input preparation, running Monte Carlo calculations, and analysis of output results. TART98 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART98 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART98 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART98 and its data files.

  11. Code System for 2-Group, 3D Neutronic Kinetics Calculations Coupled to Core Thermal Hydraulics.

    Energy Science and Technology Software Center (ESTSC)

    2000-05-12

    Version 00 QUARK is a combined computer program comprising a revised version of the QUANDRY three-dimensional, two-group neutron kinetics code and an upgraded version of the COBRA transient core analysis code (COBRA-EN). Starting from either a critical steady-state (k-effective or critical dilute Boron problem) or a subcritical steady-state (fixed source problem) in a PWR plant, the code allows one to simulate the neutronic and thermal-hydraulic core transient response to reactivity accidents initiated both inside themore » vessel (such as a control rod ejection) and outside the vessel (such as the sudden change of the Boron concentration in the coolant). QUARK output can be used as input to PSR-470/NORMA-FP to perform a subchannel analysis from converged coarse-mesh nodal solutions.« less

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

    SciTech Connect

    Lu, S.

    2002-07-01

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

  13. Coded source neutron imaging

    SciTech Connect

    Bingham, Philip R; Santos-Villalobos, Hector J

    2011-01-01

    Coded aperture techniques have been applied to neutron radiography to address limitations in neutron flux and resolution of neutron detectors in a system labeled coded source imaging (CSI). By coding the neutron source, a magnified imaging system is designed with small spot size aperture holes (10 and 100 m) for improved resolution beyond the detector limits and with many holes in the aperture (50% open) to account for flux losses due to the small pinhole size. An introduction to neutron radiography and coded aperture imaging is presented. A system design is developed for a CSI system with a development of equations for limitations on the system based on the coded image requirements and the neutron source characteristics of size and divergence. Simulation has been applied to the design using McStas to provide qualitative measures of performance with simulations of pinhole array objects followed by a quantitative measure through simulation of a tilted edge and calculation of the modulation transfer function (MTF) from the line spread function. MTF results for both 100um and 10um aperture hole diameters show resolutions matching the hole diameters.

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

    SciTech Connect

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

    2007-07-01

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

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

    SciTech Connect

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

    2004-10-15

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

  16. Parallelization of the MAAP-A code neutronics/thermal hydraulics coupling

    SciTech Connect

    Froehle, P.H.; Wei, T.Y.C.; Weber, D.P.; Henry, R.E.

    1998-12-31

    A major new feature, one-dimensional space-time kinetics, has been added to a developmental version of the MAAP code through the introduction of the DIF3D-K module. This code is referred to as MAAP-A. To reduce the overall job time required, a capability has been provided to run the MAAP-A code in parallel. The parallel version of MAAP-A utilizes two machines running in parallel, with the DIF3D-K module executing on one machine and the rest of the MAAP-A code executing on the other machine. Timing results obtained during the development of the capability indicate that reductions in time of 30--40% are possible. The parallel version can be run on two SPARC 20 (SUN OS 5.5) workstations connected through the ethernet. MPI (Message Passing Interface standard) needs to be implemented on the machines. If necessary the parallel version can also be run on only one machine. The results obtained running in this one-machine mode identically match the results obtained from the serial version of the code.

  17. Coupled moderator neutronics

    SciTech Connect

    Russell, G.J.; Pitcher, E.J.; Ferguson, P.D.

    1995-12-01

    Optimizing the neutronic performance of a coupled-moderator system for a Long-Pulse Spallation Source is a new and challenging area for the spallation target-system designer. For optimal performance of a neutron source, it is essential to have good communication with instrument scientists to obtain proper design criteria and continued interaction with mechanical, thermal-hydraulic, and materials engineers to attain a practical design. A good comprehension of the basics of coupled-moderator neutronics will aid in the proper design of a target system for a Long-Pulse Spallation Source.

  18. Estimation of relative biological effectiveness for boron neutron capture therapy using the PHITS code coupled with a microdosimetric kinetic model

    PubMed Central

    Horiguchi, Hironori; Sato, Tatsuhiko; Kumada, Hiroaki; Yamamoto, Tetsuya; Sakae, Takeji

    2015-01-01

    The absorbed doses deposited by boron neutron capture therapy (BNCT) can be categorized into four components: α and 7Li particles from the 10B(n, α)7Li reaction, 0.54-MeV protons from the 14N(n, p)14C reaction, the recoiled protons from the 1H(n, n) 1H reaction, and photons from the neutron beam and 1H(n, γ)2H reaction. For evaluating the irradiation effect in tumors and the surrounding normal tissues in BNCT, it is of great importance to estimate the relative biological effectiveness (RBE) for each dose component in the same framework. We have, therefore, established a new method for estimating the RBE of all BNCT dose components on the basis of the microdosimetric kinetic model. This method employs the probability density of lineal energy, y, in a subcellular structure as the index for expressing RBE, which can be calculated using the microdosimetric function implemented in the particle transport simulation code (PHITS). The accuracy of this method was tested by comparing the calculated RBE values with corresponding measured data in a water phantom irradiated with an epithermal neutron beam. The calculation technique developed in this study will be useful for biological dose estimation in treatment planning for BNCT. PMID:25428243

  19. Statistical Safety Evaluation of BWR Turbine Trip Scenario Using Coupled Neutron Kinetics and Thermal Hydraulics Analysis Code SKETCH-INS/TRACE5.0

    NASA Astrophysics Data System (ADS)

    Ichikawa, Ryoko; Masuhara, Yasuhiro; Kasahara, Fumio

    The Best Estimate Plus Uncertainty (BEPU) method has been prepared for the regulatory cross-check analysis at Japan Nuclear Energy Safety Organization (JNES) on base of the three-dimensional neutron-kinetics/thermal- hydraulics coupled code SKETCH-INS/TRACE5.0. In the preparation, TRACE5.0 is verified against the large-scale thermal-hydraulic tests carried out with NUPEC facility. These tests were focused on the pressure drop of steam-liquid two phase flow and void fraction distribution. From the comparison of the experimental data with other codes (RELAP5/MOD3.3 and TRAC-BF1), TRACE5.0 was judged better than other codes. It was confirmed that TRACE5.0 has high reliability for thermal hydraulics behavior and are used as a best-estimate code for the statistical safety evaluation. Next, the coupled code SKETCH-INS/TRACE5.0 was applied to turbine trip tests performed at the Peach Bottom-2 BWR4 Plant. The turbine trip event shows the rapid power peak due to the voids collapse with the pressure increase. The analyzed peak value of core power is better simulated than the previous version SKETCH-INS/TRAC-BF1. And the statistical safety evaluation using SKETCH-INS/TRACE5.0 was applied to the loss of load transient for examining the influence of the choice of sampling method.

  20. Coupled Neutron Transport for HZETRN

    NASA Technical Reports Server (NTRS)

    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.

  1. 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

    SciTech Connect

    Morgan C. White

    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, the ability to

  2. Iterative Reconstruction of Coded Source Neutron Radiographs

    SciTech Connect

    Santos-Villalobos, Hector J; Bingham, Philip R; Gregor, Jens

    2012-01-01

    Use of a coded source facilitates high-resolution neutron imaging but requires that the radiographic data be deconvolved. In this paper, we compare direct deconvolution with two different iterative algorithms, namely, one based on direct deconvolution embedded in an MLE-like framework and one based on a geometric model of the neutron beam and a least squares formulation of the inverse imaging problem.

  3. Analysis of a rod withdrawal in a PWR core with the neutronic- thermalhydraulic coupled code RELAP/PARCS and RELAP/VALKIN

    SciTech Connect

    Miro, R.; Maggini, F.; Barrachina, T.; Verdu, G.; Gomez, A.; Ortego, A.; Murillo, J. C.

    2006-07-01

    The Reactor Ejection Accident (REA) belongs to the Reactor Initiated Accidents (RIA) category of accidents and it is part of the licensing basis accident analyses required for pressure water reactors (PWR). The REA at hot zero power (HZP) is characterized by a single rod ejection from a core position with a very low power level. The evolution consists basically of a continuous reactivity insertion. The main feature limiting the consequences of the accident in a PWR is the Doppler Effect. To check the performance of the coupled code RELAP5/PARCS2.5 and RELAP5/VALKIN a REA in Trillo NPP is simulated. These analyses will allow knowing more accurately the PWR real plant phenomenology in the RIA most limiting conditions. (authors)

  4. Secondary neutron source modelling using MCNPX and ALEPH codes

    NASA Astrophysics Data System (ADS)

    Trakas, Christos; Kerkar, Nordine

    2014-06-01

    Monitoring the subcritical state and divergence of reactors requires the presence of neutron sources. But mainly secondary neutrons from these sources feed the ex-core detectors (SRD, Source Range Detector) whose counting rate is correlated with the level of the subcriticality of reactor. In cycle 1, primary neutrons are provided by sources activated outside of the reactor (e.g. Cf252); part of this source can be used for the divergence of cycle 2 (not systematic). A second family of neutron sources is used for the second cycle: the spontaneous neutrons of actinides produced after irradiation of fuel in the first cycle. Both families of sources are not sufficient to efficiently monitor the divergence of the second cycles and following ones, in most reactors. Secondary sources cluster (SSC) fulfil this role. In the present case, the SSC [Sb, Be], after activation in the first cycle (production of Sb124, unstable), produces in subsequent cycles a photo-neutron source by gamma (from Sb124)-neutron (on Be9) reaction. This paper presents the model of the process between irradiation in cycle 1 and cycle 2 results for SRD counting rate at the beginning of cycle 2, using the MCNPX code and the depletion chain ALEPH-V1 (coupling of MCNPX and ORIGEN codes). The results of this simulation are compared with two experimental results of the PWR 1450 MWe-N4 reactors. A good agreement is observed between these results and the simulations. The subcriticality of the reactors is about at -15,000 pcm. Discrepancies on the SRD counting rate between calculations and measurements are in the order of 10%, lower than the combined uncertainty of measurements and code simulation. This comparison validates the AREVA methodology, which allows having an SRD counting rate best-estimate for cycles 2 and next ones and optimizing the position of the SSC, depending on the geographic location of sources, main parameter for optimal monitoring of subcritical states.

  5. Iterative Reconstruction of Coded Source Neutron Radiographs

    SciTech Connect

    Santos-Villalobos, Hector J; Bingham, Philip R; Gregor, Jens

    2013-01-01

    Use of a coded source facilitates high-resolution neutron imaging through magnifications but requires that the radiographic data be deconvolved. A comparison of direct deconvolution with two different iterative algorithms has been performed. One iterative algorithm is based on a maximum likelihood estimation (MLE)-like framework and the second is based on a geometric model of the neutron beam within a least squares formulation of the inverse imaging problem. Simulated data for both uniform and Gaussian shaped source distributions was used for testing to understand the impact of non-uniformities present in neutron beam distributions on the reconstructed images. Results indicate that the model based reconstruction method will match resolution and improve on contrast over convolution methods in the presence of non-uniform sources. Additionally, the model based iterative algorithm provides direct calculation of quantitative transmission values while the convolution based methods must be normalized base on known values.

  6. Fast-neutron, coded-aperture imager

    NASA Astrophysics Data System (ADS)

    Woolf, Richard S.; Phlips, Bernard F.; Hutcheson, Anthony L.; Wulf, Eric A.

    2015-06-01

    This work discusses a large-scale, coded-aperture imager for fast neutrons, building off a proof-of concept instrument developed at the U.S. Naval Research Laboratory (NRL). The Space Science Division at the NRL has a heritage of developing large-scale, mobile systems, using coded-aperture imaging, for long-range γ-ray detection and localization. The fast-neutron, coded-aperture imaging instrument, designed for a mobile unit (20 ft. ISO container), consists of a 32-element array of 15 cm×15 cm×15 cm liquid scintillation detectors (EJ-309) mounted behind a 12×12 pseudorandom coded aperture. The elements of the aperture are composed of 15 cm×15 cm×10 cm blocks of high-density polyethylene (HDPE). The arrangement of the aperture elements produces a shadow pattern on the detector array behind the mask. By measuring of the number of neutron counts per masked and unmasked detector, and with knowledge of the mask pattern, a source image can be deconvolved to obtain a 2-d location. The number of neutrons per detector was obtained by processing the fast signal from each PMT in flash digitizing electronics. Digital pulse shape discrimination (PSD) was performed to filter out the fast-neutron signal from the γ background. The prototype instrument was tested at an indoor facility at the NRL with a 1.8-μCi and 13-μCi 252Cf neutron/γ source at three standoff distances of 9, 15 and 26 m (maximum allowed in the facility) over a 15-min integration time. The imaging and detection capabilities of the instrument were tested by moving the source in half- and one-pixel increments across the image plane. We show a representative sample of the results obtained at one-pixel increments for a standoff distance of 9 m. The 1.8-μCi source was not detected at the 26-m standoff. In order to increase the sensitivity of the instrument, we reduced the fastneutron background by shielding the top, sides and back of the detector array with 10-cm-thick HDPE. This shielding configuration led

  7. The CoCoNuT code: from neutron star oscillations to supernova explosions

    NASA Astrophysics Data System (ADS)

    Cerdá-Durán, P.; Gabler, M.; Müller, E.; Font, J. A.; Stergioulas, N.; Obergaulinger, M.; Aloy, M. A.; DeBrye, N.; Cordero-Carrión, I.; Ibáñez, J. M.

    2013-05-01

    CoCoNuT is a numerical code, that evolves the General relativistic magneto-hydrodynamics equations coupled to the Einstein equations in the CFC approximation. Its main purpose is to simulate astrophysical scenarios in which strong gravity is important such as the collapse of massive stars and the evolution of neutron stars. I review recent results of the numerical code regarding neutron star oscillations and core collapse supernova and its observational consequences.

  8. 3D neutronic/thermal-hydraulic coupled analysis of MYRRHA

    SciTech Connect

    Vazquez, M.; Martin-Fuertes, F.

    2012-07-01

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

  9. Neutron Unfolding Code System for Calculating Neutron Flux Spectra from Activation Data of Dosimeter Foils.

    Energy Science and Technology Software Center (ESTSC)

    1982-04-30

    Version 00 As a part of the measurement and analysis plan for the Dosimetry Experiment at the "JOYO" experimental fast reactor, neutron flux spectral analysis is performed using the NEUPAC (Neutron Unfolding Code Package) code system. NEUPAC calculates the neutron flux spectra and other integral quantities from the activation data of the dosimeter foils.

  10. Multigroup Complex Geometry Neutron Diffusion Code System.

    Energy Science and Technology Software Center (ESTSC)

    2002-12-18

    Version 01 SNAP-3D is based on SNAP2 and is a one- two- or three-dimensional multigroup diffusion code system. It is primarily intended for neutron diffusion calculations, but it can also carry out gamma-ray calculations if the diffusion approximation is accurate enough. It is suitable for fast and thermal reactor core calculations and for shield calculations. SNAP-3D can solve the multi-group neutron diffusion equations using finite difference methods in (x,y,z), (r,theta,z), (TRI,z), (HEX,z) or (spherical) coordinates.more » The one-dimensional slab and cylindrical geometries and the two-dimensional (x,y), (r,z), (r,theta), (HEX) and (TRI) are all treated as simple special cases of three-dimensional geometries. Numerous reflective and periodic symmetry options are available and may be used to reduce the number of mesh points necessary to represent the system. Extrapolation lengths can be specified at internal and external boundaries. The problem classes are: 1) eigenvalue search for critical k-effective, 2) eigenvalue search for critical buckling, 3) eigenvalue search for critical time-constant, 4) fixed source problems in which the sources are functions of regions, 5) fixed source problems in which the sources are provided, on disc, for every mesh point and group.« less

  11. TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR.

    PubMed

    Kurosawa, Masahiko

    2005-01-01

    For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54Mn and 60Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data. PMID:16604689

  12. Calculations of the giant-dipole-resonance photoneutrons using a coupled EGS4-morse code

    SciTech Connect

    Liu, J.C.; Nelson, W.R.; Kase, K.R.; Mao, X.S.

    1995-10-01

    The production and transport of the photoneutrons from the giant-dipoleresonance reaction have been implemented in a coupled EGS4-MORSE code. The total neutron yield (including both the direct neutron and evaporation neutron components) is calculated by folding the photoneutron yield cross sections with the photon track length distribution in the target. Empirical algorithms based on the measurements have been developed to estimate the fraction and energy of the direct neutron component for each photon. The statistical theory in the EVAP4 code, incorporated as a MORSE subroutine, is used to determine the energies of the evaporation neutrons. These represent major improvements over other calculations that assumed no direct neutrons, a constant fraction of direct neutrons, monoenergetic direct neutron, or a constant nuclear temperature for the evaporation neutrons. It was also assumed that the slow neutrons (< 2.5 MeV) are emitted isotropically and the fast neutrons are emitted anisotropically in the form of 1+Csin{sup 2}{theta}, which have a peak emission at 900. Comparisons between the calculated and the measured photoneutron results (spectra of the direct, evaporation and total neutrons; nuclear temperatures; direct neutron fractions) for materials of lead, tungsten, tantalum and copper have been made. The results show that the empirical algorithms, albeit simple, can produce reasonable results over the interested photon energy range.

  13. Coupling procedure for TRANSURANUS and KTF codes

    SciTech Connect

    Jimenez, J.; Alglave, S.; Avramova, M.

    2012-07-01

    The nuclear industry aims to ensure safe and economic operation of each single fuel rod introduced in the reactor core. This goal is even more challenging nowadays due to the current strategy of going for higher burn-up (fuel cycles of 18 or 24 months) and longer residence time. In order to achieve that goal, fuel modeling is the key to predict the fuel rod behavior and lifetime under thermal and pressure loads, corrosion and irradiation. In this context, fuel performance codes, such as TRANSURANUS, are used to improve the fuel rod design. The modeling capabilities of the above mentioned tools can be significantly improved if they are coupled with a thermal-hydraulic code in order to have a better description of the flow conditions within the rod bundle. For LWR applications, a good representation of the two phase flow within the fuel assembly is necessary in order to have a best estimate calculation of the heat transfer inside the bundle. In this paper we present the coupling methodology of TRANSURANUS with KTF (Karlsruhe Two phase Flow subchannel code) as well as selected results of the coupling proof of principle. (authors)

  14. Coded source neutron imaging with a MURA mask

    NASA Astrophysics Data System (ADS)

    Zou, Y. B.; Schillinger, B.; Wang, S.; Zhang, X. S.; Guo, Z. Y.; Lu, Y. R.

    2011-09-01

    In coded source neutron imaging the single aperture commonly used in neutron radiography is replaced with a coded mask. Using a coded source can improve the neutron flux at the sample plane when a very high L/ D ratio is needed. The coded source imaging is a possible way to reduce the exposure time to get a neutron image with very high L/ D ratio. A 17×17 modified uniformly redundant array coded source was tested in this work. There are 144 holes of 0.8 mm diameter on the coded source. The neutron flux from the coded source is as high as from a single 9.6 mm aperture, while its effective L/ D is the same as in the case of a 0.8 mm aperture. The Richardson-Lucy maximum likelihood algorithm was used for image reconstruction. Compared to an in-line phase contrast neutron image taken with a 1 mm aperture, it takes much less time for the coded source to get an image of similar quality.

  15. User Guide for the R5EXEC Coupling Interface in the RELAP5-3D Code

    SciTech Connect

    Forsmann, J. Hope; Weaver, Walter L.

    2015-04-01

    This report describes the R5EXEC coupling interface in the RELAP5-3D computer code from the users perspective. The information in the report is intended for users who want to couple RELAP5-3D to other thermal-hydraulic, neutron kinetics, or control system simulation codes.

  16. Monte Carlo code for neutron scattering instrumentation design and analysis

    SciTech Connect

    Daemen, L.; Fitzsimmons, M.; Hjelm, R.; Olah, G.; Roberts, J.; Seeger, P.; Smith, G.; Thelliez, T.

    1996-09-01

    This is the final report of a one-year, Laboratory-Directed Research and Development (LDRD) at the Los Alamos National Laboratory (LANL). The development of next generation, accelerator based neutron sources calls for the design of new instruments for neutron scattering studies of materials. It will be necessary, in the near future, to evaluate accurately and rapidly the performance of new and traditional neutron instruments at short- and long-pulse spallation neutron sources, as well as continuous sources. We have developed a code that is a design tool to assist the instrument designer model new or existing instruments, test their performance, and optimize their most important features.

  17. Wake coupling to full potential rotor analysis code

    NASA Technical Reports Server (NTRS)

    Torres, Francisco J.; Chang, I-Chung; Oh, Byung K.

    1990-01-01

    The wake information from a helicopter forward flight code is coupled with two transonic potential rotor codes. The induced velocities for the near-, mid-, and far-wake geometries are extracted from a nonlinear rigid wake of a standard performance and analysis code. These, together with the corresponding inflow angles, computation points, and azimuth angles, are then incorporated into the transonic potential codes. The coupled codes can then provide an improved prediction of rotor blade loading at transonic speeds.

  18. An approach for coupled-code multiphysics core simulations from a common input

    DOE PAGESBeta

    Schmidt, Rodney; Belcourt, Kenneth; Hooper, Russell; Pawlowski, Roger P.; Clarno, Kevin T.; Simunovic, Srdjan; Slattery, Stuart R.; Turner, John A.; Palmtag, Scott

    2014-12-10

    This study describes an approach for coupled-code multiphysics reactor core simulations that is being developed by the Virtual Environment for Reactor Applications (VERA) project in the Consortium for Advanced Simulation of Light-Water Reactors (CASL). In this approach a user creates a single problem description, called the “VERAIn” common input file, to define and setup the desired coupled-code reactor core simulation. A preprocessing step accepts the VERAIn file and generates a set of fully consistent input files for the different physics codes being coupled. The problem is then solved using a single-executable coupled-code simulation tool applicable to the problem, which ismore » built using VERA infrastructure software tools and the set of physics codes required for the problem of interest. The approach is demonstrated by performing an eigenvalue and power distribution calculation of a typical three-dimensional 17 × 17 assembly with thermal–hydraulic and fuel temperature feedback. All neutronics aspects of the problem (cross-section calculation, neutron transport, power release) are solved using the Insilico code suite and are fully coupled to a thermal–hydraulic analysis calculated by the Cobra-TF (CTF) code. The single-executable coupled-code (Insilico-CTF) simulation tool is created using several VERA tools, including LIME (Lightweight Integrating Multiphysics Environment for coupling codes), DTK (Data Transfer Kit), Trilinos, and TriBITS. Parallel calculations are performed on the Titan supercomputer at Oak Ridge National Laboratory using 1156 cores, and a synopsis of the solution results and code performance is presented. Finally, ongoing development of this approach is also briefly described.« less

  19. An approach for coupled-code multiphysics core simulations from a common input

    SciTech Connect

    Schmidt, Rodney; Belcourt, Kenneth; Hooper, Russell; Pawlowski, Roger P.; Clarno, Kevin T.; Simunovic, Srdjan; Slattery, Stuart R.; Turner, John A.; Palmtag, Scott

    2014-12-10

    This study describes an approach for coupled-code multiphysics reactor core simulations that is being developed by the Virtual Environment for Reactor Applications (VERA) project in the Consortium for Advanced Simulation of Light-Water Reactors (CASL). In this approach a user creates a single problem description, called the “VERAIn” common input file, to define and setup the desired coupled-code reactor core simulation. A preprocessing step accepts the VERAIn file and generates a set of fully consistent input files for the different physics codes being coupled. The problem is then solved using a single-executable coupled-code simulation tool applicable to the problem, which is built using VERA infrastructure software tools and the set of physics codes required for the problem of interest. The approach is demonstrated by performing an eigenvalue and power distribution calculation of a typical three-dimensional 17 × 17 assembly with thermal–hydraulic and fuel temperature feedback. All neutronics aspects of the problem (cross-section calculation, neutron transport, power release) are solved using the Insilico code suite and are fully coupled to a thermal–hydraulic analysis calculated by the Cobra-TF (CTF) code. The single-executable coupled-code (Insilico-CTF) simulation tool is created using several VERA tools, including LIME (Lightweight Integrating Multiphysics Environment for coupling codes), DTK (Data Transfer Kit), Trilinos, and TriBITS. Parallel calculations are performed on the Titan supercomputer at Oak Ridge National Laboratory using 1156 cores, and a synopsis of the solution results and code performance is presented. Finally, ongoing development of this approach is also briefly described.

  20. Modular Code and Data System for Fast Reactor Neutronics Analyses

    SciTech Connect

    RIMPAULT, G.

    2008-06-30

    Version 00. The European Reactor ANalysis Optimized calculation System, ERANOS, has been developed and validated with the aim of providing a suitable basis for reliable neutronic calculations of current as well as advanced fast reactor cores. It consists of data libraries, deterministic codes and calculation procedures which have been developed within the European Collaboration on Fast Reactors over the past 20 years or so, in order to answer the needs of both industrial and R&D organisations. The whole system counts roughly 250 functions and 3000 subroutines totalling 450000 lines of FORTRAN-77 and ESOPE instructions. ERANOS is written using the ALOS software which requires only standard FORTRAN compilers and includes advanced programming features. A modular structure was adopted for easier evolution and incorporation of new functionalities. Blocks of data (SETs) can be created or used by the modules themselves or by the user via the LU control language. Programming, and dynamic memory allocation, are performed by means of the ESOPE language. External temporary storage and permanent storage capabilities are provided by the GEMAT and ARCHIVE functions, respectively. ESOPE, LU, GEMAT and ARCHIVE are all part of the ALOS software. This modular structure allows different modules to be linked together in procedures corresponding to recommended calculation routes ranging from fast-running and moderately-accurate 'routine' procedures to slow-running but highly-accurate 'reference' procedures. The main contents of the ERANOS-2.0 package are: nuclear data libraries (multigroup cross-sections from the JEF-2.2 evaluated nuclear data file, and other specific data files), a cell and lattice code (ECCO), reactor flux solvers (diffusion, Sn transport, nodal variational transport), a burn-up module, various processing modules (material and neutron balance, breeding gains,...), tools related to perturbation theory and sensitivity analysis, core follow-up modules (connected in the

  1. Modular Code and Data System for Fast Reactor Neutronics Analyses

    Energy Science and Technology Software Center (ESTSC)

    2008-06-30

    Version 00. The European Reactor ANalysis Optimized calculation System, ERANOS, has been developed and validated with the aim of providing a suitable basis for reliable neutronic calculations of current as well as advanced fast reactor cores. It consists of data libraries, deterministic codes and calculation procedures which have been developed within the European Collaboration on Fast Reactors over the past 20 years or so, in order to answer the needs of both industrial and R&Dmore » organisations. The whole system counts roughly 250 functions and 3000 subroutines totalling 450000 lines of FORTRAN-77 and ESOPE instructions. ERANOS is written using the ALOS software which requires only standard FORTRAN compilers and includes advanced programming features. A modular structure was adopted for easier evolution and incorporation of new functionalities. Blocks of data (SETs) can be created or used by the modules themselves or by the user via the LU control language. Programming, and dynamic memory allocation, are performed by means of the ESOPE language. External temporary storage and permanent storage capabilities are provided by the GEMAT and ARCHIVE functions, respectively. ESOPE, LU, GEMAT and ARCHIVE are all part of the ALOS software. This modular structure allows different modules to be linked together in procedures corresponding to recommended calculation routes ranging from fast-running and moderately-accurate 'routine' procedures to slow-running but highly-accurate 'reference' procedures. The main contents of the ERANOS-2.0 package are: nuclear data libraries (multigroup cross-sections from the JEF-2.2 evaluated nuclear data file, and other specific data files), a cell and lattice code (ECCO), reactor flux solvers (diffusion, Sn transport, nodal variational transport), a burn-up module, various processing modules (material and neutron balance, breeding gains,...), tools related to perturbation theory and sensitivity analysis, core follow-up modules (connected

  2. AMPX-77: A modular code system for generating coupled multigroup neutron-gamma cross-section libraries from ENDF/B-IV and/or ENDF/B-V

    SciTech Connect

    Greene, N.M.; Ford, W.E. III; Petrie, L.M.; Arwood, J.W.

    1992-10-01

    AMPX-77 is a modular system of computer programs that pertain to nuclear analyses, with a primary emphasis on tasks associated with the production and use of multigroup cross sections. AH basic cross-section data are to be input in the formats used by the Evaluated Nuclear Data Files (ENDF/B), and output can be obtained in a variety of formats, including its own internal and very general formats, along with a variety of other useful formats used by major transport, diffusion theory, and Monte Carlo codes. Processing is provided for both neutron and gamma-my data. The present release contains codes all written in the FORTRAN-77 dialect of FORTRAN and wig process ENDF/B-V and earlier evaluations, though major modules are being upgraded in order to process ENDF/B-VI and will be released when a complete collection of usable routines is available.

  3. Current status of the PSG Monte Carlo neutron transport code

    SciTech Connect

    Leppaenen, J.

    2006-07-01

    PSG is a new Monte Carlo neutron transport code, developed at the Technical Research Centre of Finland (VTT). The code is mainly intended for fuel assembly-level reactor physics calculations, such as group constant generation for deterministic reactor simulator codes. This paper presents the current status of the project and the essential capabilities of the code. Although the main application of PSG is in lattice calculations, the geometry is not restricted in two dimensions. This paper presents the validation of PSG against the experimental results of the three-dimensional MOX fuelled VENUS-2 reactor dosimetry benchmark. (authors)

  4. CALTRANS: A parallel, deterministic, 3D neutronics code

    SciTech Connect

    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.

  5. A COUPLED TH/NEUTRONICS/CRUD FRAMEWORK IN PREDICTION OF CIPS PHENOMENON

    SciTech Connect

    Ling Zou; Hongbin Zhang; Jess Gehin; Brendan Kochunas

    2012-04-01

    A coupled TH/Neutronics/CRUD framework, which is able to simulate the CRUD deposits impact on CIPS phenomenon, was described in this paper. This framework includes the coupling among three essential physics, thermal-hydraulics, CRUD and neutronics. The overall framework was implemented by using the CFD software STAR-CCM+, developing CRUD codes, and using the neutronics code DeCART. The coupling was implemented by exchanging data between softwares using intermediate exchange files. A typical 3 by 3 PWR fuel pin problem was solved under this framework. The problem was solved in a 12 months length period of time. Time-dependent solutions were provided, including CRUD deposits inventory and their distributions on fuels, boron hideout amount inside CRUD deposits, as well as power shape changing over time. The results clearly showed the power shape suppression in regions where CRUD deposits exist, which is a strong indication of CIPS phenomenon.

  6. Magnified Neutron Radiography with Coded Sources

    NASA Astrophysics Data System (ADS)

    Bingham, P.; Santos-Villalobos, H.; Lavrik, N.; Gregor, J.; Bilheux, H.

    A coded source imaging (CSI) system has been developed and tested at the High Flux Isotope Reactor (HFIR) CG-1D beamline at Oak Ridge National Laboratory (ORNL). The goal of this system is to use magnification to improve resolution of the imaging system beyond the detector resolution. For this system, coded masks have been manufactured at 10 μm resolution with 9 μm thick Gd patterned on Si wafers, a system model base iterative reconstruction code developed, and experiments have been performed at resolutions of 200 μm, 100 μm, 50 μm, 20 μm, and 10 μm with the object place greater than 5.5m from the detector giving magnifications up to 25 times.

  7. Coded source neutron imaging at the PULSTAR reactor

    SciTech Connect

    Xiao, Ziyu; Mishra, Kaushal; Hawari, Ayman; Bingham, Philip R; Bilheux, Hassina Z; Tobin Jr, Kenneth William

    2011-01-01

    A neutron imaging facility is located on beam-tube No.5 of the 1-MW PULSTAR reactor at North Carolina State University. An investigation of high resolution imaging using the coded source imaging technique has been initiated at the facility. Coded imaging uses a mosaic of pinholes to encode an aperture, thus generating an encoded image of the object at the detector. To reconstruct the image data received by the detector, the corresponding decoding patterns are used. The optimized design of coded mask is critical for the performance of this technique and will depend on the characteristics of the imaging beam. In this work, a 34 x 38 uniformly redundant array (URA) coded aperture system is studied for application at the PULSTAR reactor neutron imaging facility. The URA pattern was fabricated on a 500 ?m gadolinium sheet. Simulations and experiments with a pinhole object have been conducted using the Gd URA and the optimized beam line.

  8. Boron injection/dilution capabilities in TRACB/NEM coupled code

    SciTech Connect

    Jambrina, A.; Barrachina, T.; Miro, R.; Verdu, G.

    2012-07-01

    The coupled code TRAC-BF1/NEM is a thermal-hydraulic-neutronic code which allows transient simulations considering neutronic 3D and thermal-hydraulic process in multiple channels with one-dimensional geometry. TRAC-BF1 and NEM can be executed either in stand-alone mode, i.e. without coupling, as well as coupled. In stand-alone calculations NEM code is used without coupling and the thermal-hydraulic conditions (fuel temperature, moderator density and boron concentration) and xenon concentration for each node are taken from the SIMULATE3 output files. The NEM's source code has been modified to be able to read these conditions from external files when it is executed without being coupled. The coupling between TRAC-BF1 and NEM follows an integration scheme in which the thermal-hydraulic solution of TRAC-BF1 is sent to NEM to incorporate the feedback effects through the cross sections. TRAC-BF1 solves heat conduction equations inside of the heat structures using the 3D power distribution from NEM. The coupling is carried out through the communication protocol functions of PVM (Parallel Virtual Machine). The present article presents a study which constitutes an advance in the simulation of injection, transport and mix of boron in the reactor, increasing the capabilities of TRAC-BF1/NEM coupled code. This article shows the modifications introduced in the TRAC-BF1/NEM's source code to allow a more realistic simulation of boron injection transients. The qualification of these improvements in both codes is performed simulating a steady state of a generic BWR at nominal power. The results have been compared with SIMULATE3 which is used as a reference to obtain the cross sections through the SIMTAB methodology. (authors)

  9. A New Code for Proto-neutron Star Evolution

    NASA Astrophysics Data System (ADS)

    Roberts, L. F.

    2012-08-01

    A new code for following the evolution and emissions of proto-neutron stars during the first minute of their lives is developed and tested. The code is one dimensional, fully implicit, and general relativistic. Multi-group, multi-flavor neutrino transport is incorporated that makes use of variable Eddington factors obtained from a formal solution of the static general relativistic Boltzmann equation with linearized scattering terms. The timescales of neutrino emission and spectral evolution obtained using the new code are broadly consistent with previous results. Unlike other recent calculations, however, the new code predicts that the neutrino-driven wind will be characterized, at least for part of its existence, by a neutron excess. This change, potentially consequential for nucleosynthesis in the wind, is due to an improved treatment of the charged current interactions of electron-flavored neutrinos and anti-neutrinos with nucleons. A comparison is also made between the results obtained using either variable Eddington factors or simple equilibrium flux-limited diffusion. The latter approximation, which has been frequently used in previous studies of proto-neutron star cooling, accurately describes the total neutrino luminosities (to within 10%) for most of the evolution, until the proto-neutron star becomes optically thin.

  10. A NEW CODE FOR PROTO-NEUTRON STAR EVOLUTION

    SciTech Connect

    Roberts, L. F.

    2012-08-20

    A new code for following the evolution and emissions of proto-neutron stars during the first minute of their lives is developed and tested. The code is one dimensional, fully implicit, and general relativistic. Multi-group, multi-flavor neutrino transport is incorporated that makes use of variable Eddington factors obtained from a formal solution of the static general relativistic Boltzmann equation with linearized scattering terms. The timescales of neutrino emission and spectral evolution obtained using the new code are broadly consistent with previous results. Unlike other recent calculations, however, the new code predicts that the neutrino-driven wind will be characterized, at least for part of its existence, by a neutron excess. This change, potentially consequential for nucleosynthesis in the wind, is due to an improved treatment of the charged current interactions of electron-flavored neutrinos and anti-neutrinos with nucleons. A comparison is also made between the results obtained using either variable Eddington factors or simple equilibrium flux-limited diffusion. The latter approximation, which has been frequently used in previous studies of proto-neutron star cooling, accurately describes the total neutrino luminosities (to within 10%) for most of the evolution, until the proto-neutron star becomes optically thin.

  11. Study of an External Neutron Source for an Accelerator-Driven System using the PHITS Code

    SciTech Connect

    Sugawara, Takanori; Iwasaki, Tomohiko; Chiba, Takashi

    2005-05-24

    A code system for the Accelerator Driven System (ADS) has been under development for analyzing dynamic behaviors of a subcritical core coupled with an accelerator. This code system named DSE (Dynamics calculation code system for a Subcritical system with an External neutron source) consists of an accelerator part and a reactor part. The accelerator part employs a database, which is calculated by using PHITS, for investigating the effect related to the accelerator such as the changes of beam energy, beam diameter, void generation, and target level. This analysis method using the database may introduce some errors into dynamics calculations since the neutron source data derived from the database has some errors in fitting or interpolating procedures. In this study, the effects of various events are investigated to confirm that the method based on the database is appropriate.

  12. Benchmarking of Neutron Production of Heavy-Ion Transport Codes

    SciTech Connect

    Remec, Igor; Ronningen, Reginald M.; Heilbronn, Lawrence

    2012-01-01

    Accurate prediction of radiation fields generated by heavy ion interactions is important in medical applications, space missions, and in design and operation of rare isotope research facilities. In recent years, several well-established computer codes in widespread use for particle and radiation transport calculations have been equipped with the capability to simulate heavy ion transport and interactions. To assess and validate these capabilities, we performed simulations of a series of benchmark-quality heavy ion experiments with the computer codes FLUKA, MARS15, MCNPX, and PHITS. We focus on the comparisons of secondary neutron production. Results are encouraging; however, further improvements in models and codes and additional benchmarking are required.

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

    SciTech Connect

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

    2012-07-01

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

  14. Interface requirements to couple thermal hydraulics codes to severe accident codes: ICARE/CATHARE

    SciTech Connect

    Camous, F.; Jacq, F.; Chatelard, P.

    1997-07-01

    In order to describe with the same code the whole sequence of severe LWR accidents, up to the vessel failure, the Institute of Protection and Nuclear Safety has performed a coupling of the severe accident code ICARE2 to the thermalhydraulics code CATHARE2. The resulting code, ICARE/CATHARE, is designed to be as pertinent as possible in all the phases of the accident. This paper is mainly devoted to the description of the ICARE2-CATHARE2 coupling.

  15. Development of a coupling code for PWR reactor cavity radiation streaming calculation

    SciTech Connect

    Zheng, Z.; Wu, H.; Cao, L.; Zheng, Y.; Zhang, H.; Wang, M.

    2012-07-01

    PWR reactor cavity radiation streaming is important for the safe of the personnel and equipment, thus calculation has to be performed to evaluate the neutron flux distribution around the reactor. For this calculation, the deterministic codes have difficulties in fine geometrical modeling and need huge computer resource; and the Monte Carlo codes require very long sampling time to obtain results with acceptable precision. Therefore, a coupling method has been developed to eliminate the two problems mentioned above in each code. In this study, we develop a coupling code named DORT2MCNP to link the Sn code DORT and Monte Carlo code MCNP. DORT2MCNP is used to produce a combined surface source containing top, bottom and side surface simultaneously. Because SDEF card is unsuitable for the combined surface source, we modify the SOURCE subroutine of MCNP and compile MCNP for this application. Numerical results demonstrate the correctness of the coupling code DORT2MCNP and show reasonable agreement between the coupling method and the other two codes (DORT and MCNP). (authors)

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

    SciTech Connect

    Calleja, M.; Stieglitz, R.; Sanchez, V.; Jimenez, J.; Imke, U.

    2012-07-01

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

  17. 3D Multigroup Sn Neutron Transport Code

    Energy Science and Technology Software Center (ESTSC)

    2001-02-14

    ATTILA is a 3D multigroup transport code with arbitrary order ansotropic scatter. The transport equation is solved in first order form using a tri-linear discontinuous spatial differencing on an arbitrary tetrahedral mesh. The overall solution technique is source iteration with DSA acceleration of the scattering source. Anisotropic boundary and internal sources may be entered in the form of spherical harmonics moments. Alpha and k eigenvalue problems are allowed, as well as fixed source problems. Forwardmore » and adjoint solutions are available. Reflective, vacumn, and source boundary conditions are available. ATTILA can perform charged particle transport calculations using slowing down (CSD) terms. ATTILA can also be used to peform infra-red steady-state calculations for radiative transfer purposes.« less

  18. 3D Multigroup Sn Neutron Transport Code

    SciTech Connect

    McGee, John; Wareing, Todd; Pautz, Shawn

    2001-02-14

    ATTILA is a 3D multigroup transport code with arbitrary order ansotropic scatter. The transport equation is solved in first order form using a tri-linear discontinuous spatial differencing on an arbitrary tetrahedral mesh. The overall solution technique is source iteration with DSA acceleration of the scattering source. Anisotropic boundary and internal sources may be entered in the form of spherical harmonics moments. Alpha and k eigenvalue problems are allowed, as well as fixed source problems. Forward and adjoint solutions are available. Reflective, vacumn, and source boundary conditions are available. ATTILA can perform charged particle transport calculations using slowing down (CSD) terms. ATTILA can also be used to peform infra-red steady-state calculations for radiative transfer purposes.

  19. Interim report on fuel cycle neutronics code development.

    SciTech Connect

    Rabiti, C; Smith, M. A.; Kaushik, D.; Yang, W. S.

    2008-05-13

    As part of the Global Nuclear Energy Partnership (GNEP), a fast reactor simulation program was launched in April 2007 to develop a suite of modern simulation tools specifically for the analysis and design of sodium cooled fast reactors. The general goal of the new suite of codes is to reduce the uncertainties and biases in the various areas of reactor design activities by enhanced prediction capabilities. Under this fast reactor simulation program, a high-fidelity deterministic neutron transport code named UNIC is being developed. The final objective is to produce an integrated, advanced neutronics code that allows the high fidelity description of a nuclear reactor and simplifies the multi-step design process by direct coupling with thermal-hydraulics and structural mechanics calculations. Currently there are three solvers for the neutron transport code incorporated in UNIC: PN2ND, SN2ND, and MOCFE. PN2ND is based on a second-order even-parity spherical harmonics discretization of the transport equation and its primary target area of use is the existing homogenization approaches that are prevalent in reactor physics. MOCFE is based upon the method of characteristics applied to an unstructured finite element mesh and its primary target area of use is the fine grained nature of the explicit geometrical problems which is the long term goal of this project. SN2ND is based on a second-order, even-parity discrete ordinates discretization of the transport equation and its primary target area is the modeling transition region between the PN2ND and MOCFE solvers. The major development goal in fiscal year 2008 for the MOCFE solver was to include a two-dimensional capability that is scalable to hundreds of processors. The short term goal of this solver is to solve two-dimensional representations of reactor systems such that the energy and spatial self-shielding are accounted for and reliable cross sections can be generated for the homogeneous calculations. In this report we

  20. A new pad-based neutron detector for stereo coded aperture thermal neutron imaging

    NASA Astrophysics Data System (ADS)

    Dioszegi, I.; Yu, B.; Smith, G.; Schaknowski, N.; Fried, J.; Vanier, P. E.; Salwen, C.; Forman, L.

    2014-09-01

    A new coded aperture thermal neutron imager system has been developed at Brookhaven National Laboratory. The cameras use a new type of position-sensitive 3He-filled ionization chamber, in which an anode plane is composed of an array of pads with independent acquisition channels. The charge is collected on each of the individual 5x5 mm2 anode pads, (48x48 in total, corresponding to 24x24 cm2 sensitive area) and read out by application specific integrated circuits (ASICs). The new design has several advantages for coded-aperture imaging applications in the field, compared to the previous generation of wire-grid based neutron detectors. Among these are its rugged design, lighter weight and use of non-flammable stopping gas. The pad-based readout occurs in parallel circuits, making it capable of high count rates, and also suitable to perform data analysis and imaging on an event-by-event basis. The spatial resolution of the detector can be better than the pixel size by using a charge sharing algorithm. In this paper we will report on the development and performance of the new pad-based neutron camera, describe a charge sharing algorithm to achieve sub-pixel spatial resolution and present the first stereoscopic coded aperture images of thermalized neutron sources using the new coded aperture thermal neutron imager system.

  1. Bragg optics computer codes for neutron scattering instrument design

    SciTech Connect

    Popovici, M.; Yelon, W.B.; Berliner, R.R.; Stoica, A.D.

    1997-09-01

    Computer codes for neutron crystal spectrometer design, optimization and experiment planning are described. Phase space distributions, linewidths and absolute intensities are calculated by matrix methods in an extension of the Cooper-Nathans resolution function formalism. For modeling the Bragg reflection on bent crystals the lamellar approximation is used. Optimization is done by satisfying conditions of focusing in scattering and in real space, and by numerically maximizing figures of merit. Examples for three-axis and two-axis spectrometers are given.

  2. Code Development in Coupled PARCS/RELAP5 for Supercritical Water Reactor

    DOE PAGESBeta

    Hu, Po; Wilson, Paul

    2014-01-01

    The new capability is added to the existing coupled code package PARCS/RELAP5, in order to analyze SCWR design under supercritical pressure with the separated water coolant and moderator channels. This expansion is carried out on both codes. In PARCS, modification is focused on extending the water property tables to supercritical pressure, modifying the variable mapping input file and related code module for processing thermal-hydraulic information from separated coolant/moderator channels, and modifying neutronics feedback module to deal with the separated coolant/moderator channels. In RELAP5, modification is focused on incorporating more accurate water properties near SCWR operation/transient pressure and temperature in themore » code. Confirming tests of the modifications is presented and the major analyzing results from the extended codes package are summarized.« less

  3. A new code for the Hall-driven magnetic evolution of neutron stars

    NASA Astrophysics Data System (ADS)

    Viganò, D.; Pons, J. A.; Miralles, J. A.

    2012-10-01

    Over the past decade, the numerical modeling of the magnetic field evolution in astrophysical scenarios has become an increasingly important field. In the crystallized crust of neutron stars the evolution of the magnetic field is governed by the Hall induction equation. In this equation the relative contribution of the two terms (Hall term and Ohmic dissipation) varies depending on the local conditions of temperature and magnetic field strength. This results in the transition from the purely parabolic character of the equations to the hyperbolic regime as the magnetic Reynolds number increases, which presents severe numerical problems. Up to now, most attempts to study this problem were based on spectral methods, but they failed in representing the transition to large magnetic Reynolds numbers. We present a new code based on upwind finite differences techniques that can handle situations with arbitrary low magnetic diffusivity and it is suitable for studying the formation of sharp current sheets during the evolution. The code is thoroughly tested in different limits and used to illustrate the evolution of the crustal magnetic field in a neutron star in some representative cases. Our code, coupled to cooling codes, can be used to perform long-term simulations of the magneto-thermal evolution of neutron stars.

  4. Coupled receiver-decoders for low rate turbo codes

    NASA Technical Reports Server (NTRS)

    Hamkins, J.; Divsalar, D.

    2003-01-01

    We present a rate 1/31 turbo code that achieves a bit error rate of 10(sup -6) at Eb/No = -0.9 db. At this coding gain, insufficient energy per symbol is present for a conventional receiver to recover the carrier phase properly. We present a method to overcome this problem by coupling the receiver and decoder functions.

  5. Adaptation of a neutron diffraction detector to coded aperture imaging

    SciTech Connect

    Vanier, P.E.; Forman, L.

    1997-02-01

    A coded aperture neutron imaging system developed at Brookhaven National Laboratory (BNL) has demonstrated that it is possible to record not only a flux of thermal neutrons at some position, but also the directions from whence they came. This realization of an idea which defied the conventional wisdom has provided a device which has never before been available to the nuclear physics community. A number of potential applications have been explored, including (1) counting warheads on a bus or in a storage area, (2) investigating inhomogeneities in drums of Pu-containing waste to facilitate non-destructive assays, (3) monitoring of vaults containing accountable materials, (4) detection of buried land mines, and (5) locating solid deposits of nuclear material held up in gaseous diffusion plants.

  6. Neutron Limit on the Strongly-Coupled Chameleon Field

    NASA Astrophysics Data System (ADS)

    Pushin, Dmitry

    2016-03-01

    One of the major open questions of cosmology is the physical origin of the dark energy. There are a few sets of theories which might explain this origin that could be tested experimentally. The chameleon dark energy theory postulates self-interacting scalar field that couples to matter. This coupling induces a screening mechanism chosen so that the field amplitude is nonzero in empty space but is greatly suppressed in regions of terrestrial matter density. On behalf of the INDEX collaboration, I will report the most stringent upper bound on the free neutron-chameleon coupling in the strongly-coupled limit of the chameleon theory using neutron interferometric techniques. In our experiment we measure neutron phase induced by chameleon field. We report a 95 % confidence level upper bound on the neutron-chameleon coupling ranging from β < 4 . 7 ×106 for a Ratra-Peebles index of n = 1 in the nonlinear scalar field potential to β < 2 . 4 ×107 for n = 6 , one order of magnitude more sensitive than the most recent free neutron limit for intermediate n. This work was supported by NIST; NSF Grants: PHY-1205342, PHY-1068712, PHY-1307426; DOE award DE-FG02-97ER41042; NSERC CREATE and DISCOVERY programs; CERC; IUCSS and IU FRS program.

  7. Analysis of the Peach Bottom Turbine Trip 2 Experiment by Coupled RELAP5-PARCS Three-Dimensional Codes

    SciTech Connect

    Bousbia-Salah, Anis; Vedovi, Juswald; D'Auria, Francesco; Ivanov, Kostadin; Galassi, Giorgio

    2004-10-15

    Thanks to continuous progress in computer technology, it is now possible to perform best-estimate simulations of complex scenarios in nuclear power plants. This method is carried out through the coupling of three-dimensional (3-D) neutron modeling of a reactor core into system codes. It is particularly appropriate for transients that involve strong interactions between core neutronics and reactor loop thermal hydraulics. For this purpose, the Peach Bottom boiling water reactor turbine trip test was selected to challenge the capability of such coupled codes. The test is characterized by a power excursion induced by rapid core pressurization and a self-limiting course behavior. In order to perform the closest simulation, the coupled thermal-hydraulic system code RELAP5 and 3-D neutron kinetic code PARCS were used. The obtained results are compared to those available from experimental data. Overall, the coupled code calculations globally predict the most significant observed aspects of the transient, such as the pressure wave amplitude across the core and the power course, with an acceptable agreement. However, sensitivity studies revealed that more-accurate code models should be considered in order to better match the void dynamic and the cross-section variations during transient conditions.

  8. Internal neutronics-temperature coupling in Serpent 2 - Reactivity differences resulting from choice of material property correlations

    SciTech Connect

    Valtavirta, V.

    2013-07-01

    This paper describes the unique way of simultaneously solving the power and temperature distributions of a nuclear system with the Monte Carlo neutron transport code Serpent 2. The coupled solution is achieved through the implementation of an internal temperature solver and material property correlations in the code. The program structure is reviewed concerning the temperature solver and the internal correlations as well as the internal coupling between these two and the neutron transport part. To estimate the reactivity differences resulting from correlation choices a simple pin-cell case has been calculated. It is established, that some correlation choices may result in difference in reactivity of approximately 100 pcm. (authors)

  9. Coupled-Channel Computation of Direct Neutron Capture on Non-Spherical Nuclei

    NASA Astrophysics Data System (ADS)

    Arbanas, Goran; Thompson, Ian; Escher, Jutta; Nunes, Filomena; Elster, Charlotte; Zhang, Shi-Sheng

    2014-09-01

    Models of direct neutron capture of neutrons have so far accounted for the effects of non-spherical nuclei either in the incoming wave functions (via non-spherical optical model potentials), or in the final bound states (via non-spherical real potential wells), but not in both. Since it is known that spherical optical potentials do not give a good reproduction of low energy neutron-scattering observables of deformed nuclei, we have performed calculations in which the initial and final states are both treated in a self-consistent, non-spherical-nucleus picture. We have done this in the coupled-channels model of nuclear reactions implemented in the FRESCO code by using the same deformation-length for the couplings to the rotational-band states in the incoming and the final state configurations. We compute direct capture using this method for even-mass calcium isotopes 40 , 42 , 44 , 46 , 48Ca to study the effect across the two closed neutron shells, for neutron-rich even-mass tin isotopes relevant to models of astrophysical nucleosynthesis, and for 56Fe that is an important structural material used in nuclear applications. Models of direct neutron capture of neutrons have so far accounted for the effects of non-spherical nuclei either in the incoming wave functions (via non-spherical optical model potentials), or in the final bound states (via non-spherical real potential wells), but not in both. Since it is known that spherical optical potentials do not give a good reproduction of low energy neutron-scattering observables of deformed nuclei, we have performed calculations in which the initial and final states are both treated in a self-consistent, non-spherical-nucleus picture. We have done this in the coupled-channels model of nuclear reactions implemented in the FRESCO code by using the same deformation-length for the couplings to the rotational-band states in the incoming and the final state configurations. We compute direct capture using this method for even

  10. An analytical demonstration of coupling schemes between magnetohydrodynamic codes and eddy current codes

    SciTech Connect

    Liu Yueqiang

    2008-07-15

    In order to model a magnetohydrodynamic (MHD) instability that strongly couples to external conducting structures (walls and/or coils) in a fusion device, it is often necessary to combine a MHD code solving for the plasma response, with an eddy current code computing the fields and currents of conductors. We present a rigorous proof of the coupling schemes between these two types of codes. One of the coupling schemes has been introduced and implemented in the CARMA code [R. Albanese, Y. Q. Liu, A. Portone, G. Rubinacci, and F. Villone, IEEE Trans. Magn. 44, 1654 (2008); A. Portone, F. Villone, Y. Q. Liu, R. Albanese, and G. Rubinacci, Plasma Phys. Controlled Fusion 50, 085004 (2008)] that couples the MHD code MARS-F[Y. Q. Liu, A. Bondeson, C. M. Fransson, B. Lennartson, and C. Breitholtz, Phys. Plasmas 7, 3681 (2000)] and the eddy current code CARIDDI[R. Albanese and G. Rubinacci, Adv. Imaging Electron Phys. 102, 1 (1998)]. While the coupling schemes are described for a general toroidal geometry, we give the analytical proof for a cylindrical plasma.

  11. Neutron limit on the strongly-coupled chameleon field

    NASA Astrophysics Data System (ADS)

    Li, K.; Arif, M.; Cory, D. G.; Haun, R.; Heacock, B.; Huber, M. G.; Nsofini, J.; Pushin, D. A.; Saggu, P.; Sarenac, D.; Shahi, C. B.; Skavysh, V.; Snow, W. M.; Young, A. R.; Index Collaboration

    2016-03-01

    The physical origin of the dark energy that causes the accelerated expansion rate of the Universe is one of the major open questions of cosmology. One set of theories postulates the existence of a self-interacting scalar field for dark energy coupling to matter. In the chameleon dark energy theory, this coupling induces a screening mechanism such that the field amplitude is nonzero in empty space but is greatly suppressed in regions of terrestrial matter density. However measurements performed under appropriate vacuum conditions can enable the chameleon field to appear in the apparatus, where it can be subjected to laboratory experiments. Here we report the most stringent upper bound on the free neutron-chameleon coupling in the strongly coupled limit of the chameleon theory using neutron interferometric techniques. Our experiment sought the chameleon field through the relative phase shift it would induce along one of the neutron paths inside a perfect crystal neutron interferometer. The amplitude of the chameleon field was actively modulated by varying the millibar pressures inside a dual-chamber aluminum cell. We report a 95% confidence level upper bound on the neutron-chameleon coupling β ranging from β <4.7 ×106 for a Ratra-Peebles index of n =1 in the nonlinear scalar field potential to β <2.4 ×107 for n =6 , one order of magnitude more sensitive than the most recent free neutron limit for intermediate n . Similar experiments can explore the full parameter range for chameleon dark energy in the foreseeable future.

  12. Neutron-induced background in charge-coupled device detectors

    SciTech Connect

    Jaanimagi, P. A.; Boni, R.; Keck, R. L.

    2001-01-01

    The inertial confinement fusion (ICF) community must become more cognizant of the neutron-induced background levels in charge-coupled device (CCD) detectors that are replacing film as the recording medium in many ICF diagnostics. This background degrades the signal-to-noise ratio (SNR) of the recorded signals and for the highest-yield shots comprises a substantial fraction of the pixel's full well capacity. CCD detectors located anywhere in the OMEGA Target Bay are precluded from recording high precision signals (SNR>30) for deuterium--tritium neutron yields greater than 10{sup 13}. CCDs make excellent calibrated neutron detectors. The average CCD background level is proportional to the neutron yield, and we have measured a linear response over four decades. The spectrum of deposited energy per pixel is heavily weighted to low energies, <50 keV, with a few isolated saturated pixels. Most of the background recorded by the CCDs is due to secondary radiation produced by interactions of the primary neutrons with all the materials in the Target Bay as well as the shield walls and the floor. Since the noise source comes from all directions it is very difficult to shield. The fallback position of using film instead of CCD cameras for high-neutron-yield target shots is flawed, as we have observed substantially increased fog levels on our x-ray recording film as a function of the neutron yield.

  13. Neutron Activation Analysis PRognosis and Optimization Code System.

    Energy Science and Technology Software Center (ESTSC)

    2004-08-20

    Version 00 NAAPRO predicts the results and main characteristics (detection limits, determination limits, measurement limits and relative precision of the analysis) of neutron activation analysis (instrumental and radiochemical). Gamma-ray dose rates for different points of time after sample irradiation and input count rate of the spectrometry system are also predicted. The code uses standard Windows user interface and extensive graphical tools for the visualization of the spectrometer characteristics (efficiency, response and background) and simulated spectrum.more » Optimization part is not included in the current version of the code. This release is designated NAAPRO, Version 01.beta. The MCNP code was used for generating detector responses. PREPRO-2000 and FCONV programs were used at the preparation of the program nuclear databases. A special program was developed for viewing, editing and updating of the program databases (not included into the present program package). The MCNP, PREPRO-2000 and FCONV software packages are not included in the NAAPRO package.« less

  14. TOPLHA and ALOHA: comparison between Lower Hybrid wave coupling codes

    NASA Astrophysics Data System (ADS)

    Meneghini, Orso; Hillairet, J.; Goniche, M.; Bilato, R.; Voyer, D.; Parker, R.

    2008-11-01

    TOPLHA and ALOHA are wave coupling simulation tools for LH antennas. Both codes are able to account for realistic 3D antenna geometries and use a 1D plasma model. In the framework of a collaboration between MIT and CEA laboratories, the two codes have been extensively compared. In TOPLHA the EM problem is self consistently formulated by means of a set of multiple coupled integral equations having as domain the triangles of the meshed antenna surface. TOPLHA currently uses the FELHS code for modeling the plasma response. ALOHA instead uses a mode matching approach and its own plasma model. Comparisons have been done for several plasma scenarios on different antenna designs: an array of independent waveguides, a multi-junction antenna and a passive/active multi-junction antenna. When simulating the same geometry and plasma conditions the two codes compare remarkably well both for the reflection coefficients and for the launched spectra. The different approach of the two codes to solve the same problem strengthens the confidence in the final results.

  15. ALOHA: an Advanced LOwer Hybrid Antenna coupling code

    NASA Astrophysics Data System (ADS)

    Hillairet, J.; Voyer, D.; Ekedahl, A.; Goniche, M.; Kazda, M.; Meneghini, O.; Milanesio, D.; Preynas, M.

    2010-12-01

    The Advanced LOwer Hybrid Antenna (ALOHA) code, has been developed to improve the modelling of the coupling of lower hybrid (LH) waves from the antenna to a cold inhomogeneous plasma while keeping a fast tool. In contrast to the previous code Slow Wave ANtenna (SWAN) (that only described the interaction of the slow wave between the waveguides and the plasma in a 1D model), the equations are now solved in 2D including the contribution of both the slow and fast waves, with a low computational cost. This approach is completed either by a full-wave computation of the antenna that takes into account its detailed geometry or by a mode-matching code dedicated to multijunctions modelling, which is convenient in preliminary design phases. Moreover, ALOHA can treat more realistic scrape-off layers in front of the antenna, by using a two-layer electron density profile. The ALOHA code has been compared with experimental results from Tore Supra LH antennas of different geometries, as well as benchmarked against other LH coupling codes, with very good results. Once validated, ALOHA has been used as a support for the design of COMPASS and ITER LH antennas and has shown to be a fast and reliable tool for LH antenna design.

  16. Coded aperture Fast Neutron Analysis: Latest design advances

    NASA Astrophysics Data System (ADS)

    Accorsi, Roberto; Lanza, Richard C.

    2001-07-01

    Past studies have showed that materials of concern like explosives or narcotics can be identified in bulk from their atomic composition. Fast Neutron Analysis (FNA) is a nuclear method capable of providing this information even when considerable penetration is needed. Unfortunately, the cross sections of the nuclear phenomena and the solid angles involved are typically small, so that it is difficult to obtain high signal-to-noise ratios in short inspection times. CAFNAaims at combining the compound specificity of FNA with the potentially high SNR of Coded Apertures, an imaging method successfully used in far-field 2D applications. The transition to a near-field, 3D and high-energy problem prevents a straightforward application of Coded Apertures and demands a thorough optimization of the system. In this paper, the considerations involved in the design of a practical CAFNA system for contraband inspection, its conclusions, and an estimate of the performance of such a system are presented as the evolution of the ideas presented in previous expositions of the CAFNA concept.

  17. Coupled full core neutron transport/CFD simulations of pressurized water reactors

    SciTech Connect

    Kochunas, B.; Stimpson, S.; Collins, B.; Downar, T.; Brewster, R.; Baglietto, E.; Yan, J.

    2012-07-01

    Recently as part of the CASL project, a capability to perform 3D whole-core coupled neutron transport and computational fluid dynamics (CFD) calculations was demonstrated. This work uses the 2D/1D transport code DeCART and the commercial CFD code STAR-CCM+. It builds on previous CASL work demonstrating coupling for smaller spatial domains. The coupling methodology is described along with the problem simulated and results are presented for fresh hot full power conditions. An additional comparison is made to an equivalent model that uses lower order T/H feedback to assess the importance and cost of high fidelity feedback to the neutronics problem. A simulation of a quarter core Combustion Engineering (CE) PWR core was performed with the coupled codes using a Fixed Point Gauss-Seidel iteration technique. The total approximate calculation requirements are nearly 10,000 CPU hours and 1 TB of memory. The problem took 6 coupled iterations to converge. The CFD coupled model and low order T/H feedback model compared well for global solution parameters, with a difference in the critical boron concentration and average outlet temperature of 14 ppm B and 0.94 deg. C, respectively. Differences in the power distribution were more significant with maximum relative differences in the core-wide pin peaking factor (Fq) of 5.37% and average relative differences in flat flux region power of 11.54%. Future work will focus on analyzing problems more relevant to CASL using models with less approximations. (authors)

  18. A new neutron energy spectrum unfolding code using a two steps genetic algorithm

    NASA Astrophysics Data System (ADS)

    Shahabinejad, H.; Hosseini, S. A.; Sohrabpour, M.

    2016-03-01

    A new neutron spectrum unfolding code TGASU (Two-steps Genetic Algorithm Spectrum Unfolding) has been developed to unfold the neutron spectrum from a pulse height distribution which was calculated using the MCNPX-ESUT computational Monte Carlo code. To perform the unfolding process, the response matrices were generated using the MCNPX-ESUT computational code. Both one step (common GA) and two steps GAs have been implemented to unfold the neutron spectra. According to the obtained results, the new two steps GA code results has shown closer match in all energy regions and particularly in the high energy regions. The results of the TGASU code have been compared with those of the standard spectra, LSQR method and GAMCD code. The results of the TGASU code have been demonstrated to be more accurate than that of the existing computational codes for both under-determined and over-determined problems.

  19. Impact of the phonon coupling on the radiative neutron capture

    SciTech Connect

    Avdeenkov, A. V.; Goriely, S.; Kamerdzhiev, S. P.

    2010-07-15

    Inclusion of the coupling of quasiparticle degrees of freedom with phonon degrees is a natural extention of the standard QRPA approach. The paper presents the quantitative impact of this phonon coupling on the dipole strength and radiative neutron capture for the stable {sup 124}Sn and very exotic {sup 150}Sn isotopes, as an illustration, using the self-consistent version of the Extended Theory of Finite Fermi Systems. It was found that the phonon contribution to the pygmy-dipole resonance and radiative neutron capture cross section is increased with the (N - Z) difference growth. The results show that the self-consistent nuclear structure calculations are important for unstable nuclei, where phenomenological approaches do not work.

  20. Measurement of the spin-rotation coupling in neutron polarimetry

    NASA Astrophysics Data System (ADS)

    Demirel, Bülent; Sponar, Stephan; Hasegawa, Yuji

    2015-02-01

    The effect of spin-rotation coupling is measured for the first time with neutrons. The coupling of spin with the angular velocity of a rotating spin turner can be observed as a phase shift in a neutron polarimeter set-up. After the neutron’s spin is rotated by 2π through a rotating magnetic field, different phase shifts are induced for ‘up’ and ‘down’ spin eigenstates. This phase difference results in the rotation of the neutron’s spin-vector, which turns out to depend solely on the frequency of the rotation of the magnetic field. The experimental results agree well with the solutions acquired by the Pauli-Schrödinger equation.

  1. Validation of a coupled reactive transport code in porous media

    NASA Astrophysics Data System (ADS)

    Mugler, C.; Montarnal, P.; Dimier, A.

    2003-04-01

    The safety assessment of nuclear waste disposals needs to predict the migration of radionuclides and chemical species through a geological medium. It is therefore necessary to develop and assess qualified and validated tools which integrate both the transport mechanisms through the geological media and the chemical mechanisms governing the mobility of radionuclides. In this problem, both geochemical and hydrodynamic phenomena are tightly linked together. That is the reason why the French Nuclear Energy Agency (CEA) and the French Agency for the Management of Radioactive Wastes (ANDRA) are conjointly developping a coupled reactive transport code that solves simultaneously a geochemical model and a transport model. This code, which is part of the software project ALLIANCES, leans on the libraries of two geochemical codes solving the complex ensemble of reacting chemical species: CHESS and PHREEQC. Geochemical processes considered here include ion exchange, redox reactions, acid-base reactions, surface complexation and mineral dissolution and/or precipitation. Transport is simulated using the mixed-hybrid finite element scheme CAST3M or the finite volume scheme MT3D. All together solve Darcy's law and simulate several hydrological processes such as advection, diffusion and dispersion. The coupling algorithm is an iterative sequential algorithm. Several analytical test cases have been defined and used to validate the reactive transport code. Numerical results can be compared to analytical solutions.

  2. Limits on tensor coupling from neutron β decay

    NASA Astrophysics Data System (ADS)

    Pattie, R. W., Jr.; Hickerson, K. P.; Young, A. R.

    2013-10-01

    Limits on the tensor couplings generating a Fierz interference term b in mixed Gamow-Teller Fermi decays can be derived by combining data from measurements of angular correlation parameters in neutron decay, the neutron lifetime, and GV=GFVud as extracted from measurements of the Ft values from the 0+→0+ superallowed decay data set. These limits are derived by comparing the neutron β-decay rate as predicted in the standard model with the measured decay rate while allowing for the existence of beyond the standard model (BSM) couplings. We analyze limits derived from the electron-neutrino asymmetry a, or the beta asymmetry A, finding that the most stringent limits for CT/CA under the assumption of no right-handed neutrinos is -0.0026couplings have the useful property that they are independent of BSM extensions with vector or axial-vector symmetry to first order.

  3. An example of neutronic penalizations in reactivity transient analysis using 3D coupled chain HEMERA

    SciTech Connect

    Dubois, F.; Normand, B.; Sargeni, A.

    2012-07-01

    HEMERA (Highly Evolutionary Methods for Extensive Reactor Analyses), is a fully coupled 3D computational chain developed jointly by IRSN and CEA. It is composed of CRONOS2 (core neutronics, cross sections library from APOLLO2), FLICA4 (core thermal-hydraulics) and the system code CATHARE. Multi-level and multi-dimensional models are developed to account for neutronics, core thermal-hydraulics, fuel thermal analysis and system thermal-hydraulics, dedicated to best-estimate, conservative simulations and sensitivity analysis. In IRSN, the HEMERA chain is widely used to study several types of reactivity accidents and for sensitivity studies. Just as an example of the HEMERA possibilities, we present here two types of neutronic penalizations and their impact on a power transient due to a REA (Rod Ejection Accident): in the first one, we studied a bum-up distribution modification and in the second one, a delayed-neutron fraction modification. Both modifications are applied to the whole core or localized in a few assemblies. Results show that it is possible to use global or local changes but 1) in case of bum-up modification, the total core power can increase when assembly peak power decrease so, care has to be taken if the goal is to maximize a local power peak and 2) for delayed-neutron fraction, a local modification can have the same effect as the one on the whole core, provided that it is large enough. (authors)

  4. General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. We recommend C00474/ALLCP/02 MORSE-CGA.

    Energy Science and Technology Software Center (ESTSC)

    1991-08-01

    Version: 00 The original MORSE code was a multipurpose neutron and gamma-ray transport Monte Carlo code. It was designed as a tool for solving most shielding problems. Through the use of multigroup cross sections, the solution of neutron, gamma-ray, or coupled neutron-gamma-ray problems could be obtained in either the forward or adjoint mode. Time dependence for both shielding and criticality problems is provided. General three-dimensional geometry could be used with an albedo option available atmore » any material surface. Isotropic or anisotropic scattering up to a P16 expansion of the angular distribution was allowed. MORSE-CG incorporated the Mathematical Applications, Inc. (MAGI) combinatorial geometry routines. MORSE-B modifies the Monte Carlo neutron and photon transport computer code MORSE-CG by adding routines which allow various flexible options.« less

  5. Interface requirements to couple thermal-hydraulic codes to severe accident codes: ATHLET-CD

    SciTech Connect

    Trambauer, K.

    1997-07-01

    The system code ATHLET-CD is being developed by GRS in cooperation with IKE and IPSN. Its field of application comprises the whole spectrum of leaks and large breaks, as well as operational and abnormal transients for LWRs and VVERs. At present the analyses cover the in-vessel thermal-hydraulics, the early phases of core degradation, as well as fission products and aerosol release from the core and their transport in the Reactor Coolant System. The aim of the code development is to extend the simulation of core degradation up to failure of the reactor pressure vessel and to cover all physically reasonable accident sequences for western and eastern LWRs including RMBKs. The ATHLET-CD structure is highly modular in order to include a manifold spectrum of models and to offer an optimum basis for further development. The code consists of four general modules to describe the reactor coolant system thermal-hydraulics, the core degradation, the fission product core release, and fission product and aerosol transport. Each general module consists of some basic modules which correspond to the process to be simulated or to its specific purpose. Besides the code structure based on the physical modelling, the code follows four strictly separated steps during the course of a calculation: (1) input of structure, geometrical data, initial and boundary condition, (2) initialization of derived quantities, (3) steady state calculation or input of restart data, and (4) transient calculation. In this paper, the transient solution method is briefly presented and the coupling methods are discussed. Three aspects have to be considered for the coupling of different modules in one code system. First is the conservation of masses and energy in the different subsystems as there are fluid, structures, and fission products and aerosols. Second is the convergence of the numerical solution and stability of the calculation. The third aspect is related to the code performance, and running time.

  6. Cooperative solutions coupling a geometry engine and adaptive solver codes

    NASA Technical Reports Server (NTRS)

    Dickens, Thomas P.

    1995-01-01

    Follow-on work has progressed in using Aero Grid and Paneling System (AGPS), a geometry and visualization system, as a dynamic real time geometry monitor, manipulator, and interrogator for other codes. In particular, AGPS has been successfully coupled with adaptive flow solvers which iterate, refining the grid in areas of interest, and continuing on to a solution. With the coupling to the geometry engine, the new grids represent the actual geometry much more accurately since they are derived directly from the geometry and do not use refits to the first-cut grids. Additional work has been done with design runs where the geometric shape is modified to achieve a desired result. Various constraints are used to point the solution in a reasonable direction which also more closely satisfies the desired results. Concepts and techniques are presented, as well as examples of sample case studies. Issues such as distributed operation of the cooperative codes versus running all codes locally and pre-calculation for performance are discussed. Future directions are considered which will build on these techniques in light of changing computer environments.

  7. The Application Programming Interface for the PVMEXEC Program and Associated Code Coupling System

    SciTech Connect

    Walter L. Weaver III

    2005-03-01

    This report describes the Application Programming Interface for the PVMEXEC program and the code coupling systems that it implements. The information in the report is intended for programmers wanting to add a new code into the coupling system.

  8. Coupling MCNP-DSP and LAHET Monte Carlo Codes for Designing Subcriticality Monitors for Accelerator-Driven Systems

    SciTech Connect

    Valentine, T.E.; Rugama, Y. Munoz-Cobos, J.; Perez, R.

    2000-10-23

    The design of reactivity monitoring systems for accelerator-driven systems must be investigated to ensure that such systems remain subcritical during operation. The Monte Carlo codes LAHET and MCNP-DSP were combined together to facilitate the design of reactivity monitoring systems. The coupling of LAHET and MCNP-DSP provides a tool that can be used to simulate a variety of subcritical measurements such as the pulsed neutron, Rossi-{alpha}, or noise analysis measurements.

  9. A novel approach to correct the coded aperture misalignment for fast neutron imaging

    NASA Astrophysics Data System (ADS)

    Zhang, F. N.; Hu, H. S.; Zhang, T. K.; Jia, Q. G.; Wang, D. M.; Jia, J.

    2015-12-01

    Aperture alignment is crucial for the diagnosis of neutron imaging because it has significant impact on the coding imaging and the understanding of the neutron source. In our previous studies on the neutron imaging system with coded aperture for large field of view, "residual watermark," certain extra information that overlies reconstructed image and has nothing to do with the source is discovered if the peak normalization is employed in genetic algorithms (GA) to reconstruct the source image. Some studies on basic properties of residual watermark indicate that the residual watermark can characterize coded aperture and can thus be used to determine the location of coded aperture relative to the system axis. In this paper, we have further analyzed the essential conditions for the existence of residual watermark and the requirements of the reconstruction algorithm for the emergence of residual watermark. A gamma coded imaging experiment has been performed to verify the existence of residual watermark. Based on the residual watermark, a correction method for the aperture misalignment has been studied. A multiple linear regression model of the position of coded aperture axis, the position of residual watermark center, and the gray barycenter of neutron source with twenty training samples has been set up. Using the regression model and verification samples, we have found the position of the coded aperture axis relative to the system axis with an accuracy of approximately 20 μm. Conclusively, a novel approach has been established to correct the coded aperture misalignment for fast neutron coded imaging.

  10. A novel approach to correct the coded aperture misalignment for fast neutron imaging

    SciTech Connect

    Zhang, F. N.; Hu, H. S. Wang, D. M.; Jia, J.; Zhang, T. K.; Jia, Q. G.

    2015-12-15

    Aperture alignment is crucial for the diagnosis of neutron imaging because it has significant impact on the coding imaging and the understanding of the neutron source. In our previous studies on the neutron imaging system with coded aperture for large field of view, “residual watermark,” certain extra information that overlies reconstructed image and has nothing to do with the source is discovered if the peak normalization is employed in genetic algorithms (GA) to reconstruct the source image. Some studies on basic properties of residual watermark indicate that the residual watermark can characterize coded aperture and can thus be used to determine the location of coded aperture relative to the system axis. In this paper, we have further analyzed the essential conditions for the existence of residual watermark and the requirements of the reconstruction algorithm for the emergence of residual watermark. A gamma coded imaging experiment has been performed to verify the existence of residual watermark. Based on the residual watermark, a correction method for the aperture misalignment has been studied. A multiple linear regression model of the position of coded aperture axis, the position of residual watermark center, and the gray barycenter of neutron source with twenty training samples has been set up. Using the regression model and verification samples, we have found the position of the coded aperture axis relative to the system axis with an accuracy of approximately 20 μm. Conclusively, a novel approach has been established to correct the coded aperture misalignment for fast neutron coded imaging.

  11. A novel approach to correct the coded aperture misalignment for fast neutron imaging.

    PubMed

    Zhang, F N; Hu, H S; Zhang, T K; Jia, Q G; Wang, D M; Jia, J

    2015-12-01

    Aperture alignment is crucial for the diagnosis of neutron imaging because it has significant impact on the coding imaging and the understanding of the neutron source. In our previous studies on the neutron imaging system with coded aperture for large field of view, "residual watermark," certain extra information that overlies reconstructed image and has nothing to do with the source is discovered if the peak normalization is employed in genetic algorithms (GA) to reconstruct the source image. Some studies on basic properties of residual watermark indicate that the residual watermark can characterize coded aperture and can thus be used to determine the location of coded aperture relative to the system axis. In this paper, we have further analyzed the essential conditions for the existence of residual watermark and the requirements of the reconstruction algorithm for the emergence of residual watermark. A gamma coded imaging experiment has been performed to verify the existence of residual watermark. Based on the residual watermark, a correction method for the aperture misalignment has been studied. A multiple linear regression model of the position of coded aperture axis, the position of residual watermark center, and the gray barycenter of neutron source with twenty training samples has been set up. Using the regression model and verification samples, we have found the position of the coded aperture axis relative to the system axis with an accuracy of approximately 20 μm. Conclusively, a novel approach has been established to correct the coded aperture misalignment for fast neutron coded imaging. PMID:26724035

  12. FRAPTRAN Fuel Rod Code and its Coupled Transient Analysis with the GENFLO Thermal-Hydraulic Code

    SciTech Connect

    Valtonen, Keijo; Hamalainen, Anitta; Cunningham, Mitchel E. )

    2002-05-01

    The FRAPTRAN computer code has been developed for the U.S. Nuclear Regulatory Commission (NRC) to calculate fuel behavior during power and/or cooling transients at burnup levels up to 65 MWd/kgM. FRAPTRAN has now been assessed and peer reviewed. STUK/VTT have coupled GENFLO to FRAPTRAN for calculations with improved coolant boundary conditions and prepared example calculations to show the effect of improving the coolant boundary conditions.

  13. A neutron spectrum unfolding computer code based on artificial neural networks

    NASA Astrophysics Data System (ADS)

    Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

    2014-02-01

    The Bonner Spheres Spectrometer consists of a thermal neutron sensor placed at the center of a number of moderating polyethylene spheres of different diameters. From the measured readings, information can be derived about the spectrum of the neutron field where measurements were made. Disadvantages of the Bonner system are the weight associated with each sphere and the need to sequentially irradiate the spheres, requiring long exposure periods. Provided a well-established response matrix and adequate irradiation conditions, the most delicate part of neutron spectrometry, is the unfolding process. The derivation of the spectral information is not simple because the unknown is not given directly as a result of the measurements. The drawbacks associated with traditional unfolding procedures have motivated the need of complementary approaches. Novel methods based on Artificial Intelligence, mainly Artificial Neural Networks, have been widely investigated. In this work, a neutron spectrum unfolding code based on neural nets technology is presented. This code is called Neutron Spectrometry and Dosimetry with Artificial Neural networks unfolding code that was designed in a graphical interface. The core of the code is an embedded neural network architecture previously optimized using the robust design of artificial neural networks methodology. The main features of the code are: easy to use, friendly and intuitive to the user. This code was designed for a Bonner Sphere System based on a 6LiI(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. The main feature of the code is that as entrance data, for unfolding the neutron spectrum, only seven rate counts measured with seven Bonner spheres are required; simultaneously the code calculates 15 dosimetric quantities as well as the total flux for radiation protection purposes. This code generates a full report with all information of the unfolding in

  14. Neutron Transport Models and Methods for HZETRN and Coupling to Low Energy Light Ion Transport

    NASA Technical Reports Server (NTRS)

    Blattnig, S.R.; Slaba, T.C.; Heinbockel, J.H.

    2008-01-01

    Exposure estimates inside space vehicles, surface habitats, and high altitude aircraft 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 HETCHEDS and FLUKA, 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 ion (A<4) 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.

  15. Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

    SciTech Connect

    Baratta, A.J.

    1997-07-01

    To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts and engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.

  16. Rod Ejection Accident by the Coupled System Code ATHLET-QUABOX/CUBBOX

    NASA Astrophysics Data System (ADS)

    Perin, Yann; Velkov, Kiril; Pasichnyk, Igor; Langenbuch, Siegfried

    The paper considers a Rod Ejection Accident (REA) which has been calculated by the coupled-code system ATHLET-QUABOX/CUBBOX. For the present study, a MOX/UOX mixed core loading was developed on the basis of a generic PWR. The results are particularly focused on the fuel enthalpy rise which is the main safety criterion for such transient. A parametric REA study has been performed, showing the influence of some important thermal-hydraulic and neutron-physical parameters. Simulations have been performed using realistic or artificially decreased delayed neutron fractions for two different core states (HZP and 30% of the nominal power). Effective fuel rod temperature influence (i.e. Doppler coefficient) has been studied by using different correlations (0.5/0.5 weighting factors or the typical TDoppler = 0.7 TSurface + 0.3 TCenter) or by changing the fuel gap conductance. It is shown that the maximum enthalpy (and enthalpy increase) does not always appear in the affected fuel assembly but can also appear in the neighboring ones. This result is a direct consequence of the burn up dependence of the enthalpy. The paper also considers the case of local delayed neutron parameters and briefly describes the future REA studies foreseen at GRS such as an investigation of quantitative uncertainty propagation from the nuclear data to the transient behavior.

  17. Development of a coupled dynamics code with transport theory capability and application to accelerator driven systems transients

    SciTech Connect

    Cahalan, J. E.; Ama, T.; Palmiotti, G.; Taiwo, T. A.; Yang, W. S.

    2000-03-09

    The VARIANT-K and DIF3D-K nodal spatial kinetics computer codes have been coupled to the SAS4A and SASSYS-1 liquid metal reactor accident and systems analysis codes. SAS4A and SASSYS-1 have been extended with the addition of heavy liquid metal (Pb and Pb-Bi) thermophysical properties, heat transfer correlations, and fluid dynamics correlations. The coupling methodology and heavy liquid metal modeling additions are described. The new computer code suite has been applied to analysis of neutron source and thermal-hydraulics transients in a model of an accelerator-driven minor actinide burner design proposed in an OECD/NEA/NSC benchmark specification. Modeling assumptions and input data generation procedures are described. Results of transient analyses are reported, with emphasis on comparison of P1 and P3 variational nodal transport theory results with nodal diffusion theory results, and on significance of spatial kinetics effects.

  18. The simulation of a criticality accident excursion occurring in a simple fast metal system using the coupled neutronic-hydrodynamic method

    SciTech Connect

    Myers, W.L.

    1996-12-31

    Analysis of a criticality accident scenario occuring in a simple fast metal system using the coupled neutronic-hydrodynamic method is demonstrated by examining the last Godiva-I criticality accident. The basis tools and information for creating a coupled neutronic-hydrodynamic code are presented. Simplifying assumptions and approximations for creating an idealized model for the Godiva-I system are discussed. Estimates of the total energy generation and the maximum attainable kinetic energy yield are the most important results that are obtained from the code. With modifications, the methodology presented in this paper can be extended to analyze criticality accident excursions in other kinds of nuclear systems.

  19. Non-Uniform Contrast and Noise Correction for Coded Source Neutron Imaging

    SciTech Connect

    Santos-Villalobos, Hector J; Bingham, Philip R

    2012-01-01

    Since the first application of neutron radiography in the 1930s, the field of neutron radiography has matured enough to develop several applications. However, advances in the technology are far from concluded. In general, the resolution of scintillator-based detection systems is limited to the $10\\mu m$ range, and the relatively low neutron count rate of neutron sources compared to other illumination sources restricts time resolved measurement. One path toward improved resolution is the use of magnification; however, to date neutron optics are inefficient, expensive, and difficult to develop. There is a clear demand for cost-effective scintillator-based neutron imaging systems that achieve resolutions of $1 \\mu m$ or less. Such imaging system would dramatically extend the application of neutron imaging. For such purposes a coded source imaging system is under development. The current challenge is to reduce artifacts in the reconstructed coded source images. Artifacts are generated by non-uniform illumination of the source, gamma rays, dark current at the imaging sensor, and system noise from the reconstruction kernel. In this paper, we describe how to pre-process the coded signal to reduce noise and non-uniform illumination, and how to reconstruct the coded signal with three reconstruction methods correlation, maximum likelihood estimation, and algebraic reconstruction technique. We illustrates our results with experimental examples.

  20. Preliminary study of coded-source-based neutron imaging at the CPHS

    NASA Astrophysics Data System (ADS)

    Li, Yuanji; Huang, Zhifeng; Chen, Zhiqiang; Kang, Kejun; Xiao, Yongshun; Wang, Xuewu; Wei, Jie; Loong, C.-K.

    2011-09-01

    A cold neutron radiography/tomography instrument is under construction at the Compact Pulsed Hadron Source (CPHS) at Tsinghua University, China. The neutron flux is so low that an acceptable neutron radiographic image requires a long exposure time in the single-hole imaging mode. The coded-source-based imaging technique is helpful to increase the utilization of neutron flux to reduce the exposure time without loss in spatial resolution and provides high signal-to-noise ratio (SNR) images. Here we report a preliminary study on the feasibility of coded-source-based technique applied to the cold neutron imaging with a low-brilliance neutron source at the CPHS. A proper coded aperture is designed to be used in the beamline instead of the single-hole aperture. Two image retrieval algorithms, the Wiener filter algorithm and the Richardson-Lucy algorithm, are evaluated by using analytical and Monte Carlo simulations. The simulation results reveal that the coded source imaging technique is suitable for the CPHS to partially solve the problem of low neutron flux.

  1. Monte Carlo Predictions of Prompt Fission Neutrons and Photons: a Code Comparison

    NASA Astrophysics Data System (ADS)

    Talou, P.; Kawano, T.; Stetcu, I.; Vogt, R.; Randrup, J.

    2014-04-01

    This paper reports on initial comparisons between the LANL CGMF and LBNL/LLNL FREYA codes, which both aim at computing prompt fission neutrons and gammas. While the methodologies used in both codes are somewhat similar, the detailed implementations and physical assumptions are different. We are investigating how some of these differences impact predictions.

  2. Interpretation of TRIGA reactivity transients with RELAP5/PARCS coupled-code

    SciTech Connect

    Bandini, G.; Meloni, P.; Polidori, M.

    2006-07-01

    In the frame of future experiments to carried out upon TRIGA reactors, which aim to verify the real feasibility of the ADS (Accelerator Driven System) concept, it is essential to build a numerical tool able to simulate the dynamic behaviour of the reactor in subcritical configuration. This model developed to support the design of subcritical experiments and the safety analysis of the reactor, as a first step has to be assessed against the experimental data available for the critical reactor. To this purpose the thermal-hydraulic/ neutronic numerical model based on the RELAP5/PARCS coupled-code is been tested against the experimental reactivity transients conducted on the RC1-TRIGA reactor at the ENEA Casaccia Research Center in forecast of the TRADE (TRIGA Accelerator Driven Experiment) subcritical experience. The results of the calculations already performed show a qualitative good agreement with the experimental data and allow to address the future developments and improvements of the numerical model. (authors)

  3. Current and anticipated uses of thermalhydraulic and neutronic codes at PSI

    SciTech Connect

    Aksan, S.N.; Zimmermann, M.A.; Yadigaroglu, G.

    1997-07-01

    The thermalhydraulic and/or neutronic codes in use at PSI mainly provide the capability to perform deterministic safety analysis for Swiss NPPs and also serve as analysis tools for experimental facilities for LWR and ALWR simulations. In relation to these applications, physical model development and improvements, and assessment of the codes are also essential components of the activities. In this paper, a brief overview is provided on the thermalhydraulic and/or neutronic codes used for safety analysis of LWRs, at PSI, and also of some experiences and applications with these codes. Based on these experiences, additional assessment needs are indicated, together with some model improvement needs. The future needs that could be used to specify both the development of a new code and also improvement of available codes are summarized.

  4. General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System with Array Geometry Capability. We recommend C00474/ALLCP/02 MORSE-CGA.

    Energy Science and Technology Software Center (ESTSC)

    1991-05-01

    Version 00 MORSE-CGA was developed to add the capability of modelling rectangular lattices for nuclear reactor cores or for multipartitioned structures. It thus enhances the capability of the MORSE code system. The MORSE code is a multipurpose neutron and gamma-ray transport Monte Carlo code. It has been designed as a tool for solving most shielding problems. Through the use of multigroup cross sections, the solution of neutron, gamma-ray, or coupled neutron-gamma-ray problems may be obtainedmore » in either the forward or adjoint mode. Time dependence for both shielding and criticality problems is provided. General three-dimensional geometry may be used with an albedo option available at any material surface. Isotropic or anisotropic scattering up to a P16 expansion of the angular distribution is allowed.« less

  5. Thermal neutron response of a boron-coated GEM detector via GEANT4 Monte Carlo code.

    PubMed

    Jamil, M; Rhee, J T; Kim, H G; Ahmad, Farzana; Jeon, Y J

    2014-10-22

    In this work, we report the design configuration and the performance of the hybrid Gas Electron Multiplier (GEM) detector. In order to make the detector sensitive to thermal neutrons, the forward electrode of the GEM has been coated with the enriched boron-10 material, which works as a neutron converter. A total of 5×5cm(2) configuration of GEM has been used for thermal neutron studies. The response of the detector has been estimated via using GEANT4 MC code with two different physics lists. Using the QGSP_BIC_HP physics list, the neutron detection efficiency was determined to be about 3%, while with QGSP_BERT_HP physics list the efficiency was around 2.5%, at the incident thermal neutron energies of 25meV. The higher response of the detector proves that GEM-coated with boron converter improves the efficiency for thermal neutrons detection. PMID:25464183

  6. Neutron characterization study for D-T, p-7Li neutron sources with new BCA based direct collision coupling method

    NASA Astrophysics Data System (ADS)

    Wang, Guan-bo; Yang, Xin; Qian, Da-zhi; Li, Run-dong; Tang, Bin

    2014-09-01

    The T(D,n)4He and 7Li(p,n)7Be neutron sources have been used for decades in nuclear physics research, stellar nucleosynthesis research and neutron therapy research. In this work, the neutron characterization including neutron yield, spectra, and angular distribution for D-T and p-7Li sources have been studied with our new binary collision approximation (BCA) based direct collision coupling method. Distinguished from the traditional path integration method for getting the neutron weight, the new model establishes a relationship between the scattering cross section and the impact parameter, which allows the secondary neutron generation carrying out jointly with ions BCA tracking. The experimental measurements of neutron characterizations have been employed for these two reactions, and the new algorithm is validated.

  7. A theory manual for multi-physics code coupling in LIME.

    SciTech Connect

    Belcourt, Noel; Bartlett, Roscoe Ainsworth; Pawlowski, Roger Patrick; Schmidt, Rodney Cannon; Hooper, Russell Warren

    2011-03-01

    The Lightweight Integrating Multi-physics Environment (LIME) is a software package for creating multi-physics simulation codes. Its primary application space is when computer codes are currently available to solve different parts of a multi-physics problem and now need to be coupled with other such codes. In this report we define a common domain language for discussing multi-physics coupling and describe the basic theory associated with multiphysics coupling algorithms that are to be supported in LIME. We provide an assessment of coupling techniques for both steady-state and time dependent coupled systems. Example couplings are also demonstrated.

  8. CAFNA{reg{underscore}sign}, coded aperture fast neutron analysis for contraband detection: Preliminary results

    SciTech Connect

    Zhang, L.; Lanza, R.C.

    1999-12-01

    The authors have developed a near field coded aperture imaging system for use with fast neutron techniques as a tool for the detection of contraband and hidden explosives through nuclear elemental analysis. The technique relies on the prompt gamma rays produced by fast neutron interactions with the object being examined. The position of the nuclear elements is determined by the location of the gamma emitters. For existing fast neutron techniques, in Pulsed Fast Neutron Analysis (PFNA), neutrons are used with very low efficiency; in Fast Neutron Analysis (FNS), the sensitivity for detection of the signature gamma rays is very low. For the Coded Aperture Fast Neutron Analysis (CAFNA{reg{underscore}sign}) the authors have developed, the efficiency for both using the probing fast neutrons and detecting the prompt gamma rays is high. For a probed volume of n{sup 3} volume elements (voxels) in a cube of n resolution elements on a side, they can compare the sensitivity with other neutron probing techniques. As compared to PFNA, the improvement for neutron utilization is n{sup 2}, where the total number of voxels in the object being examined is n{sup 3}. Compared to FNA, the improvement for gamma-ray imaging is proportional to the total open area of the coded aperture plane; a typical value is n{sup 2}/2, where n{sup 2} is the number of total detector resolution elements or the number of pixels in an object layer. It should be noted that the actual signal to noise ratio of a system depends also on the nature and distribution of background events and this comparison may reduce somewhat the effective sensitivity of CAFNA. They have performed analysis, Monte Carlo simulations, and preliminary experiments using low and high energy gamma-ray sources. The results show that a high sensitivity 3-D contraband imaging and detection system can be realized by using CAFNA.

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

    SciTech Connect

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

    1998-03-01

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

  10. Analyses of the MSLB benchmark V1000CT-2 by the coupled system code ATHLET-BIPR8KN

    SciTech Connect

    Nikonov, S. P.; Langenbuch, S.; Lizorkin, M. P.; Velkov, K.

    2006-07-01

    Within the activities of OECD/NEA is being initiated the second phase of the VVER-1000 Coolant Transient Benchmark (V1000CT-2). It considers the best estimate analyses of a Main Steam Line Break (MSLB) of a VVER-1000 NPP with two exercises. The analyses have been performed with the coupled system code ATHLET-BIPR8KN which enables to perform realistic simulation of three-dimensional neutron kinetics and thermal-hydraulic processes in VVER NPP. Results are presented and analysed for the two proposed scenarios. These results are supplemented by sensitivity studies varying the number of the thermo-hydraulic channels (THC) in the core and by comparisons with point kinetics calculations. This work is of considerable importance for the validation of the coupled system code ATHLET-BIPR8KN in case of asymmetric core inlet conditions. (authors)

  11. A 2D/1D coupling neutron transport method based on the matrix MOC and NEM methods

    SciTech Connect

    Zhang, H.; Zheng, Y.; Wu, H.; Cao, L.

    2013-07-01

    A new 2D/1D coupling method based on the matrix MOC method (MMOC) and nodal expansion method (NEM) is proposed for solving the three-dimensional heterogeneous neutron transport problem. The MMOC method, used for radial two-dimensional calculation, constructs a response matrix between source and flux with only one sweep and then solves the linear system by using the restarted GMRES algorithm instead of the traditional trajectory sweeping process during within-group iteration for angular flux update. Long characteristics are generated by using the customization of commercial software AutoCAD. A one-dimensional diffusion calculation is carried out in the axial direction by employing the NEM method. The 2D and ID solutions are coupled through the transverse leakage items. The 3D CMFD method is used to ensure the global neutron balance and adjust the different convergence properties of the radial and axial solvers. A computational code is developed based on these theories. Two benchmarks are calculated to verify the coupling method and the code. It is observed that the corresponding numerical results agree well with references, which indicates that the new method is capable of solving the 3D heterogeneous neutron transport problem directly. (authors)

  12. Image enhancement using MCNP5 code and MATLAB in neutron radiography.

    PubMed

    Tharwat, Montaser; Mohamed, Nader; Mongy, T

    2014-07-01

    This work presents a method that can be used to enhance the neutron radiography (NR) image for objects with high scattering materials like hydrogen, carbon and other light materials. This method used Monte Carlo code, MCNP5, to simulate the NR process and get the flux distribution for each pixel of the image and determines the scattered neutron distribution that caused image blur, and then uses MATLAB to subtract this scattered neutron distribution from the initial image to improve its quality. This work was performed before the commissioning of digital NR system in Jan. 2013. The MATLAB enhancement method is quite a good technique in the case of static based film neutron radiography, while in neutron imaging (NI) technique, image enhancement and quantitative measurement were efficient by using ImageJ software. The enhanced image quality and quantitative measurements were presented in this work. PMID:24583508

  13. Coupled Receiver/Decoders for Low-Rate Turbo Codes

    NASA Technical Reports Server (NTRS)

    Hamkins, Jon; Divsalar, Dariush

    2005-01-01

    been proposed for receiving weak single- channel phase-modulated radio signals bearing low-rate-turbo-coded binary data. Originally intended for use in receiving telemetry signals from distant spacecraft, the proposed receiver/ decoders may also provide enhanced reception in mobile radiotelephone systems. A radio signal of the type to which the proposal applies comprises a residual carrier signal and a phase-modulated data signal. The residual carrier signal is needed as a phase reference for demodulation as a prerequisite to decoding. Low-rate turbo codes afford high coding gains and thereby enable the extraction of data from arriving radio signals that might otherwise be too weak. In the case of a conventional receiver, if the signal-to-noise ratio (specifically, the symbol energy to one-sided noise power spectral density) of the arriving signal is below approximately 0 dB, then there may not be enough energy per symbol to enable the receiver to recover properly the carrier phase. One could solve the problem at the transmitter by diverting some power from the data signal to the residual carrier. A better solution . a coupled receiver/decoder according to the proposal . could reduce the needed amount of residual carrier power. In all that follows, it is to be understood that all processing would be digital and the incoming signals to be processed would be, more precisely, outputs of analog-to-digital converters that preprocess the residual carrier and data signals at a rate of multiple samples per symbol. The upper part of the figure depicts a conventional receiving system, in which the receiver and decoder are uncoupled, and which is also called a non-data-aided system because output data from the decoder are not used in the receiver to aid in recovering the carrier phase. The receiver tracks the carrier phase from the residual carrier signal and uses the carrier phase to wipe phase noise off the data signal. The receiver typically includes a phase-locked loop

  14. Investigation of Coded Source Neutron Imaging at the North Carolina State University PULSTAR Reactor

    SciTech Connect

    Xiao, Ziyu; Mishra, Kaushal; Hawari, Ayman; Bingham, Philip R; Bilheux, Hassina Z; Tobin Jr, Kenneth William

    2010-10-01

    A neutron imaging facility is located on beam-tube #5 of the 1-MWth PULSTAR reactor at the North Carolina State University. An investigation has been initiated to explore the application of coded imaging techniques at the facility. Coded imaging uses a mosaic of pinholes to encode an aperture, thus generating an encoded image of the object at the detector. To reconstruct the image recorded by the detector, corresponding decoding patterns are used. The optimized design of coded masks is critical for the performance of this technique and will depend on the characteristics of the imaging beam. In this work, Monte Carlo (MCNP) simulations were utilized to explore the needed modifications to the PULSTAR thermal neutron beam to support coded imaging techniques. In addition, an assessment of coded mask design has been performed. The simulations indicated that a 12 inch single crystal sapphire filter is suited for such an application at the PULSTAR beam in terms of maximizing flux with good neutron-to-gamma ratio. Computational simulations demonstrate the feasibility of correlation reconstruction methods on neutron transmission imaging. A gadolinium aperture with thickness of 500 m was used to construct the mask using a 38 34 URA pattern. A test experiment using such a URA design has been conducted and the point spread function of the system has been measured.

  15. Development of an integrated thermal-hydraulics capability incorporating RELAP5 and PANTHER neutronics code

    SciTech Connect

    Page, R.; Jones, J.R.

    1997-07-01

    Ensuring that safety analysis needs are met in the future is likely to lead to the development of new codes and the further development of existing codes. It is therefore advantageous to define standards for data interfaces and to develop software interfacing techniques which can readily accommodate changes when they are made. Defining interface standards is beneficial but is necessarily restricted in application if future requirements are not known in detail. Code interfacing methods are of particular relevance with the move towards automatic grid frequency response operation where the integration of plant dynamic, core follow and fault study calculation tools is considered advantageous. This paper describes the background and features of a new code TALINK (Transient Analysis code LINKage program) used to provide a flexible interface to link the RELAP5 thermal hydraulics code with the PANTHER neutron kinetics and the SIBDYM whole plant dynamic modelling codes used by Nuclear Electric. The complete package enables the codes to be executed in parallel and provides an integrated whole plant thermal-hydraulics and neutron kinetics model. In addition the paper discusses the capabilities and pedigree of the component codes used to form the integrated transient analysis package and the details of the calculation of a postulated Sizewell `B` Loss of offsite power fault transient.

  16. General Purpose Monte Carlo Codes for Neutron and Photon Transport Calculations based on Continuous Energy and Multigroup Methods.

    Energy Science and Technology Software Center (ESTSC)

    2008-02-29

    Version 00 (1) Problems to be solved: MVP/GMVP II can solve eigenvalue and fixed-source problems. The multigroup code GMVP can solve forward and adjoint problems for neutron, photon and neutron-photon coupled transport. The continuous-energy code MVP can solve only the forward problems. Both codes can also perform time-dependent calculations. (2) Geometry description: MVP/GMVP employs combinatorial geometry to describe the calculation geometry. It describes spatial regions by the combination of the 3-dimensional objects (BODIes). Currently, themore » following objects (BODIes) can be used. - BODIes with linear surfaces : half space, parallelepiped, right parallelepiped, wedge, right hexagonal prism - BODIes with quadratic surface and linear surfaces : cylinder, sphere, truncated right cone, truncated elliptic cone, ellipsoid by rotation, general ellipsoid - Arbitrary quadratic surface and torus The rectangular and hexagonal lattice geometry can be used to describe the repeated geometry. Furthermore, the statistical geometry model is available to treat coated fuel particles or pebbles for high temperature reactors. (3) Particle sources: The various forms of energy-, angle-, space- and time-dependent distribution functions can be specified. See Abstract for more detail.« less

  17. Multigroup Neutron/Gamma-Ray Direct Integration Transport Code System for Two-Dimensional Cylindrical Geometry.

    Energy Science and Technology Software Center (ESTSC)

    1980-10-15

    Version 00 PALLAS-2DCY-FX is a code for direct integration of the transport equation in two-dimensional (r,z) geometry. It solves the energy and angular-dependent Boltzmann transport equation with general anisotropic scattering in cylindrical geometry. Its principal applications are to neutron or gamma-ray transport problems in the forward mode. The code is particularly designed for and suited to the solution of deep penetration radiation transport problems with an external (fixed) source.

  18. ACDOS2: an improved neutron-induced dose rate code

    SciTech Connect

    Lagache, J.C.

    1981-06-01

    To calculate the expected dose rate from fusion reactors as a function of geometry, composition, and time after shutdown a computer code, ACDOS2, was written, which utilizes up-to-date libraries of cross-sections and radioisotope decay data. ACDOS2 is in ANSI FORTRAN IV, in order to make it readily adaptable elsewhere.

  19. Neutron Activation Analysis and Product Isotope Inventory Code System.

    Energy Science and Technology Software Center (ESTSC)

    1990-10-31

    Version 00 NAC was designed to predict the neutron-induced gamma-ray radioactivity for a wide variety of composite materials. The NAC output includes the input data, a list of all reactions for each constituent element, and the end-of-irradiation disintegration rates for each reaction. NAC also compiles a product isotope inventory containing the isotope name, the disintegration rate, the gamma-ray source strength, and the absorbed dose rate at 1 meter from an unshielded point source. The inducedmore » activity is calculated as a function of irradiation and decay times; the effect of cyclic irradiation can also be calculated.« less

  20. Comparison of direct and quasi-static methods for neutron kinetic calculations with the EDF R and D COCAGNE code

    SciTech Connect

    Girardi, E.; Guerin, P.; Dulla, S.; Nervo, M.; Ravetto, P.

    2012-07-01

    Quasi-Static (QS) methods are quite popular in the reactor physics community and they exhibit two main advantages. First, these methods overcome both the limits of the Point Kinetic (PK) approach and the issues of the computational effort related to the direct discretization of the time-dependent neutron transport equation. Second, QS methods can be implemented in such a way that they can be easily coupled to very different external spatial solvers. In this paper, the results of the coupling between the QS methods developed by Politecnico di Torino and the EDF R and D core code COCAGNE are presented. The goal of these activities is to evaluate the performances of QS methods (in term of computational cost and precision) with respect to the direct kinetic solver (e.g. {theta}-scheme) already available in COCAGNE. Additionally, they allow to perform an extensive cross-validation of different kinetic models (QS and direct methods). (authors)

  1. SOURCES: a code for calculating (alpha,n), spontaneous fission, and delayed neutron sources and spectra.

    PubMed

    Wilson, W B; Perry, R T; Charlton, W S; Parish, T A; Shores, E F

    2005-01-01

    SOURCES is a computer code that determines neutron production rates and spectra from (alpha,n) reactions, spontaneous fission and delayed neutron emission owing to the decay of radionuclides in homogeneous media, interface problems and three-region interface problems. The code is also capable of calculating the neutron production rates due to (alpha,n) reactions induced by a monoenergetic beam of alpha particles incident on a slab of target material. The (alpha,n) spectra are calculated using an assumed isotropic angular distribution in the centre-of-mass system with a library of 107 nuclide decay alpha-particle spectra, 24 sets of measured and/or evaluated (alpha,n) cross sections and product nuclide level branching fractions, and functional alpha particle stopping cross sections for Z < 106. Spontaneous fission sources and spectra are calculated with evaluated half-life, spontaneous fission branching and Watt spectrum parameters for 44 actinides. The delayed neutron spectra are taken from an evaluated library of 105 precursors. The code outputs the magnitude and spectra of the resultant neutron sources. It also provides an analysis of the contributions to that source by each nuclide in the problem. PMID:16381695

  2. Neutronics-thermalhydraulics coupling in a CANDU SCWR

    NASA Astrophysics Data System (ADS)

    Adouki, Pierre

    In order to implement new nuclear technologies as a solution to the growing demand for energy, 10 countries agreed on a framework for international cooperation in 2002, to form the Generation IV International Forum (GIF). The goal of the GIF is to design the next generation of nuclear reactors that would be cost effective and would enhance safety. This forum has proposed several types of Generation IV reactors including the Supercritical Water-Cooled Reactor (SCWR). The SCWR comes in two main configurations: pressure vessel SCWR and pressure tube SCWR (PT-SCWR). In this study, the CANDU SCWR (a PT-SCWR) is considered. This reactor is oriented vertically and contains 336 channels with a length of 5 m. The target coolant inlet and outlet temperatures are 350 Celsius and 625 Celsius, respectively. The coolant flows downwards, and the reactor power is 2540 MWth. Various fuel designs have been considered in order not to exceed the linear element rating. However, the dependency between the core power and thermalhydraulics parameters results in the necessity to use a neutronics/thermalhydaulics coupling scheme to determine the core power and the thermalhydraulics parameters. The core power obtained has a power peaking factor of 1.4. The bundle power distribution for all channels has a peak at the third bundle from the inlet, but the value of this peak increases with the channel power. The heat-transfer coefficient and the specific-heat capacity have a peak at the same location in a channel, and this location shifts toward the inlet as the channel power increases. The exit coolant temperature increases with the channel power, while the exit coolant density and pressure decrease with the channel power. Also, higher channel powers lead to higher fuel and cladding temperatures. Moreover, as the coupling method is applied, the effective multiplication factor and the values of thermalhydaulics parameters oscillate as they converge.

  3. The Application of the PEBBED Code Suite to the PBMR-400 Coupled Code Benchmark - FY 2006 Annual Report

    SciTech Connect

    Not Available

    2006-09-01

    This document describes the recent developments of the PEBBED code suite and its application to the PBMR-400 Coupled Code Benchmark. This report addresses an FY2006 Level 2 milestone under the NGNP Design and Evaluation Methods Work Package. The milestone states "Complete a report describing the results of the application of the integrated PEBBED code package to the PBMR-400 coupled code benchmark". The report describes the current state of the PEBBED code suite, provides an overview of the Benchmark problems to which it was applied, discusses the code developments achieved in the past year, and states some of the results attained. Results of the steady state problems generated by the PEBBED fuel management code compare favorably to the preliminary results generated by codes from other participating institutions and to similar non-Benchmark analyses. Partial transient analysis capability has been achieved through the acquisition of the NEM-THERMIX code from Penn State University. Phase I of the task has been achieved through the development of a self-consistent set of tools for generating cross sections for design and transient analysis and in the successful execution of the steady state benchmark exercises.

  4. Neutron star structure in the presence of conformally coupled scalar fields

    NASA Astrophysics Data System (ADS)

    Sultana, Joseph; Bose, Benjamin; Kazanas, Demosthenes

    2014-10-01

    Neutron star models are studied in the context of scalar-tensor theories of gravity in the presence of a conformally coupled scalar field, using two different numerical equations of state (EoS) representing different degrees of stiffness. In both cases we obtain a complete solution by matching the interior numerical solution of the coupled Einstein-scalar field hydrostatic equations, with an exact metric on the surface of the star. These are then used to find the effect of the scalar field and its coupling to geometry, on the neutron star structure, particularly the maximum neutron star mass and radius. We show that in the presence of a conformally coupled scalar field, neutron stars are less dense and have smaller masses and radii than their counterparts in the minimally coupled case, and the effect increases with the magnitude of the scalar field at the center of the star.

  5. Neutron Star Structure in the Presence of Conformally Coupled Scalar Fields

    NASA Technical Reports Server (NTRS)

    Sultana, Joseph; Bose, Benjamin; Kazanas, Demosthenes

    2014-01-01

    Neutron star models are studied in the context of scalar-tensor theories of gravity in the presence of a conformally coupled scalar field, using two different numerical equations of state (EoS) representing different degrees of stiffness. In both cases we obtain a complete solution by matching the interior numerical solution of the coupled Einstein-scalar field hydrostatic equations, with an exact metric on the surface of the star. These are then used to find the effect of the scalar field and its coupling to geometry, on the neutron star structure, particularly the maximum neutron star mass and radius. We show that in the presence of a conformally coupled scalar field, neutron stars are less dense and have smaller masses and radii than their counterparts in the minimally coupled case, and the effect increases with the magnitude of the scalar field at the center of the star.

  6. Neutron imaging with coded sources: new challenges and the implementation of a simultaneous iterative reconstruction technique

    SciTech Connect

    Santos-Villalobos, Hector J; Bingham, Philip R; Gregor, Jens

    2013-01-01

    The limitations in neutron flux and resolution (L/D) of current neutron imaging systems can be addressed with a Coded Source Imaging system with magnification (xCSI). More precisely, the multiple sources in an xCSI system can exceed the flux of a single pinhole system for several orders of magnitude, while maintaining a higher L/D with the small sources. Moreover, designing for an xCSI system reduces noise from neutron scattering, because the object is placed away from the detector to achieve magnification. However, xCSI systems are adversely affected by correlated noise such as non-uniform illumination of the neutron source, incorrect sampling of the coded radiograph, misalignment of the coded masks, mask transparency, and the imperfection of the system Point Spread Function (PSF). We argue that a model-based reconstruction algorithm can overcome these problems and describe the implementation of a Simultaneous Iterative Reconstruction Technique algorithm for coded sources. Design pitfalls that preclude a satisfactory reconstruction are documented.

  7. Verification of a neutronic code for transient analysis in reactors with Hex-z geometry

    SciTech Connect

    Gonzalez-Pintor, S.; Verdu, G.; Ginestar, D.

    2012-07-01

    Due to the geometry of the fuel bundles, to simulate reactors such as VVER reactors it is necessary to develop methods that can deal with hexagonal prisms as basic elements of the spatial discretization. The main features of a code based on a high order finite element method for the spatial discretization of the neutron diffusion equation and an implicit difference method for the time discretization of this equation are presented and the performance of the code is tested solving the first exercise of the AER transient benchmark. The obtained results are compared with the reference results of the benchmark and with the results provided by PARCS code. (authors)

  8. ABAREX -- A neutron spherical optical-statistical-model code -- A user`s manual

    SciTech Connect

    Smith, A.B.; Lawson, R.D.

    1998-06-01

    The contemporary version of the neutron spherical optical-statistical-model code ABAREX is summarized with the objective of providing detailed operational guidance for the user. The physical concepts involved are very briefly outlined. The code is described in some detail and a number of explicit examples are given. With this document one should very quickly become fluent with the use of ABAREX. While the code has operated on a number of computing systems, this version is specifically tailored for the VAX/VMS work station and/or the IBM-compatible personal computer.

  9. Borehole parametric study for neutron induced capture gamma-ray spectrometry using the MCNP code.

    PubMed

    Shahriari, M; Sohrabpour, M

    2000-01-01

    The MCNP Monte Carlo code has been used to simulate neutron transport from an Am-Be source into a granite formation surrounding a borehole. The effects of the moisture and the neutron poison on the thermal neutron flux distribution and the capture by the absorbing elements has been calculated. Thermal and nonthermal captures for certain absorbers having resonance structures in the epithermal and fast energy regions such as W and Si were performed. It is shown that for those absorbers having large resonances in the epithermal regions when they are present in dry formation or when accompanied by neutron poisons the resonance captures may be significant compared to the thermal captures. PMID:10670932

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

    SciTech Connect

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

    2006-07-01

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

  11. A fully coupled Monte Carlo/discrete ordinates solution to the neutron transport equation. Final report

    SciTech Connect

    Filippone, W.L.; Baker, R.S.

    1990-12-31

    The neutron transport equation is solved by a hybrid method that iteratively couples regions where deterministic (S{sub N}) and stochastic (Monte Carlo) methods are applied. Unlike previous hybrid methods, the Monte Carlo and S{sub N} regions are fully coupled in the sense that no assumption is made about geometrical separation or decoupling. The hybrid method provides a new means of solving problems involving both optically thick and optically thin regions that neither Monte Carlo nor S{sub N} is well suited for by themselves. The fully coupled Monte Carlo/S{sub N} technique consists of defining spatial and/or energy regions of a problem in which either a Monte Carlo calculation or an S{sub N} calculation is to be performed. The Monte Carlo region may comprise the entire spatial region for selected energy groups, or may consist of a rectangular area that is either completely or partially embedded in an arbitrary S{sub N} region. The Monte Carlo and S{sub N} regions are then connected through the common angular boundary fluxes, which are determined iteratively using the response matrix technique, and volumetric sources. The hybrid method has been implemented in the S{sub N} code TWODANT by adding special-purpose Monte Carlo subroutines to calculate the response matrices and volumetric sources, and linkage subrountines to carry out the interface flux iterations. The common angular boundary fluxes are included in the S{sub N} code as interior boundary sources, leaving the logic for the solution of the transport flux unchanged, while, with minor modifications, the diffusion synthetic accelerator remains effective in accelerating S{sub N} calculations. The special-purpose Monte Carlo routines used are essentially analog, with few variance reduction techniques employed. However, the routines have been successfully vectorized, with approximately a factor of five increase in speed over the non-vectorized version.

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

    SciTech Connect

    Ebert, D.

    1997-07-01

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

  13. GPU-accelerated 3D neutron diffusion code based on finite difference method

    SciTech Connect

    Xu, Q.; Yu, G.; Wang, K.

    2012-07-01

    Finite difference method, as a traditional numerical solution to neutron diffusion equation, although considered simpler and more precise than the coarse mesh nodal methods, has a bottle neck to be widely applied caused by the huge memory and unendurable computation time it requires. In recent years, the concept of General-Purpose computation on GPUs has provided us with a powerful computational engine for scientific research. In this study, a GPU-Accelerated multi-group 3D neutron diffusion code based on finite difference method was developed. First, a clean-sheet neutron diffusion code (3DFD-CPU) was written in C++ on the CPU architecture, and later ported to GPUs under NVIDIA's CUDA platform (3DFD-GPU). The IAEA 3D PWR benchmark problem was calculated in the numerical test, where three different codes, including the original CPU-based sequential code, the HYPRE (High Performance Pre-conditioners)-based diffusion code and CITATION, were used as counterpoints to test the efficiency and accuracy of the GPU-based program. The results demonstrate both high efficiency and adequate accuracy of the GPU implementation for neutron diffusion equation. A speedup factor of about 46 times was obtained, using NVIDIA's Geforce GTX470 GPU card against a 2.50 GHz Intel Quad Q9300 CPU processor. Compared with the HYPRE-based code performing in parallel on an 8-core tower server, the speedup of about 2 still could be observed. More encouragingly, without any mathematical acceleration technology, the GPU implementation ran about 5 times faster than CITATION which was speeded up by using the SOR method and Chebyshev extrapolation technique. (authors)

  14. A Numerical Model for Coupling of Neutron Diffusion and Thermomechanics in Fast Burst Reactors

    SciTech Connect

    Samet Y. Kadioglu; Dana A. Knoll; Cassiano De Oliveira

    2008-11-01

    We develop a numerical model for coupling of neutron diffusion adn termomechanics in order to stimulate transient behavior of a fast burst reactor. The problem involves solving a set of non-linear different equations which approximate neutron diffusion, temperature change, and material behavior. With this equation set we will model the transition from a supercritical to subcritical state and possible mechanical vibration.

  15. Neutronics code VALE for two-dimensional triagonal (hexagonal) and three-dimensional geometries

    SciTech Connect

    Vondy, D.R.; Fowler, T.B.

    1981-08-01

    This report documents the computer code VALE designed to solve multigroup neutronics problems with the diffusion theory approximation to neutron transport for a triagonal arrangement of mesh points on planes in two- and three-dimensional geometry. This code parallels the VENTURE neutronics code in the local computation system, making exposure and fuel management capabilities available. It uses and generates interface data files adopted in the cooperative effort sponsored by Reactor Physics RRT Division of the US DOE. The programming in FORTRAN is straightforward, although data is transferred in blocks between auxiliary storage devices and main core, and direct access schemes are used. The size of problems which can be handled is essentially limited only by cost of calculation since the arrays are variably dimensioned. The memory requirement is held down while data transfer during iteration is increased only as necessary with problem size. There is provision for the more common boundary conditions including the repeating boundary, 180/sup 0/ rotational symmetry, and the rotational symmetry conditions for the 30/sup 0/, 60/sup 0/, and 120/sup 0/ triangular grids on planes. A variety of types of problems may be solved: the usual neutron flux eignevalue problem, or a direct criticality search on the buckling, on a reciprocal velocity absorber (prompt mode), or on nuclide concentrations. The adjoint problem and fixed source problem may be solved, as well as the dominating higher harmonic, or the importance problem for an arbitrary fixed source.

  16. a New Method for Neutron Capture Therapy (nct) and Related Simulation by MCNP4C Code

    NASA Astrophysics Data System (ADS)

    Shirazi, Mousavi; Alireza, Seyed; Ali, Taheri

    2010-01-01

    Neutron capture therapy (NCT) is enumerated as one of the most important methods for treatment of some strong maladies among cancers in medical science thus is unavoidable controlling and protecting instances in use of this science. Among of treatment instances of this maladies with use of nuclear medical science is use of neutron therapy that is one of the most important and effective methods in treatment of cancers. But whereas fast neutrons have too destroyer effects and also sake of protection against additional absorbed energy (absorbed dose) by tissue during neutron therapy and also naught damaging to rest of healthy tissues, should be measured absorbed energy by tissue accurately, because destroyer effects of fast neutrons is almost quintuple more than gamma photons. In this article for neutron therapy act of male's liver has been simulated a system by the Monte Carlo method (MCNP4C code) and also with use of analytical method, thus absorbed dose by this tissue has been obtained for sources with different energies accurately and has been compared results of this two methods together.

  17. Coupling external radiation transport code results to the GADRAS detector response function.

    SciTech Connect

    Mitchell, Dean J; Thoreson, Gregory G.; Horne, Steven M.

    2014-01-01

    Simulating gamma spectra is useful for analyzing special nuclear materials. Gamma spectra are influenced not only by the source and the detector, but also by the external, and potentially complex scattering environment. The scattering environment can make accurate representations of gamma spectra difficult to obtain. By coupling the Monte Carlo Nuclear Particle (MCNP) code with the Gamma Detector Response and Analysis Software (GADRAS) detector response function, gamma spectrum simulations can be computed with a high degree of fidelity even in the presence of a complex scattering environment. Traditionally, GADRAS represents the external scattering environment with empirically derived scattering parameters. By modeling the external scattering environment in MCNP and using the results as input for the GADRAS detector response function, gamma spectra can be obtained with a high degree of fidelity. This method was verified with experimental data obtained in an environment with a significant amount of scattering material. The experiment used both gamma-emitting sources and moderated and bare neutron-emitting sources. The sources were modeled using GADRAS and MCNP in the presence of the external scattering environment, producing accurate representations of the experimental data.

  18. Determination of the axial-vector weak coupling constant with ultracold neutrons.

    PubMed

    Liu, J; Mendenhall, M P; Holley, A T; Back, H O; Bowles, T J; Broussard, L J; Carr, R; Clayton, S; Currie, S; Filippone, B W; García, A; Geltenbort, P; Hickerson, K P; Hoagland, J; Hogan, G E; Hona, B; Ito, T M; Liu, C-Y; Makela, M; Mammei, R R; Martin, J W; Melconian, D; Morris, C L; Pattie, R W; Pérez Galván, A; Pitt, M L; Plaster, B; Ramsey, J C; Rios, R; Russell, R; Saunders, A; Seestrom, S J; Sondheim, W E; Tatar, E; Vogelaar, R B; VornDick, B; Wrede, C; Yan, H; Young, A R

    2010-10-29

    A precise measurement of the neutron decay β asymmetry A₀ has been carried out using polarized ultracold neutrons from the pulsed spallation ultracold neutron source at the Los Alamos Neutron Science Center. Combining data obtained in 2008 and 2009, we report A₀ = -0.119 66±0.000 89{-0.001 40}{+0.001 23}, from which we determine the ratio of the axial-vector to vector weak coupling of the nucleon g{A}/g{V}=-1.275 90{-0.004 45}{+0.004 09}. PMID:21231098

  19. VENTURE/PC manual: A multidimensional multigroup neutron diffusion code system. Version 3

    SciTech Connect

    Shapiro, A.; Huria, H.C.; Cho, K.W.

    1991-12-01

    VENTURE/PC is a recompilation of part of the Oak Ridge BOLD VENTURE code system, which will operate on an IBM PC or compatible computer. Neutron diffusion theory solutions are obtained for multidimensional, multigroup problems. This manual contains information associated with operating the code system. The purpose of the various modules used in the code system, and the input for these modules are discussed. The PC code structure is also given. Version 2 included several enhancements not given in the original version of the code. In particular, flux iterations can be done in core rather than by reading and writing to disk, for problems which allow sufficient memory for such in-core iterations. This speeds up the iteration process. Version 3 does not include any of the special processors used in the previous versions. These special processors utilized formatted input for various elements of the code system. All such input data is now entered through the Input Processor, which produces standard interface files for the various modules in the code system. In addition, a Standard Interface File Handbook is included in the documentation which is distributed with the code, to assist in developing the input for the Input Processor.

  20. VENTURE/PC manual: A multidimensional multigroup neutron diffusion code system

    SciTech Connect

    Shapiro, A.; Huria, H.C.; Cho, K.W. )

    1991-12-01

    VENTURE/PC is a recompilation of part of the Oak Ridge BOLD VENTURE code system, which will operate on an IBM PC or compatible computer. Neutron diffusion theory solutions are obtained for multidimensional, multigroup problems. This manual contains information associated with operating the code system. The purpose of the various modules used in the code system, and the input for these modules are discussed. The PC code structure is also given. Version 2 included several enhancements not given in the original version of the code. In particular, flux iterations can be done in core rather than by reading and writing to disk, for problems which allow sufficient memory for such in-core iterations. This speeds up the iteration process. Version 3 does not include any of the special processors used in the previous versions. These special processors utilized formatted input for various elements of the code system. All such input data is now entered through the Input Processor, which produces standard interface files for the various modules in the code system. In addition, a Standard Interface File Handbook is included in the documentation which is distributed with the code, to assist in developing the input for the Input Processor.

  1. DANDE: a linked code system for core neutronics/depletion analysis

    SciTech Connect

    LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.

    1985-06-01

    This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the course of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of reactor fuel under increased burn conditions. The operation of the code system is made clear in this report by following a sample problem.

  2. Crust-core coupling in rotating neutron stars

    SciTech Connect

    Glampedakis, Kostas; Andersson, Nils

    2006-08-15

    Motivated by their gravitational wave driven instability, we investigate the influence of the crust on r-mode oscillations in a neutron star. Using a simplistic model of an elastic neutron star crust with constant shear modulus, we carry out an analytic calculation with the main objective of deriving an expression for the slippage between the core and the crust. Our analytic estimates support previous numerical results and provide useful insights into the details of the problem.

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

    SciTech Connect

    Chatani, K.; Slater, C.O.

    1990-01-01

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

  4. Simulations of inspiraling and merging double neutron stars using the Spectral Einstein Code

    NASA Astrophysics Data System (ADS)

    Haas, Roland; Ott, Christian D.; Szilagyi, Bela; Kaplan, Jeffrey D.; Lippuner, Jonas; Scheel, Mark A.; Barkett, Kevin; Muhlberger, Curran D.; Dietrich, Tim; Duez, Matthew D.; Foucart, Francois; Pfeiffer, Harald P.; Kidder, Lawrence E.; Teukolsky, Saul A.

    2016-06-01

    We present results on the inspiral, merger, and postmerger evolution of a neutron star-neutron star (NSNS) system. Our results are obtained using the hybrid pseudospectral-finite volume Spectral Einstein Code (SpEC). To test our numerical methods, we evolve an equal-mass system for ≈22 orbits before merger. This waveform is the longest waveform obtained from fully general-relativistic simulations for NSNSs to date. Such long (and accurate) numerical waveforms are required to further improve semianalytical models used in gravitational wave data analysis, for example, the effective one body models. We discuss in detail the improvements to SpEC's ability to simulate NSNS mergers, in particular mesh refined grids to better resolve the merger and postmerger phases. We provide a set of consistency checks and compare our results to NSNS merger simulations with the independent bam code. We find agreement between them, which increases confidence in results obtained with either code. This work paves the way for future studies using long waveforms and more complex microphysical descriptions of neutron star matter in SpEC.

  5. RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - user`s manual

    SciTech Connect

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

    1998-03-01

    This document is the User`s Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code`s capabilities and limitations; Chapter 2 describes the code`s structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARC and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs.

  6. Assessement of Codes and Standards Applicable to a Hydrogen Production Plant Coupled to a Nuclear Reactor

    SciTech Connect

    M. J. Russell

    2006-06-01

    This is an assessment of codes and standards applicable to a hydrogen production plant to be coupled to a nuclear reactor. The result of the assessment is a list of codes and standards that are expected to be applicable to the plant during its design and construction.

  7. Evaluating the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks

    SciTech Connect

    Ortiz-Rodriguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solis Sanches, L. O.; Miranda, R. Castaneda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

    2013-07-03

    In this work the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks is evaluated. The first one code based on traditional iterative procedures and called Neutron spectrometry and dosimetry from the Universidad Autonoma de Zacatecas (NSDUAZ) use the SPUNIT iterative algorithm and was designed to unfold neutron spectrum and calculate 15 dosimetric quantities and 7 IAEA survey meters. The main feature of this code is the automated selection of the initial guess spectrum trough a compendium of neutron spectrum compiled by the IAEA. The second one code known as Neutron spectrometry and dosimetry with artificial neural networks (NDSann) is a code designed using neural nets technology. The artificial intelligence approach of neural net does not solve mathematical equations. By using the knowledge stored at synaptic weights on a neural net properly trained, the code is capable to unfold neutron spectrum and to simultaneously calculate 15 dosimetric quantities, needing as entrance data, only the rate counts measured with a Bonner spheres system. Similarities of both NSDUAZ and NSDann codes are: they follow the same easy and intuitive user's philosophy and were designed in a graphical interface under the LabVIEW programming environment. Both codes unfold the neutron spectrum expressed in 60 energy bins, calculate 15 dosimetric quantities and generate a full report in HTML format. Differences of these codes are: NSDUAZ code was designed using classical iterative approaches and needs an initial guess spectrum in order to initiate the iterative procedure. In NSDUAZ, a programming routine was designed to calculate 7 IAEA instrument survey meters using the fluence-dose conversion coefficients. NSDann code use artificial neural networks for solving the ill-conditioned equation system of neutron spectrometry problem through synaptic weights of a properly trained neural network. Contrary to iterative procedures, in neural

  8. Evaluating the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks

    NASA Astrophysics Data System (ADS)

    Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solís Sánches, L. O.; Miranda, R. Castañeda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

    2013-07-01

    In this work the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks is evaluated. The first one code based on traditional iterative procedures and called Neutron spectrometry and dosimetry from the Universidad Autonoma de Zacatecas (NSDUAZ) use the SPUNIT iterative algorithm and was designed to unfold neutron spectrum and calculate 15 dosimetric quantities and 7 IAEA survey meters. The main feature of this code is the automated selection of the initial guess spectrum trough a compendium of neutron spectrum compiled by the IAEA. The second one code known as Neutron spectrometry and dosimetry with artificial neural networks (NDSann) is a code designed using neural nets technology. The artificial intelligence approach of neural net does not solve mathematical equations. By using the knowledge stored at synaptic weights on a neural net properly trained, the code is capable to unfold neutron spectrum and to simultaneously calculate 15 dosimetric quantities, needing as entrance data, only the rate counts measured with a Bonner spheres system. Similarities of both NSDUAZ and NSDann codes are: they follow the same easy and intuitive user's philosophy and were designed in a graphical interface under the LabVIEW programming environment. Both codes unfold the neutron spectrum expressed in 60 energy bins, calculate 15 dosimetric quantities and generate a full report in HTML format. Differences of these codes are: NSDUAZ code was designed using classical iterative approaches and needs an initial guess spectrum in order to initiate the iterative procedure. In NSDUAZ, a programming routine was designed to calculate 7 IAEA instrument survey meters using the fluence-dose conversion coefficients. NSDann code use artificial neural networks for solving the ill-conditioned equation system of neutron spectrometry problem through synaptic weights of a properly trained neural network. Contrary to iterative procedures, in neural

  9. A Computer Code System for the Calculation of Reactivity and Kinetic Parameters by One-Dimensional Neutron Transport Perturbation Theory.

    Energy Science and Technology Software Center (ESTSC)

    1985-02-01

    Version 00 TP1 is a transport theory code, developed to determine reactivity effects and kinetic parameters such as effective delayed neutron fractions and mean generation time by applying the usual perturbation formalism for one-dimensional geometry.

  10. 3D Neutron Transport PWR Full-core Calculation with RMC code

    NASA Astrophysics Data System (ADS)

    Qiu, Yishu; She, Ding; Fan, Xiao; Wang, Kan; Li, Zeguang; Liang, Jingang; Leroyer, Hadrien

    2014-06-01

    Nowadays, there are more and more interests in the use of Monte Carlo codes to calculate the detailed power density distributions in full-core reactors. With the Inspur TS1000 HPC Server of Tsinghua University, several calculations have been done based on the EDF 3D Neutron Transport PWR Full-core benchmark through large-scale parallelism. To investigate and compare the results of the deterministic method and Monte Carlo method, EDF R&D and Department of Engineering Physics of Tsinghua University are having a collaboration to make code to code verification. So in this paper, two codes are used. One is the code COCAGNE developed by the EDF R&D, a deterministic core code, and the other is the Monte Carlo code RMC developed by Department of Engineering Physics in Tsinghua University. First, the full-core model is described and a 26-group calculation was performed by these two codes using the same 26-group cross-section library provided by EDF R&D. Then the parallel and tally performance of RMC is discussed. RMC employs a novel algorithm which can cut down most of the communications. It can be seen clearly that the speedup ratio almost linearly increases with the nodes. Furthermore the cell-mapping method applied by RMC consumes little time to tally even millions of cells. The results of the codes COCAGNE and RMC are compared in three ways. The results of these two codes agree well with each other. It can be concluded that both COCAGNE and RMC are able to provide 3D-transport solutions associated with detailed power density distributions calculation in PWR full-core reactors. Finally, to investigate how many histories are needed to obtain a given standard deviation for a full 3D solution, the non-symmetrized condensed 2-group fluxes of RMC are discussed.

  11. VizieR Online Data Catalog: Neutron-star matter unified EoS FORTRAN codes (Potekhin+, 2013)

    NASA Astrophysics Data System (ADS)

    Potekhin, A. Y.; Fantina, A. F.; Chamel, N.; Pearson, J. M.; Goriely, S.

    2013-09-01

    The FORTRAN codes implement the formulae for the unified EoSs of neutron-star matter BSk19, BSk20, and BSk21, supplemented with codes for masses and radii of static neutron stars, number fractions of particles, nuclear size and shape parameters, and effective masses of nucleons. Detailed descriptions of all subroutines are given by comment lines in the files (4 data files).

  12. Development and preliminary verification of the 3D core neutronic code: COCO

    SciTech Connect

    Lu, H.; Mo, K.; Li, W.; Bai, N.; Li, J.

    2012-07-01

    As the recent blooming economic growth and following environmental concerns (China)) is proactively pushing forward nuclear power development and encouraging the tapping of clean energy. Under this situation, CGNPC, as one of the largest energy enterprises in China, is planning to develop its own nuclear related technology in order to support more and more nuclear plants either under construction or being operation. This paper introduces the recent progress in software development for CGNPC. The focus is placed on the physical models and preliminary verification results during the recent development of the 3D Core Neutronic Code: COCO. In the COCO code, the non-linear Green's function method is employed to calculate the neutron flux. In order to use the discontinuity factor, the Neumann (second kind) boundary condition is utilized in the Green's function nodal method. Additionally, the COCO code also includes the necessary physical models, e.g. single-channel thermal-hydraulic module, burnup module, pin power reconstruction module and cross-section interpolation module. The preliminary verification result shows that the COCO code is sufficient for reactor core design and analysis for pressurized water reactor (PWR). (authors)

  13. Facing Challenges for Monte Carlo Analysis of Full PWR Cores : Towards Optimal Detail Level for Coupled Neutronics and Proper Diffusion Data for Nodal Kinetics

    NASA Astrophysics Data System (ADS)

    Nuttin, A.; Capellan, N.; David, S.; Doligez, X.; El Mhari, C.; Méplan, O.

    2014-06-01

    Safety analysis of innovative reactor designs requires three dimensional modeling to ensure a sufficiently realistic description, starting from steady state. Actual Monte Carlo (MC) neutron transport codes are suitable candidates to simulate large complex geometries, with eventual innovative fuel. But if local values such as power densities over small regions are needed, reliable results get more difficult to obtain within an acceptable computation time. In this scope, NEA has proposed a performance test of full PWR core calculations based on Monte Carlo neutron transport, which we have used to define an optimal detail level for convergence of steady state coupled neutronics. Coupling between MCNP for neutronics and the subchannel code COBRA for thermal-hydraulics has been performed using the C++ tool MURE, developed for about ten years at LPSC and IPNO. In parallel with this study and within the same MURE framework, a simplified code of nodal kinetics based on two-group and few-point diffusion equations has been developed and validated on a typical CANDU LOCA. Methods for the computation of necessary diffusion data have been defined and applied to NU (Nat. U) and Th fuel CANDU after assembly evolutions by MURE. Simplicity of CANDU LOCA model has made possible a comparison of these two fuel behaviours during such a transient.

  14. Automated design of coupled RF cavities using 2-D and 3-D codes

    SciTech Connect

    Smith, Peter; Christiansen, D. W.; Greninger, P. T.; Spalek, G.

    2001-01-01

    Coupled RF cavities in the Accelerator Production of Tritium Project have been designed using a procedure in which a 2-D code (CCT) searches for a design that meets frequency and coupling requirements, while a 3-D code (HFSS) is used to obtain empirical factors used by CCT to characterize the coupling slot between cavities. Using assumed values of the empirical factors, CCT runs the Superfish code iteratively to solve for a trial cavity design that has a specified frequency and coupling. The frequency shifts and the coupling constant k of the slot are modeled in CCT using a perturbation theory, the results of which are adjusted using the empirical factors. Given a trial design, HFSS is run using periodic boundary conditions to obtain a mode spectrum. The mode spectrum is processed using the DISPER code to obtain values of the coupling and the frequencies with slots. These results are used to calculate a new set of empirical factors, which are fed back into CCT for another design iteration. Cold models have been fabricated and tested to validate the codes, and results will be presented.

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

    SciTech Connect

    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.

  16. Thermal neutron imaging through XRQA2 GAFCHROMIC films coupled with a cadmium radiator

    NASA Astrophysics Data System (ADS)

    Sacco, D.; Bedogni, R.; Bortot, D.; Palomba, M.; Pola, A.; Introini, M. V.; Lorenzoli, M.; Gentile, A.; Strigari, L.; Pressello, C.; Soriani, A.; Gómez-Ros, J. M.

    2015-10-01

    A simple and inexpensive method to perform passive thermal neutron imaging on large areas was developed on the basis of XRQA2 GAFCHROMIC films, commonly employed for quality assurance in radiology. To enhance their thermal neutron response, the sensitive face of film was coupled with a 1 mm thick cadmium radiator, forming a sandwich. By exchanging the order of Cd filter and sensitive film with respect to the incident neutron beam direction, two different configurations (beam-Cd-film and beam-film-Cd) were identified. These configurations were tested at thermal neutrons fluence values in the range 109-1010 cm-2, using the ex-core radial thermal neutron column of the ENEA Casaccia - TRIGA reactor. The results are presented in this work.

  17. A new paradigm for local-global coupling in whole-core neutron transport.

    SciTech Connect

    Lewis, E.; Smith, M.; Palmiotti, G,; Nuclear Engineering Division; Northwestern Univ.; INL

    2009-01-01

    A new paradigm that increases the efficiency of whole-core neutron transport calculations without lattice homogenization is introduced. Quasi-reflected interface conditions are formulated to partially decouple periodic lattice effects from global flux gradients. The starting point is the finite subelement form of the variational nodal code VARIANT that eliminates fuel-coolant homogenization through the use of heterogeneous nodes. The interface spherical harmonics expansions that couple pin-cell-sized nodes are divided into low-order and high-order terms, and reflected interface conditions are applied to the high-order terms. Combined with an integral transport method within the node, the new approach dramatically reduces both the formation time and the dimensions of the nodal response matrices and leads to sharply reduced memory requirements and computational time. The method is applied to the two-dimensional C5G7 problem, an Organisation for Economic Co-operation and Development/Nuclear Energy Agency pressurized water reactor benchmark containing mixed oxide (MOX) and UO{sub 2} fuel assemblies, as well as to a three-dimensional MOX fuel assembly. Results indicate the new approach results in very little loss of accuracy relative to the corresponding full spherical harmonics expansions while reducing computational times by well over an order of magnitude.

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

    SciTech Connect

    Mocko, Michal; Muhrer, Guenter; Daemen, Luke L; Kelsey, Charles T; Duran, Michael A; Tovesson, Fredrik K

    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.

  19. Improved Convergence Rate of Multi-Group Scattering Moment Tallies for Monte Carlo Neutron Transport Codes

    NASA Astrophysics Data System (ADS)

    Nelson, Adam

    Multi-group scattering moment matrices are critical to the solution of the multi-group form of the neutron transport equation, as they are responsible for describing the change in direction and energy of neutrons. These matrices, however, are difficult to correctly calculate from the measured nuclear data with both deterministic and stochastic methods. Calculating these parameters when using deterministic methods requires a set of assumptions which do not hold true in all conditions. These quantities can be calculated accurately with stochastic methods, however doing so is computationally expensive due to the poor efficiency of tallying scattering moment matrices. This work presents an improved method of obtaining multi-group scattering moment matrices from a Monte Carlo neutron transport code. This improved method of tallying the scattering moment matrices is based on recognizing that all of the outgoing particle information is known a priori and can be taken advantage of to increase the tallying efficiency (therefore reducing the uncertainty) of the stochastically integrated tallies. In this scheme, the complete outgoing probability distribution is tallied, supplying every one of the scattering moment matrices elements with its share of data. In addition to reducing the uncertainty, this method allows for the use of a track-length estimation process potentially offering even further improvement to the tallying efficiency. Unfortunately, to produce the needed distributions, the probability functions themselves must undergo an integration over the outgoing energy and scattering angle dimensions. This integration is too costly to perform during the Monte Carlo simulation itself and therefore must be performed in advance by way of a pre-processing code. The new method increases the information obtained from tally events and therefore has a significantly higher efficiency than the currently used techniques. The improved method has been implemented in a code system

  20. A development of NRESPG Monte Carlo code for the calculation of neutron response function for gas counters

    NASA Astrophysics Data System (ADS)

    Takeda, N.; Kudo, K.; Toyokawa, H.; Torii, T.; Hashimoto, M.; Sugita, T.; Dietze, G.; Yang, X.

    1999-02-01

    A Monte Carlo code Neutron RESPonse function for Gas counters (NRESPG) has been developed for the calculation of neutron response functions and efficiencies for neutron energies up to 20 MeV, which can be applied for 3He, H 2, or BF 3 gas proportional counters with or without moderator. This code can simulate the neutron behavior in a two-dimensional detector configuration and treat the thermal motion of a moderator atom which becomes important as the neutron energy becomes sufficiently low. Further, a more precise measured data was taken to simulate the position-dependent gas multiplication in the sensitive and insensitive gas region of a proportional counter. The NRESPG code has been applied for the calculation of response functions of 3He cylindrical proportional counters to determine neutron energy and neutron fluence in a monoenergetic calibration field. Thus, a remarkable discrepancy in the lower portion of the full-energy peak produced by the 3He(n,p)T reaction can be removed which results in a good agreement between simulations and experiments. The code has been also used for the simulation of the response of a McTaggart-type long counter consisting of a central cylindrical BF 3 counter surrounded by a polyethylene moderator. The results of the NRESPG simulations were compared with those obtained from MCNP calculations.

  1. Search for strongly coupled Chameleon scalar field with neutron interferometry

    NASA Astrophysics Data System (ADS)

    Li, K.; Arif, M.; Cory, D.; Haun, R.; Heacock, B.; Huber, M.; Nsofini, J.; Pushin, D. A.; Saggu, P.; Sarenac, D.; Shahi, C.; Skavysh, V.; Snow, M.; Young, A.

    2015-04-01

    The dark energy proposed to explain the observed accelerated expansion of the universe is not understood. A chameleon scalar field proposed as a dark energy candidate can explain the accelerated expansion and evade all current gravity experimental bounds. It features an effective range of the chameleon scalar field that depends on the local mass density. Hence a perfect crystal neutron interferometer, that measures relative phase shift between two paths, is a prefect tool to search for the chameleon field. We are preparing a two-chamber helium gas cell for the neutron interferometer. We can lower the pressure in one cell so low that the chameleon field range expands into the cell and causes a measurable neutron phase shift while keeping the pressure difference constant. We expect to set a new upper limit of the Chameleon field by at least one order of magnitude. This work is supported by NSF Grant 1205977, DOE Grant DE-FG02-97ER41042, Canadian Excellence Research Chairs program, Natural Sciences and Engineering Research Council of Canada and Collaborative Research and Training Experience Program

  2. MCNP: a general Monte Carlo code for neutron and photon transport

    SciTech Connect

    Forster, R.A.; Godfrey, T.N.K.

    1985-01-01

    MCNP is a very general Monte Carlo neutron photon transport code system with approximately 250 person years of Group X-6 code development invested. It is extremely portable, user-oriented, and a true production code as it is used about 60 Cray hours per month by about 150 Los Alamos users. It has as its data base the best cross-section evaluations available. MCNP contains state-of-the-art traditional and adaptive Monte Carlo techniques to be applied to the solution of an ever-increasing number of problems. Excellent user-oriented documentation is available for all facets of the MCNP code system. Many useful and important variants of MCNP exist for special applications. The Radiation Shielding Information Center (RSIC) in Oak Ridge, Tennessee is the contact point for worldwide MCNP code and documentation distribution. A much improved MCNP Version 3A will be available in the fall of 1985, along with new and improved documentation. Future directions in MCNP development will change the meaning of MCNP to Monte Carlo N Particle where N particle varieties will be transported.

  3. The Transient 3-D Transport Coupled Code TORT-TD/ATTICA3D for High-Fidelity Pebble-Bed HTGR Analyses

    NASA Astrophysics Data System (ADS)

    Seubert, Armin; Sureda, Antonio; Lapins, Janis; Bader, Johannes; Laurien, Eckart

    2012-01-01

    This article describes the 3D discrete ordinates-based coupled code system TORT-TD/ATTICA3D that aims at steady state and transient analyses of pebble-bed high-temperature gas cooled reactors. In view of increasing computing power, the application of time-dependent neutron transport methods becomes feasible for best estimate evaluations of safety margins. The calculation capabilities of TORT-TD/ATTICA3D are presented along with the coupling approach, with focus on the time-dependent neutron transport features of TORT-TD. Results obtained for the OECD/NEA/NSC PBMR-400 benchmark demonstrate the transient capabilities of TORT-TD/ATTICA3D.

  4. Bruyères-le-Châtel Neutron Evaluations of Actinides with the TALYS Code: The Fission Channel

    NASA Astrophysics Data System (ADS)

    Romain, P.; Morillon, B.; Duarte, H.

    2016-01-01

    For several years, various neutron evaluations of plutonium and uranium isotopes have been performed at Bruyères-le-Châtel (BRC), from 1 keV up to 30 MeV. Since only nuclear reaction models have been used to produce these evaluations, our approach was named the "Full Model" approach. Total, shape elastic and direct inelastic cross sections were obtained from the coupled channels model using a dispersive optical potential developed for actinides, with a large enough coupling scheme including the lowest octupolar band. All other cross sections were calculated using the Hauser-Feshbach theory (TALYS code) with a pre-equilibrium component above 8-10 MeV. In this paper, we focus our attention on the fission channel. More precisely, we will present the BRC contribution to fission modeling and the philosophy adopted in our "Full Model" approach. Performing evaluations with the "Full Model" approach implies the optimization of a large number of model parameters. With increasing neutron incident energy, many residual nuclei produced by nucleon emission also lead to fission. All available experimental data assigned to various fission mechanisms of the same nucleus were used to determine fission barrier parameters. For uranium isotopes, triple-humped fission barriers were required in order to reproduce accurately variations of the experimental fission cross sections. Our BRC fission modeling has shown that the effects of the class II or class III states located in the wells of the fission barrier sometimes provide an anti-resonant transmission rather than a resonant one. Consistent evaluations were produced for a large series of U and Pu isotopes. Resulting files were tested against integral data.

  5. Quantum coupled-channels model of nuclear fusion with a semiclassical consideration of neutron rearrangement

    NASA Astrophysics Data System (ADS)

    Karpov, A. V.; Rachkov, V. A.; Samarin, V. V.

    2015-12-01

    Background: Significant enhancement of sub-barrier fusion cross sections owing to neutron transfer with positive Q values was observed in many combinations of colliding nuclei. This degree of freedom has not yet been included into the rigorous quantum coupled-channels (QCC) approach. However, the empirical coupled-channels model with neutron rearrangement [Zagrebaev, Phys. Rev. C 67, 061601 (2003), 10.1103/PhysRevC.67.061601] has already been successfully used in several papers to reproduce and predict cross sections for sub-barrier fusion reactions of stable nuclei. Purpose: The objective of this study is to combine the QCC approach and the empirical model to account for additional channels of neutron rearrangement. Method: Coupling of relative motion to collective degrees of freedom (rotation of nuclei and/or their surface vibrations) are taken into account within the QCC approach. The probability of transfer of x neutrons with a given Q value is estimated semiclassically. Results: The proposed new model was successfully tested on a few combinations of fusing nuclei 40Ca+90,96Zr, 32S+96,90, and 60,64Ni+100Mo. The calculated fusion cross sections and barrier distribution functions agree well with experimental data. Conclusions: The model developed in this work confirms all the conclusions previously made within the empirical coupled-channels model with neutron rearrangement [see Rachkov et al., Phys. Rev. C 90, 014614 (2014), 10.1103/PhysRevC.90.014614]. Moreover, it has an advantage of a more reliable microscopic account for the coupling between relative motion and the collective degrees of freedom. The proposed model can also be used to reproduce the structure of the barrier distribution function. This is a step forward to a complete solution of the long-term problem of accounting for neutron transfer channels in the QCC model.

  6. Verification of SMART Neutronics Design Methodology by the MCNAP Monte Carlo Code

    SciTech Connect

    Jong Sung Chung; Kyung Jin Shim; Chang Hyo Kim; Chungchan Lee; Sung Quun Zee

    2000-11-12

    SMART is a small advanced integral pressurized water reactor (PWR) of 330 MW(thermal) designed for both electricity generation and seawater desalinization. The CASMO-3/MASTER nuclear analysis system, a design-basis of Korean PWR plants, has been employed for the SMART core nuclear design and analysis because the fuel assembly (FA) characteristics and reactor operating conditions in temperature and pressure are similar to those of PWR plants. However, the SMART FAs are highly poisoned with more than 20 Al{sub 2}O{sub 3}-B{sub 4}C plus additional Gd{sub 2}O{sub 3}/UO{sub 2} BPRs each FA. The reactor is operated with control rods inserted. Therefore, the flux and power distribution may become more distorted than those of commercial PWR plants. In addition, SMART should produce power from room temperature to hot-power operating condition because it employs nuclear heating from room temperature. This demands reliable predictions of core criticality, shutdown margin, control rod worth, power distributions, and reactivity coefficients at both room temperature and hot operating condition, yet no such data are available to verify the CASMO-3/MASTER (hereafter MASTER) code system. In the absence of experimental verification data for the SMART neutronics design, the Monte Carlo depletion analysis program MCNAP is adopted as near-term alternatives for qualifying MASTER neutronics design calculations. The MCNAP is a personal computer-based continuous energy Monte Carlo neutronics analysis program written in C++ language. We established its qualification by presenting its prediction accuracy on measurements of Venus critical facilities and core neutronics analysis of a PWR plant in operation, and depletion characteristics of integral burnable absorber FAs of the current PWR. Here, we present a comparison of MASTER and MCNAP neutronics design calculations for SMART and establish the qualification of the MASTER system.

  7. Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data.

    Energy Science and Technology Software Center (ESTSC)

    1996-12-19

    Version 03 The NJOY nuclear data processing system is a comprehensive computer code system for producing pointwise and multigroup cross sections and related quantities from ENDF/B evaluated nuclear data in the ENDF format, including the latest US library, ENDF/B-VI. The NJOY code works with neutrons, photons, and charged particles and produces libraries for a wide variety of particle transport and reactor analysis codes.

  8. A portable, parallel, object-oriented Monte Carlo neutron transport code in C++

    SciTech Connect

    Lee, S.R.; Cummings, J.C.; Nolen, S.D. |

    1997-05-01

    We have developed a multi-group Monte Carlo neutron transport code using C++ and the Parallel Object-Oriented Methods and Applications (POOMA) class library. This transport code, called MC++, currently computes k and {alpha}-eigenvalues and is portable to and runs parallel on a wide variety of platforms, including MPPs, clustered SMPs, and individual workstations. It contains appropriate classes and abstractions for particle transport and, through the use of POOMA, for portable parallelism. Current capabilities of MC++ are discussed, along with physics and performance results on a variety of hardware, including all Accelerated Strategic Computing Initiative (ASCI) hardware. Current parallel performance indicates the ability to compute {alpha}-eigenvalues in seconds to minutes rather than hours to days. Future plans and the implementation of a general transport physics framework are also discussed.

  9. Genetic algorithms applied to reconstructing coded imaging of neutrons and analysis of residual watermark

    SciTech Connect

    Zhang Tiankui; Hu Huasi; Jia Qinggang; Zhang Fengna; Liu Zhihua; Hu Guang; Guo Wei; Chen Da; Li Zhenghong; Wu Yuelei

    2012-11-15

    Monte-Carlo simulation of neutron coded imaging based on encoding aperture for Z-pinch of large field-of-view with 5 mm radius has been investigated, and then the coded image has been obtained. Reconstruction method of source image based on genetic algorithms (GA) has been established. 'Residual watermark,' which emerges unavoidably in reconstructed image, while the peak normalization is employed in GA fitness calculation because of its statistical fluctuation amplification, has been discovered and studied. Residual watermark is primarily related to the shape and other parameters of the encoding aperture cross section. The properties and essential causes of the residual watermark were analyzed, while the identification on equivalent radius of aperture was provided. By using the equivalent radius, the reconstruction can also be accomplished without knowing the point spread function (PSF) of actual aperture. The reconstruction result is close to that by using PSF of the actual aperture.

  10. Osiris: A Modern, High-Performance, Coupled, Multi-Physics Code For Nuclear Reactor Core Analysis

    SciTech Connect

    Procassini, R J; Chand, K K; Clouse, C J; Ferencz, R M; Grandy, J M; Henshaw, W D; Kramer, K J; Parsons, I D

    2007-02-26

    To meet the simulation needs of the GNEP program, LLNL is leveraging a suite of high-performance codes to be used in the development of a multi-physics tool for modeling nuclear reactor cores. The Osiris code project, which began last summer, is employing modern computational science techniques in the development of the individual physics modules and the coupling framework. Initial development is focused on coupling thermal-hydraulics and neutral-particle transport, while later phases of the project will add thermal-structural mechanics and isotope depletion. Osiris will be applicable to the design of existing and future reactor systems through the use of first-principles, coupled physics models with fine-scale spatial resolution in three dimensions and fine-scale particle-energy resolution. Our intent is to replace an existing set of legacy, serial codes which require significant approximations and assumptions, with an integrated, coupled code that permits the design of a reactor core using a first-principles physics approach on a wide range of computing platforms, including the world's most powerful parallel computers. A key research activity of this effort deals with the efficient and scalable coupling of physics modules which utilize rather disparate mesh topologies. Our approach allows each code module to use a mesh topology and resolution that is optimal for the physics being solved, and employs a mesh-mapping and data-transfer module to effect the coupling. Additional research is planned in the area of scalable, parallel thermal-hydraulics, high-spatial-accuracy depletion and coupled-physics simulation using Monte Carlo transport.

  11. Asymptotic Analysis of Time-Dependent Neutron Transport Coupled with Isotopic Depletion and Radioactive Decay

    SciTech Connect

    Brantley, P S

    2006-09-27

    We describe an asymptotic analysis of the coupled nonlinear system of equations describing time-dependent three-dimensional monoenergetic neutron transport and isotopic depletion and radioactive decay. The classic asymptotic diffusion scaling of Larsen and Keller [1], along with a consistent small scaling of the terms describing the radioactive decay of isotopes, is applied to this coupled nonlinear system of equations in a medium of specified initial isotopic composition. The analysis demonstrates that to leading order the neutron transport equation limits to the standard time-dependent neutron diffusion equation with macroscopic cross sections whose number densities are determined by the standard system of ordinary differential equations, the so-called Bateman equations, describing the temporal evolution of the nuclide number densities.

  12. Incorporation of coupled nonequilibrium chemistry into a two-dimensional nozzle code (SEAGULL)

    NASA Technical Reports Server (NTRS)

    Ratliff, A. W.

    1979-01-01

    A two-dimensional multiple shock nozzle code (SEAGULL) was extended to include the effects of finite rate chemistry. The basic code that treats multiple shocks and contact surfaces was fully coupled with a generalized finite rate chemistry and vibrational energy exchange package. The modified code retains all of the original SEAGULL features plus the capability to treat chemical and vibrational nonequilibrium reactions. Any chemical and/or vibrational energy exchange mechanism can be handled as long as thermodynamic data and rate constants are available for all participating species.

  13. Status report on multigroup cross section generation code development for high-fidelity deterministic neutronics simulation system.

    SciTech Connect

    Yang, W. S.; Lee, C. H.

    2008-05-16

    Under the fast reactor simulation program launched in April 2007, development of an advanced multigroup cross section generation code was initiated in July 2007, in conjunction with the development of the high-fidelity deterministic neutron transport code UNIC. The general objectives are to simplify the existing multi-step schemes and to improve the resolved and unresolved resonance treatments. Based on the review results of current methods and the fact that they have been applied successfully to fast critical experiment analyses and fast reactor designs for last three decades, the methodologies of the ETOE-2/MC{sup 2}-2/SDX code system were selected as the starting set of methodologies for multigroup cross section generation for fast reactor analysis. As the first step for coupling with the UNIC code and use in a parallel computing environment, the MC{sup 2}-2 code was updated by modernizing the memory structure and replacing old data management package subroutines and functions with FORTRAN 90 based routines. Various modifications were also made in the ETOE-2 and MC{sup 2}-2 codes to process the ENDF/B-VII.0 data properly. Using the updated ETOE-2/MC{sup 2}-2 code system, the ENDF/B-VII.0 data was successfully processed for major heavy and intermediate nuclides employed in sodium-cooled fast reactors. Initial verification tests of the MC{sup 2}-2 libraries generated from ENDF/B-VII.0 data were performed by inter-comparison of twenty-one group infinite dilute total cross sections obtained from MC{sup 2}-2, VIM, and NJOY. For almost all nuclides considered, MC{sup 2}-2 cross sections agreed very well with those from VIM and NJOY. Preliminary validation tests of the ENDF/B-VII.0 libraries of MC{sup 2}-2 were also performed using a set of sixteen fast critical benchmark problems. The deterministic results based on MC{sup 2}-2/TWODANT calculations were in good agreement with MCNP solutions within {approx}0.25% {Delta}{rho}, except a few small LANL fast assemblies

  14. New code for quasiequilibrium initial data of binary neutron stars: Corotating, irrotational, and slowly spinning systems

    NASA Astrophysics Data System (ADS)

    Tsokaros, Antonios; UryÅ«, Kōji; Rezzolla, Luciano

    2015-05-01

    We present the extension of our cocal—Compact Object CALculator—code to compute general-relativistic initial data for binary compact-star systems. In particular, we construct quasiequilibrium initial data for equal-mass binaries with spins that are either aligned or antialigned with the orbital angular momentum. The Isenberg-Wilson-Mathews formalism is adopted and the constraint equations are solved using the representation formula with a suitable choice of a Green's function. We validate the new code with solutions for equal-mass binaries and explore its capabilities for a wide range of compactnesses, from a white dwarf binary with compactness ˜1 0-4, up to a highly relativistic neutron-star binary with compactness ˜0.22 . We also present a comparison with corotating and irrotational quasiequilibrium sequences from the spectral code lorene [Taniguchi and Gourgoulhon, Phys. Rev. D 66, 104019 (2002)] and with different compactness, showing that the results from the two codes agree to a precision of the order of 0.05%. Finally, we present equilibria for spinning configurations with a nuclear-physics equation of state in a piecewise polytropic representation.

  15. Model-Based Least Squares Reconstruction of Coded Source Neutron Radiographs: Integrating the ORNL HFIR CG1D Source Model

    SciTech Connect

    Santos-Villalobos, Hector J; Gregor, Jens; Bingham, Philip R

    2014-01-01

    At the present, neutron sources cannot be fabricated small and powerful enough in order to achieve high resolution radiography while maintaining an adequate flux. One solution is to employ computational imaging techniques such as a Magnified Coded Source Imaging (CSI) system. A coded-mask is placed between the neutron source and the object. The system resolution is increased by reducing the size of the mask holes and the flux is increased by increasing the size of the coded-mask and/or the number of holes. One limitation of such system is that the resolution of current state-of-the-art scintillator-based detectors caps around 50um. To overcome this challenge, the coded-mask and object are magnified by making the distance from the coded-mask to the object much smaller than the distance from object to detector. In previous work, we have shown via synthetic experiments that our least squares method outperforms other methods in image quality and reconstruction precision because of the modeling of the CSI system components. However, the validation experiments were limited to simplistic neutron sources. In this work, we aim to model the flux distribution of a real neutron source and incorporate such a model in our least squares computational system. We provide a full description of the methodology used to characterize the neutron source and validate the method with synthetic experiments.

  16. RSAP - A Code for Display of Neutron Cross Section Data and SAMMY Fit Results

    SciTech Connect

    Sayer, R.O.

    2001-02-02

    RSAP is a computer code for display of neutron cross section data and selected SAMMY output. SAMMY is a multilevel R-matrix code for fitting neutron time-of-flight cross-section data using Bayes' method. RSAP, which runs on the Digital Unix Alpha platform, reads ORELA Data Files (ODF) created by SAMMY and uses graphics routines from the PLPLOT package. In addition, RSAP can read data and/or computed values from ASCII files with a format specified by the user. Plot output may be displayed in an X window, sent to a postscript file (rsap.ps), or sent to a color postscript file (rsap.psc). Thirteen plot types are supported, allowing the user to display cross section data, transmission data, errors, theory, Bayes fits, and residuals in various combinations. In this document the designations theory and Bayes refer to the initial and final theoretical cross sections, respectively, as evaluated by SAMMY. Special plot types include Bayes/Data, Theory--Data, and Bayes--Data. Output from two SAMMY runs may be compared by plotting the ratios Theory2/Theory1 and Bayes2/Bayes1 or by plotting the differences (Theory2-Theory1) and (Bayes2-Bayes1).

  17. Validation of the MCNPX-PoliMi Code to Design a Fast-Neutron Multiplicity Counter

    SciTech Connect

    J. L. Dolan; A. C. Kaplan; M. Flaska; S. A. Pozzi; D. L. Chichester

    2012-07-01

    Many safeguards measurement systems used at nuclear facilities, both domestically and internationally, rely on He-3 detectors and well established mathematical equations to interpret coincidence and multiplicity-type measurements for verifying quantities of special nuclear material. Due to resource shortages alternatives to these existing He-3 based systems are being sought. Work is also underway to broaden the capabilities of these types of measurement systems in order to improve current multiplicity analysis techniques. As a part of a Material Protection, Accounting, and Control Technology (MPACT) project within the U.S. Department of Energy's Fuel Cycle Technology Program we are designing a fast-neutron multiplicity counter with organic liquid scintillators to quantify important quantities such as plutonium mass. We are also examining the potential benefits of using fast-neutron detectors for multiplicity analysis of advanced fuels in comparison with He-3 detectors and testing the performance of such designs. The designs are being developed and optimized using the MCNPX-PoliMi transport code to study detector response. In the full paper, we will discuss validation measurements used to justify the use of the MCNPX-PoliMi code paired with the MPPost multiplicity routine to design a fast neutron multiplicity counter with liquid scintillators. This multiplicity counter will be designed with the end goal of safeguarding advanced nuclear fuels. With improved timing qualities associated with liquid scintillation detectors, we can design a system that is less limited by nuclear materials of high activities. Initial testing of the designed system with nuclear fuels will take place at Idaho National Laboratory in a later stage of this collaboration.

  18. NSDann2BS, a neutron spectrum unfolding code based on neural networks technology and two bonner spheres

    NASA Astrophysics Data System (ADS)

    Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solís Sánches, L. O.; Miranda, R. Castañeda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

    2013-07-01

    In this work a neutron spectrum unfolding code, based on artificial intelligence technology is presented. The code called "Neutron Spectrometry and Dosimetry with Artificial Neural Networks and two Bonner spheres", (NSDann2BS), was designed in a graphical user interface under the LabVIEW programming environment. The main features of this code are to use an embedded artificial neural network architecture optimized with the "Robust design of artificial neural networks methodology" and to use two Bonner spheres as the only piece of information. In order to build the code here presented, once the net topology was optimized and properly trained, knowledge stored at synaptic weights was extracted and using a graphical framework build on the LabVIEW programming environment, the NSDann2BS code was designed. This code is friendly, intuitive and easy to use for the end user. The code is freely available upon request to authors. To demonstrate the use of the neural net embedded in the NSDann2BS code, the rate counts of 252Cf, 241AmBe and 239PuBe neutron sources measured with a Bonner spheres system.

  19. NSDann2BS, a neutron spectrum unfolding code based on neural networks technology and two bonner spheres

    SciTech Connect

    Ortiz-Rodriguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solis Sanches, L. O.; Miranda, R. Castaneda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

    2013-07-03

    In this work a neutron spectrum unfolding code, based on artificial intelligence technology is presented. The code called ''Neutron Spectrometry and Dosimetry with Artificial Neural Networks and two Bonner spheres'', (NSDann2BS), was designed in a graphical user interface under the LabVIEW programming environment. The main features of this code are to use an embedded artificial neural network architecture optimized with the ''Robust design of artificial neural networks methodology'' and to use two Bonner spheres as the only piece of information. In order to build the code here presented, once the net topology was optimized and properly trained, knowledge stored at synaptic weights was extracted and using a graphical framework build on the LabVIEW programming environment, the NSDann2BS code was designed. This code is friendly, intuitive and easy to use for the end user. The code is freely available upon request to authors. To demonstrate the use of the neural net embedded in the NSDann2BS code, the rate counts of {sup 252}Cf, {sup 241}AmBe and {sup 239}PuBe neutron sources measured with a Bonner spheres system.

  20. A three-dimensional transient neutronics routine for the TRAC-PF1 reactor thermal hydraulic computer code

    SciTech Connect

    Bandini, B.R. Los Alamos National Lab., NM

    1990-05-01

    No present light water reactor accident analysis code employs both high state of the art neutronics and thermal-hydraulics computational algorithms. Adding a modern three-dimensional neutron kinetics model to the present TRAC-PFI/MOD2 code would create a fully up to date pressurized water reactor accident evaluation code. After reviewing several options, it was decided that the Nodal Expansion Method would best provide the basis for this multidimensional transient neutronic analysis capability. Steady-state and transient versions of the Nodal Expansion Method were coded in both three-dimensional Cartesian and cylindrical geometries. In stand-alone form this method of solving the few group neutron diffusion equations was shown to yield efficient and accurate results for a variety of steady-state and transient benchmark problems. The Nodal Expansion Method was then incorporated into TRAC-PFl/MOD2. The combined NEM/TRAC code results agreed well with the EPRI-ARROTTA core-only transient analysis code when modelling a severe PWR control rod ejection accident.

  1. Effects of Secondary Circuit Modeling on Results of Pressurized Water Reactor Main Steam Line Break Benchmark Calculations with New Coupled Code TRAB-3D/SMABRE

    SciTech Connect

    Daavittila, Antti; Haemaelaeinen, Anitta; Kyrki-Rajamaeki, Riitta

    2003-05-15

    All of the three exercises of the Organization for Economic Cooperation and Development/Nuclear Regulatory Commission pressurized water reactor main steam line break (PWR MSLB) benchmark were calculated at VTT, the Technical Research Centre of Finland. For the first exercise, the plant simulation with point-kinetic neutronics, the thermal-hydraulics code SMABRE was used. The second exercise was calculated with the three-dimensional reactor dynamics code TRAB-3D, and the third exercise with the combination TRAB-3D/SMABRE. VTT has over ten years' experience of coupling neutronic and thermal-hydraulic codes, but this benchmark was the first time these two codes, both developed at VTT, were coupled together. The coupled code system is fast and efficient; the total computation time of the 100-s transient in the third exercise was 16 min on a modern UNIX workstation. The results of all the exercises are similar to those of the other participants. In order to demonstrate the effect of secondary circuit modeling on the results, three different cases were calculated. In case 1 there is no phase separation in the steam lines and no flow reversal in the aspirator. In case 2 the flow reversal in the aspirator is allowed, but there is no phase separation in the steam lines. Finally, in case 3 the drift-flux model is used for the phase separation in the steam lines, but the aspirator flow reversal is not allowed. With these two modeling variations, it is possible to cover a remarkably broad range of results. The maximum power level reached after the reactor trip varies from 534 to 904 MW, the range of the time of the power maximum being close to 30 s. Compared to the total calculated transient time of 100 s, the effect of the secondary side modeling is extremely important.

  2. Quantum description of coupling to neutron-rearrangement channels in fusion reactions near the Coulomb barrier

    SciTech Connect

    Samarin, V. V.

    2015-10-15

    The fusion cross sections for the {sup 17,18}O+{sup 27}Al, {sup 18}O+{sup 58}Ni, and {sup 6}He+{sup 197}Au reactions were calculated by the coupled-channel method. The radial dependence of matrices that describe coupling to valence-neutron-rearrangement channels was determined with the aid of two-center wave functions. The coupling-strength parameters were evaluated on the basis of numerically solving the time-dependent Schrödinger equation. Satisfactory agreement with experimental data was obtained.

  3. Comparisons between the TOPLHA and the ALOHA codes on Lower Hybrid antenna coupling

    NASA Astrophysics Data System (ADS)

    Hillairet, J.; Meneghini, O.; Milanesio, D.; Voyer, D.; Goniche, M.

    2009-11-01

    Comparisons have been made between TOPLHA and ALOHA, two codes for Lower Hybrid antenna coupling to cold inhomogeneous plasmas, on the Tore-Supra C2 multi-junction and an ITER relevant Passive-Active Multi-junction (PAM). Calculation for several plasma scenarios of reflexion coefficients and launched spectra are in good agreement.

  4. Modeling equal and unequal mass binary neutron star mergers using public codes

    NASA Astrophysics Data System (ADS)

    De Pietri, Roberto; Feo, Alessandra; Maione, Francesco; Löffler, Frank

    2016-03-01

    We present three-dimensional simulations of the dynamics of binary neutron star mergers from the late stage of the inspiral process up to ˜20 ms after the system has merged, either to form a hypermassive neutron star or a rotating black hole. We investigate five equal mass models of total gravitational mass 2.207, 2.373, 2.537, 2.697, and 2.854 M⊙, respectively, and four unequal mass models with MADM≃2.53 M⊙ and q ≃0.94 , 0.88, 0.83, and 0.77 (where q =M(1 )/M(2 ) is the mass ratio). We use a semirealistic equation of state, namely, the seven-segment piecewise polytropic SLyPP with a thermal component given by Γth=1.8 . We have also compared the resulting dynamics (for one model) using both the BSSN-NOK and CCZ4 methods for the evolution of the gravitational sector and also different reconstruction methods for the matter sector, namely, PPM, WENO, and MP5. Our results show agreement at high resolution, but superiority of BSSN-NOK supplemented by WENO reconstruction at lower resolutions. One of the important characteristics of the present investigation is that for the first time it has been done using only publicly available open source software: the Einstein Toolkit code, deployed for the dynamical evolution, and the LORENE code, for the generation of the initial models. All of the source code and parameters used for the simulations have been made publicly available. This not only makes it possible to rerun and reanalyze our data but also enables others to directly build upon this work for future research.

  5. A Complex-Geometry Validation Experiment for Advanced Neutron Transport Codes

    SciTech Connect

    David W. Nigg; Anthony W. LaPorta; Joseph W. Nielsen; James Parry; Mark D. DeHart; Samuel E. Bays; William F. Skerjanc

    2013-11-01

    The Idaho National Laboratory (INL) has initiated a focused effort to upgrade legacy computational reactor physics software tools and protocols used for support of core fuel management and experiment management in the Advanced Test Reactor (ATR) and its companion critical facility (ATRC) at the INL.. This will be accomplished through the introduction of modern high-fidelity computational software and protocols, with appropriate new Verification and Validation (V&V) protocols, over the next 12-18 months. Stochastic and deterministic transport theory based reactor physics codes and nuclear data packages that support this effort include MCNP5[1], SCALE/KENO6[2], HELIOS[3], SCALE/NEWT[2], and ATTILA[4]. Furthermore, a capability for sensitivity analysis and uncertainty quantification based on the TSUNAMI[5] system has also been implemented. Finally, we are also evaluating the Serpent[6] and MC21[7] codes, as additional verification tools in the near term as well as for possible applications to full three-dimensional Monte Carlo based fuel management modeling in the longer term. On the experimental side, several new benchmark-quality code validation measurements based on neutron activation spectrometry have been conducted using the ATRC. Results for the first four experiments, focused on neutron spectrum measurements within the Northwest Large In-Pile Tube (NW LIPT) and in the core fuel elements surrounding the NW LIPT and the diametrically opposite Southeast IPT have been reported [8,9]. A fifth, very recent, experiment focused on detailed measurements of the element-to-element core power distribution is summarized here and examples of the use of the measured data for validation of corresponding MCNP5, HELIOS, NEWT, and Serpent computational models using modern least-square adjustment methods are provided.

  6. Inlet-Compressor Analysis Performed Using Coupled Computational Fluid Dynamics Codes

    NASA Technical Reports Server (NTRS)

    Cole, Gary L.; Suresh, Ambady; Townsend, Scott

    1999-01-01

    A thorough understanding of dynamic interactions between inlets and compressors is extremely important to the design and development of propulsion control systems, particularly for supersonic aircraft such as the High-Speed Civil Transport (HSCT). Computational fluid dynamics (CFD) codes are routinely used to analyze individual propulsion components. By coupling the appropriate CFD component codes, it is possible to investigate inlet-compressor interactions. The objectives of this work were to gain a better understanding of inlet-compressor interaction physics, formulate a more realistic compressor-face boundary condition for time-accurate CFD simulations of inlets, and to take a first step toward the CFD simulation of an entire engine by coupling multidimensional component codes. This work was conducted at the NASA Lewis Research Center by a team of civil servants and support service contractors as part of the High Performance Computing and Communications Program (HPCCP).

  7. Safety Related Investigations of the VVER-1000 Reactor Type by the Coupled Code System TRACE/PARCS

    NASA Astrophysics Data System (ADS)

    Jaeger, Wadim; Espinoza, Victor Hugo Sánchez; Lischke, Wolfgang

    This study was performed at the Institute of Reactor Safety at the Forschungszentrum Karlsruhe. It is embedded in the ongoing investigations of the international code assessment and maintenance program (CAMP) for qualification and validation of system codes like TRACE(1) and PARCS(2). The chosen reactor type used to validate these two codes was the Russian designed VVER-1000 because the OECD/NEA VVER-1000 Coolant Transient Benchmark Phase 2(3) includes detailed information of the Bulgarian nuclear power plant (NPP) Kozloduy unit 6. The post-test investigations of a coolant mixing experiment have shown that the predicted parameters (coolant temperature, pressure drop, etc.) are in good agreement with the measured data. The coolant mixing pattern, especially in the downcomer, has been also reproduced quiet well by TRACE. The coupled code system TRACE/PARCS which was applied on a postulated main steam line break (MSLB) provided good results compared to reference values and the ones of other participants of the benchmark. The results show that the developed three-dimensional nodalization of the reactor pressure vessel (RPV) is appropriate to describe the coolant mixing phenomena in the downcomer and the lower plenum of a VVER-1000 reactor. This phenomenon is a key issue for investigations of MSLB transient where the thermal hydraulics and the core neutronics are strongly linked. The simulation of the RPV and core behavior for postulated transients using the validated 3D TRACE RPV model, taking into account boundary conditions at vessel in- and outlet, indicates that the results are physically sound and in good agreement to other participant's results.

  8. An introduction to LIME 1.0 and its use in coupling codes for multiphysics simulations.

    SciTech Connect

    Belcourt, Noel; Pawlowski, Roger Patrick; Schmidt, Rodney Cannon; Hooper, Russell Warren

    2011-11-01

    LIME is a small software package for creating multiphysics simulation codes. The name was formed as an acronym denoting 'Lightweight Integrating Multiphysics Environment for coupling codes.' LIME is intended to be especially useful when separate computer codes (which may be written in any standard computer language) already exist to solve different parts of a multiphysics problem. LIME provides the key high-level software (written in C++), a well defined approach (with example templates), and interface requirements to enable the assembly of multiple physics codes into a single coupled-multiphysics simulation code. In this report we introduce important software design characteristics of LIME, describe key components of a typical multiphysics application that might be created using LIME, and provide basic examples of its use - including the customized software that must be written by a user. We also describe the types of modifications that may be needed to individual physics codes in order for them to be incorporated into a LIME-based multiphysics application.

  9. A mono-dimensional nuclear fuel performance analysis code, PUMA, development from a coupled approach

    SciTech Connect

    Cheon, J. S.; Lee, B. O.; Lee, C. B.; Yacout, A. M.

    2013-07-01

    Multidimensional-multi-physical phenomena in nuclear fuels are treated as a set of mono-dimensional-coupled problems which encompass heat, displacement, fuel constituent redistribution, and fission gas release. Rather than uncoupling these coupled equations as in conventional fuel performance analysis codes, efforts are put into to obtain fully coupled solutions by relying on the recent advances of numerical analysis. Through this approach, a new SFR metal fuel performance analysis code, called PUMA (Performance of Uranium Metal fuel rod Analysis code) is under development. Although coupling between temperature and fuel constituent was made easily, the coupling between the mechanical equilibrium equation and a set of stiff kinetics equations for fission gas release is accomplished by introducing one-level Newton scheme through backward differentiation formula. Displacement equations from 1D finite element formulation of the mechanical equilibrium equation are solved simultaneously with stress equation, creep equation, swelling equation, and FGR equations. Calculations was made successfully such that the swelling and the hydrostatic pressure are interrelated each other. (authors)

  10. A comparison of the COG and MCNP codes in computational neutron capture therapy modeling, Part II: gadolinium neutron capture therapy models and therapeutic effects.

    PubMed

    Wangerin, K; Culbertson, C N; Jevremovic, T

    2005-08-01

    The goal of this study was to evaluate the COG Monte Carlo radiation transport code, developed and tested by Lawrence Livermore National Laboratory, for gadolinium neutron capture therapy (GdNCT) related modeling. The validity of COG NCT model has been established for this model, and here the calculation was extended to analyze the effect of various gadolinium concentrations on dose distribution and cell-kill effect of the GdNCT modality and to determine the optimum therapeutic conditions for treating brain cancers. The computational results were compared with the widely used MCNP code. The differences between the COG and MCNP predictions were generally small and suggest that the COG code can be applied to similar research problems in NCT. Results for this study also showed that a concentration of 100 ppm gadolinium in the tumor was most beneficial when using an epithermal neutron beam. PMID:16010124

  11. Updated Users' Guide for RSAP -- A Code for Display and Manipulation of Neutron Cross Section Data and SAMMY Fit Results

    SciTech Connect

    Sayer, R.O.

    2003-07-29

    RSAP [1] is a computer code for display and manipulation of neutron cross section data and selected SAMMY output. SAMMY [2] is a multilevel R-matrix code for fitting neutron time-of-flight cross-section data using Bayes' method. This users' guide provides documentation for the recently updated RSAP code (version 6). The code has been ported to the Linux platform, and several new features have been added, including the capability to read cross section data from ASCII pointwise ENDF files as well as double-precision PLT output from SAMMY. A number of bugs have been found and corrected, and the input formats have been improved. Input items are parsed so that items may be separated by spaces or commas.

  12. Secondary fusion coupled deuteron/triton transport simulation and thermal-to-fusion neutron convertor measurement

    SciTech Connect

    Wang, G. B.; Wang, K.; Liu, H. G.; Li, R. D.

    2013-07-01

    A Monte Carlo tool RSMC (Reaction Sequence Monte Carlo) was developed to simulate deuteron/triton transportation and reaction coupled problem. The 'Forced particle production' variance reduction technique was used to improve the simulation speed, which made the secondary product play a major role. The mono-energy 14 MeV fusion neutron source was employed as a validation. Then the thermal-to-fusion neutron convertor was studied with our tool. Moreover, an in-core conversion efficiency measurement experiment was performed with {sup 6}LiD and {sup 6}LiH converters. Threshold activation foils was used to indicate the fast and fusion neutron flux. Besides, two other pivotal parameters were calculated theoretically. Finally, the conversion efficiency of {sup 6}LiD is obtained as 1.97x10{sup -4}, which matches well with the theoretical result. (authors)

  13. Code System for Calculating Alpha, N; Spontaneous Fission; and Delayed Neutron Sources and Spectra.

    Energy Science and Technology Software Center (ESTSC)

    2002-07-18

    Version: 04 SOURCES4C is a code system that determines neutron production rates and spectra from (alpha,n) reactions, spontaneous fission, and delayed neutron emission due to radionuclide decay. In this release the three-region problem was modified to correct several bugs, and new documentation was added to the package. Details are available in the included LA-UR-02-1617 (2002) report. The code is capable of calculating (alpha,n) source rates and spectra in four types of problems: homogeneous media (i.e.,more » an intimate mixture of alpha-emitting source material and low-Z target material), two-region interface problems (i.e., a slab of alpha-emitting source material in contact with a slab of low-Z target material), three-region interface problems (i.e., a thin slab of low-Z target material sandwiched between alpha-emitting source material and low-Z target material), and (alpha,n) reactions induced by a monoenergetic beam of alpha-particles incident on a slab of target material. The process of creating a SOURCES input file (tape1) is streamlined with the Los Alamos SOURCES Tape1 Creator and Library Link (LASTCALL) Version 1. Intended to supplement the SOURCES manual, LASTCALL is a simple graphical user interface designed to minimize common errors made during input. The optional application, LASTCALL, consists of a single dialog window launched from an executable (lastcall.exe) on Windows-based personal computers.« less

  14. Validation of the new code package APOLLO2.8 for accurate PWR neutronics calculations

    SciTech Connect

    Santamarina, A.; Bernard, D.; Blaise, P.; Leconte, P.; Palau, J. M.; Roque, B.; Vaglio, C.; Vidal, J. F.

    2013-07-01

    This paper summarizes the Qualification work performed to demonstrate the accuracy of the new APOLLO2.S/SHEM-MOC package based on JEFF3.1.1 nuclear data file for the prediction of PWR neutronics parameters. This experimental validation is based on PWR mock-up critical experiments performed in the EOLE/MINERVE zero-power reactors and on P.I. Es on spent fuel assemblies from the French PWRs. The Calculation-Experiment comparison for the main design parameters is presented: reactivity of UOX and MOX lattices, depletion calculation and fuel inventory, reactivity loss with burnup, pin-by-pin power maps, Doppler coefficient, Moderator Temperature Coefficient, Void coefficient, UO{sub 2}-Gd{sub 2}O{sub 3} poisoning worth, Efficiency of Ag-In-Cd and B4C control rods, Reflector Saving for both standard 2-cm baffle and GEN3 advanced thick SS reflector. From this qualification process, calculation biases and associated uncertainties are derived. This code package APOLLO2.8 is already implemented in the ARCADIA new AREVA calculation chain for core physics and is currently under implementation in the future neutronics package of the French utility Electricite de France. (authors)

  15. A coupled deterministic/stochastic method for computing neutron capture therapy dose rates

    NASA Astrophysics Data System (ADS)

    Hubbard, Thomas Richard

    Neutron capture therapy (NCT) is an experimental method of treating brain tumors and other cancers by: (1) injecting or infusing the patient with a tumor-seeking, neutron target-labeled drug; and (2) irradiating the patient in an intense epithermal neutron fluence. The nuclear reaction between the neutrons and the target nuclei (e.g. sp{10}B(n,alpha)sp7Lirbrack releases energy in the form of high-LET (i.e. energy deposited within the range of a cell diameter) reaction particles which selectively kill the tumor cell. The efficacy of NCT is partly dependent on the delivery of maximum thermal neutron fluence to the tumor and the minimization of radiation dose to healthy tissue. Since the filtered neutron source (e.g. research reactor) usually provides a broad energy spectrum of highly-penetrating neutron and gamma-photon radiation, detailed transport calculations are necessary in order to plan treatments that use optimal treatment facility configurations and patient positioning. Current computational methods for NCT use either discrete ordinates calculation or, more often, Monte Carlo simulation to predict neutron fluences in the vicinity of the tumor. These methods do not, however, accurately calculate the transport of radiation throughout the entire facility or the deposition of dose in all the various parts of the body due to shortcomings of using either method alone. A computational method, specifically designed for NCT problems, has been adapted from the MASH methodology and couples a forward discrete ordinates (Ssb{n}) calculation with an adjoint Monte Carlo run to predict the dose at any point within the patient. The transport from the source through the filter/collimator is performed with a forward DORT run, and this is then coupled to adjoint MORSE results at a selected coupling parallelepiped which surrounds human phantom. Another routine was written to allow the user to generate the MORSE models at various angles and positions within the treatment room. The

  16. Observational constraints on neutron star crust-core coupling during glitches

    NASA Astrophysics Data System (ADS)

    Newton, W. G.; Berger, S.; Haskell, B.

    2015-12-01

    We demonstrate that observations of glitches in the Vela pulsar can be used to investigate the strength of the crust-core coupling in a neutron star and provide a powerful probe of the internal structure of neutron stars. We assume that glitch recovery is dominated by the torque exerted by the mutual friction-mediated recoupling of superfluid components of the core that were decoupled from the crust during the glitch. Then we use the observations of the recoveries from two recent glitches in the Vela pulsar to infer the fraction of the core that is coupled to the crust during the glitch. We then analyse whether crustal neutrons alone are sufficient to drive glitches in the Vela pulsar, taking into account crustal entrainment. We use two sets of neutron star equations of state (EOSs) which span crust and core consistently and cover a conservative range of the slope of the symmetry energy at saturation density 30 < L < 120 MeV. The two sets differ in the stiffness of the high density EOS. We find that for medium to stiff EOSs, observations imply >70 per cent of the moment of inertia of the core is coupled to the crust during the glitch, though for softer EOSs L ≈ 30 MeV as little as 5 per cent could be coupled. We find that only by extending the region where superfluid vortices are strongly pinned into the core by densities at least 0.016 fm-3 above the crust-core transition density does any EOS reproduce the observed glitch activity.

  17. Radiation transport codes for potential applications related to radiobiology and radiotherapy using protons, neutrons, and negatively charged pions

    NASA Technical Reports Server (NTRS)

    Armstrong, T. W.

    1972-01-01

    Several Monte Carlo radiation transport computer codes are used to predict quantities of interest in the fields of radiotherapy and radiobiology. The calculational methods are described and comparisions of calculated and experimental results are presented for dose distributions produced by protons, neutrons, and negatively charged pions. Comparisons of calculated and experimental cell survival probabilities are also presented.

  18. ITS Version 6 : the integrated TIGER series of coupled electron/photon Monte Carlo transport codes.

    SciTech Connect

    Franke, Brian Claude; Kensek, Ronald Patrick; Laub, Thomas William

    2008-04-01

    ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of lineartime-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 6, the latest version of ITS, contains (1) improvements to the ITS 5.0 codes, and (2) conversion to Fortran 90. The general user friendliness of the software has been enhanced through memory allocation to reduce the need for users to modify and recompile the code.

  19. Neutron and ARPES constraints on the couplings of the multiorbital Hubbard model for the iron pnictides

    SciTech Connect

    Luo, Qinlong; Martins, G. B.; Daghofer, M.; Yu, Rong; Yildirim, Yucel; Moreo, Adriana; Dagotto, Elbio R

    2010-01-01

    The results of neutron-scattering and angle-resolved photoemission experiments for the Fe-pnictide parent compounds, and their metallic nature, are shown to impose severe constraints on the range of values that can be considered realistic for the intraorbital Hubbard repulsion U and Hund coupling J in multiorbital Hubbard models treated in the mean-field approximation. Phase diagrams for three- and five-orbital models are here provided, and the physically realistic regime of couplings is highlighted, to guide future theoretical work into the proper region of parameters of Hubbard models. In addition, using the random phase approximation, the pairing tendencies in these realistic coupling regions are investigated. It is shown that the dominant spin-singlet pairing channels in these coupling regimes correspond to nodal superconductivity, with strong competition between several states that belong to different irreducible representations. This is compatible with experimental bulk measurements that have reported the existence of nodes in several Fe-pnictide compounds.

  20. A Visualization Code System for Gamma and Neutron Shielding Calculations, Version 2.0

    Energy Science and Technology Software Center (ESTSC)

    2008-08-01

    EASYQAD, Version 2.0, is a standalone Windows XP or Windows 7 code system which facilitates gamma and neutron shielding calculations with user friendly graphical interfaces. It is used to analyze radiation shielding problems and includes: - 8 kinds of geometry types - Various flexible source options - Common material library - Various detector types The update contents of EASYQAD Version 2.0 are below: - Addition of starting option with ‘P-code’ files - Addition of multi-sourcemore » calculation function - Expansion of source geometries - Addition of warning message - Modifications of EASYQAD program errors a. Coordination application problem in source division b. Source position error c. Rotation problem of source geometry d. Program running error in using more than six gamma energy distribution e. EASYQAD display problem of the right elliptic cylinder, ellipsoid and truncated right cone geometries Through intuitive windows and their interactions inside EASYQAD, the user can specify the dimensions of 3D-shapes, their material compositions, their densities, the type of radioactive sources, the locations of the sources, the type and positions of detectors. With the ease of using these sequences, shielding problems will become simpler and more clearly understandable to the analyzer. Furthermore, the error checking system can prevent users from making mistakes by automatically debugging the user inputs and giving modal dialog windows. The included AECL implementation of QAD-CGGP-A, Version 95.2 (C00645MNYCP00), is run from the user interface.« less

  1. The TORT three-dimensional discrete ordinates neutron/photon transport code (TORT version 3)

    SciTech Connect

    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.

  2. Measured performances on vectorization and multitasking with a Monte Carlo code for neutron transport problems

    NASA Astrophysics Data System (ADS)

    Chauvet, Yves

    1985-07-01

    This paper summarized two improvements of a real production code by using vectorization and multitasking techniques. After a short description of Monte Carlo algorithms employed in our neutron transport problems, we briefly describe the work we have done in order to get a vector code. Vectorization principles will be presented and measured performances on the CRAY 1S, CYBER 205 and CRAY X-MP compared in terms of vector lengths. The second part of this work is an adaptation to multitasking on the CRAY X-MP using exclusively standard multitasking tools available with FORTRAN under the COS 1.13 system. Two examples will be presented. The goal of the first one is to measure the overhead inherent to multitasking when tasks become too small and to define a granularity threshold that is to say a minimum size for a task. With the second example we propose a method that is very X-MP oriented in order to get the best speedup factor on such a computer. In conclusion we prove that Monte Carlo algorithms are very well suited to future vector and parallel computers.

  3. Use of SUSA in Uncertainty and Sensitivity Analysis for INL VHTR Coupled Codes

    SciTech Connect

    Gerhard Strydom

    2010-06-01

    The need for a defendable and systematic Uncertainty and Sensitivity approach that conforms to the Code Scaling, Applicability, and Uncertainty (CSAU) process, and that could be used for a wide variety of software codes, was defined in 2008.The GRS (Gesellschaft für Anlagen und Reaktorsicherheit) company of Germany has developed one type of CSAU approach that is particularly well suited for legacy coupled core analysis codes, and a trial version of their commercial software product SUSA (Software for Uncertainty and Sensitivity Analyses) was acquired on May 12, 2010. This interim milestone report provides an overview of the current status of the implementation and testing of SUSA at the INL VHTR Project Office.

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

    SciTech Connect

    Burns, Kimberly A.

    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. Computer codes for checking, plotting and processing of neutron cross-section covariance data and their application

    SciTech Connect

    Sartori, E.

    1992-12-31

    This paper presents a brief review of computer codes concerned with checking, plotting, processing and using of covariances of neutron cross-section data. It concentrates on those available from the computer code information centers of the United States and the OECD/Nuclear Energy Agency. Emphasis will be placed also on codes using covariances for specific applications such as uncertainty analysis, data adjustment and data consistency analysis. Recent evaluations contain neutron cross section covariance information for all isotopes of major importance for technological applications of nuclear energy. It is therefore important that the available software tools needed for taking advantage of this information are widely known as hey permit the determination of better safety margins and allow the optimization of more economic, I designs of nuclear energy systems.

  6. Studies of coupled cavity LINAC (CCL) accelerating structures with 3-D codes

    SciTech Connect

    Spalek, G.

    2000-08-01

    The cw CCL being designed for the Accelerator Production of Tritium (APT) project accelerates protons from 96 MeV to 211 MeV. It consists of 99 segments each containing up to seven accelerating cavities. Segments are coupled by intersegment coupling cavities and grouped into supermodules. The design method needs to address not only basic cavity sizing for a given coupling and pi/2 mode frequency, but also the effects of high power densities on the cavity frequency, mechanical stresses, and the structure's stop band during operation. On the APT project, 3-D RF (Ansoft Corp.'s HFSS) and coupled RF/structural (Ansys Inc.'s ANSYS) codes are being used. to develop tools to address the above issues and guide cooling channel design. The code's predictions are being checked against available low power Aluminum models. Stop band behavior under power will be checked once the tools are extended to CCDTL structures that have been tested at high power. A summary of calculations made to date and agreement with measured results will be presented.

  7. Mixing models for the two-way-coupling of CFD codes and zero-dimensional multi-zone codes to model HCCI combustion

    SciTech Connect

    Barths, H.; Felsch, C.; Peters, N.

    2008-11-15

    The objective of this work is the development of a consistent mixing model for the two-way-coupling of a CFD code and a multi-zone code based on multiple zero-dimensional reactors. The two-way-coupling allows for a computationally efficient modeling of HCCI combustion. The physical domain in the CFD code is subdivided into multiple zones based on three phase variables (fuel mixture fraction, dilution, and total enthalpy). Those phase variables are sufficient for the description of the thermodynamic state of each zone, assuming that each zone is at the same pressure. Each zone in the CFD code is represented by a corresponding zone in the zero-dimensional code. The zero-dimensional code solves the chemistry for each zone, and the heat release is fed back into the CFD code. The difficulty in facing this kind of methodology is to keep the thermodynamic state of each zone consistent between the CFD code and the zero-dimensional code after the initialization of the zones in the multi-zone code has taken place. The thermodynamic state of each zone (and thereby the phase variables) will change in time due to mixing and source terms (e.g., vaporization of fuel, wall heat transfer). The focus of this work lies on a consistent description of the mixing between the zones in phase space in the zero-dimensional code, based on the solution of the CFD code. Two mixing models with different degrees of accuracy, complexity, and numerical effort are described. The most elaborate mixing model (and an appropriate treatment of the source terms) keeps the thermodynamic state of the zones in the CFD code and the zero-dimensional code identical. The models are applied to a test case of HCCI combustion in an engine. (author)

  8. Mixing models for the two-way-coupling of CFD codes and zero-dimensional multi-zone codes to model HCCI combustion

    SciTech Connect

    Barths, H.; Felsch, C.; Peters, N.

    2009-01-15

    The objective of this work is the development of a consistent mixing model for the two-way-coupling of a CFD code and a multi-zone code based on multiple zero-dimensional reactors. The two-way-coupling allows for a computationally efficient modeling of HCCI combustion. The physical domain in the CFD code is subdivided into multiple zones based on three phase variables (fuel mixture fraction, dilution, and total enthalpy). Those phase variables are sufficient for the description of the thermodynamic state of each zone, assuming that each zone is at the same pressure. Each zone in the CFD code is represented by a corresponding zone in the zero-dimensional code. The zero-dimensional code solves the chemistry for each zone, and the heat release is fed back into the CFD code. The difficulty in facing this kind of methodology is to keep the thermodynamic state of each zone consistent between the CFD code and the zero-dimensional code after the initialization of the zones in the multi-zone code has taken place. The thermodynamic state of each zone (and thereby the phase variables) will change in time due to mixing and source terms (e.g., vaporization of fuel, wall heat transfer). The focus of this work lies on a consistent description of the mixing between the zones in phase space in the zero-dimensional code, based on the solution of the CFD code. Two mixing models with different degrees of accuracy, complexity, and numerical effort are described. The most elaborate mixing model (and an appropriate treatment of the source terms) keeps the thermodynamic state of the zones in the CFD code and the zero-dimensional code identical. The models are applied to a test case of HCCI combustion in an engine. (author)

  9. Development of Tritium Permeation Analysis Code and Tritium Transport in a High Temperature Gas-Cooled Reactor Coupled with Hydrogen Production System

    SciTech Connect

    Chang H. Oh; Eung S. Kim; Mike Patterson

    2010-06-01

    Abstract – A tritium permeation analyses code (TPAC) was developed by Idaho National Laboratory for the purpose of analyzing tritium distributions in very high temperature reactor (VHTR) systems, including integrated hydrogen production systems. A MATLAB SIMULINK software package was used in developing the code. The TPAC is based on the mass balance equations of tritium-containing species and various forms of hydrogen coupled with a variety of tritium sources, sinks, and permeation models. In the TPAC, ternary fission and neutron reactions with 6Li, 7Li 10B, and 3He were taken into considerations as tritium sources. Purification and leakage models were implemented as main tritium sinks. Permeation of tritium and H2 through pipes, vessels, and heat exchangers were considered as main tritium transport paths. In addition, electroyzer and isotope exchange models were developed for analyzing hydrogen production systems, including high temperature electrolysis and sulfur-iodine processes.

  10. MULTIDIMENSIONAL COUPLED PHOTON-ELECTRON TRANSPORT SIMULATIONS USING NEUTRAL PARTICLE SN CODES

    SciTech Connect

    Ilas, Dan; Williams, Mark L; Peplow, Douglas E.; Kirk, Bernadette Lugue

    2008-01-01

    During the past two years a study was underway at ORNL to assess the suitability of the popular SN neutral particle codes ANISN, DORT and TORT for coupled photon-electron calculations specific to external beam therapy of medical physics applications. The CEPXS-BFP code was used to generate the cross sections. The computational tests were performed on phantoms typical of those used in medical physics for external beam therapy, with materials simulated by water at different densities and the comparisons were made against Monte Carlo simulations that served as benchmarks. Although the results for one-dimensional calculations were encouraging, it appeared that the higher dimensional transport codes had fundamental difficulties in handling the electron transport. The results of two-dimensional simulations using the code DORT with an S16 fully symmetric quadrature set agree fairly with the reference Monte Carlo results but not well enough for clinical applications. While the photon fluxes are in better agreement (generally, within less than 5% from the reference), the discrepancy increases, sometimes very significantly, for the electron fluxes. The paper, however, focuses on the results obtained with the three-dimensional code TORT which had convergence difficulties for the electron groups. Numerical instabilities occurred in these groups. These instabilities were more pronounced with the degree of anisotropy of the problem.

  11. The stability of tidally deformed neutron stars to three- and four-mode coupling

    SciTech Connect

    Venumadhav, Tejaswi; Zimmerman, Aaron; Hirata, Christopher M.

    2014-01-20

    It has recently been suggested that the tidal deformation of a neutron star excites daughter p- and g-modes to large amplitudes via a quasi-static instability. This would remove energy from the tidal bulge, resulting in dissipation and possibly affecting the phase evolution of inspiralling binary neutron stars and hence the extraction of binary parameters from gravitational wave observations. This instability appears to arise because of a large three-mode interaction among the tidal mode and high-order p- and g-modes of similar radial wavenumber. We show that additional four-mode interactions enter into the analysis at the same order as the three-mode terms previously considered. We compute these four-mode couplings by finding a volume-preserving coordinate transformation that relates the energy of a tidally deformed star to that of a radially perturbed spherical star. Using this method, we relate the four-mode coupling to three-mode couplings and show that there is a near-exact cancellation between the destabilizing effect of the three-mode interactions and the stabilizing effect of the four-mode interaction. We then show that the equilibrium tide is stable against the quasi-static decay into daughter p- and g-modes to leading order. The leading deviation from the quasi-static approximation due to orbital motion of the binary is considered; while it may slightly spoil the near-cancellation, any resulting instability timescale is at least of order the gravitational wave inspiral time. We conclude that the p-/g-mode coupling does not lead to a quasi-static instability, and does not impact the phase evolution of gravitational waves from binary neutron stars.

  12. A new three-tier architecture design for multi-sphere neutron spectrometer with the FLUKA code

    NASA Astrophysics Data System (ADS)

    Huang, Hong; Yang, Jian-Bo; Tuo, Xian-Guo; Liu, Zhi; Wang, Qi-Biao; Wang, Xu

    2016-07-01

    The current commercially, available Bonner sphere neutron spectrometer (BSS) has high sensitivity to neutrons below 20 MeV, which causes it to be poorly placed to measure neutrons ranging from a few MeV to 100 MeV. The paper added moderator layers and the auxiliary material layer upon 3He proportional counters with FLUKA code, with a view to improve. The results showed that the responsive peaks to neutrons below 20 MeV gradually shift to higher energy region and decrease slightly with the increasing moderator thickness. On the contrary, the response for neutrons above 20 MeV was always very low until we embed auxiliary materials such as copper (Cu), lead (Pb), tungsten (W) into moderator layers. This paper chose the most suitable auxiliary material Pb to design a three-tier architecture multi-sphere neutron spectrometer (NBSS). Through calculating and comparing, the NBSS was advantageous in terms of response for 5-100 MeV and the highest response was 35.2 times the response of polyethylene (PE) ball with the same PE thickness.

  13. Radiation Coupling with the FUN3D Unstructured-Grid CFD Code

    NASA Technical Reports Server (NTRS)

    Wood, William A.

    2012-01-01

    The HARA radiation code is fully-coupled to the FUN3D unstructured-grid CFD code for the purpose of simulating high-energy hypersonic flows. The radiation energy source terms and surface heat transfer, under the tangent slab approximation, are included within the fluid dynamic ow solver. The Fire II flight test, at the Mach-31 1643-second trajectory point, is used as a demonstration case. Comparisons are made with an existing structured-grid capability, the LAURA/HARA coupling. The radiative surface heat transfer rates from the present approach match the benchmark values within 6%. Although radiation coupling is the focus of the present work, convective surface heat transfer rates are also reported, and are seen to vary depending upon the choice of mesh connectivity and FUN3D ux reconstruction algorithm. On a tetrahedral-element mesh the convective heating matches the benchmark at the stagnation point, but under-predicts by 15% on the Fire II shoulder. Conversely, on a mixed-element mesh the convective heating over-predicts at the stagnation point by 20%, but matches the benchmark away from the stagnation region.

  14. Integrated TIGER Series of Coupled Electron/Photon Monte Carlo Transport Codes System.

    Energy Science and Technology Software Center (ESTSC)

    2012-11-30

    Version: 00 Distribution is restricted to US Government Agencies and Their Contractors Only. The Integrated Tiger Series (ITS) is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of linear time-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. The goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects onemore » of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 6, the latest version of ITS, contains (1) improvements to the ITS 5.0 codes, and (2) conversion to Fortran 95. The general user friendliness of the software has been enhanced through memory allocation to reduce the need for users to modify and recompile the code.« less

  15. Integrated TIGER Series of Coupled Electron/Photon Monte Carlo Transport Codes System.

    SciTech Connect

    VALDEZ, GREG D.

    2012-11-30

    Version: 00 Distribution is restricted to US Government Agencies and Their Contractors Only. The Integrated Tiger Series (ITS) is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of linear time-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. The goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 6, the latest version of ITS, contains (1) improvements to the ITS 5.0 codes, and (2) conversion to Fortran 95. The general user friendliness of the software has been enhanced through memory allocation to reduce the need for users to modify and recompile the code.

  16. Monte Carlo Code System for Calculation of Multiple Scattering of Neutrons in the Resonance Region.

    Energy Science and Technology Software Center (ESTSC)

    1983-01-25

    Version 00 MCRTOF systematically calculates capture and scattering probabilities for neutrons incident on a material disk, with neutron cross sections calculated from the resonance parameters. Capture, front and rear face scattering, transmission, etc., probabilities are obtained from the average destinations of the incident neutrons.

  17. ACDOS1: a computer code to calculate dose rates from neutron activation of neutral beamlines and other fusion-reactor components

    SciTech Connect

    Keney, G.S.

    1981-08-01

    A computer code has been written to calculate neutron induced activation of neutral-beam injector components and the corresponding dose rates as a function of geometry, component composition, and time after shutdown. The code, ACDOS1, was written in FORTRAN IV to calculate both activity and dose rates for up to 30 target nuclides and 50 neutron groups. Sufficient versatility has also been incorporated into the code to make it applicable to a variety of general activation problems due to neutrons of energy less than 20 MeV.

  18. Pedestal Fueling Simulations with a Coupled Kinetic-kinetic Plasma-neutral Transport Code

    SciTech Connect

    D.P. Stotler, C.S. Chang, S.H. Ku, J. Lang and G.Y. Park

    2012-08-29

    A Monte Carlo neutral transport routine, based on DEGAS2, has been coupled to the guiding center ion-electron-neutral neoclassical PIC code XGC0 to provide a realistic treatment of neutral atoms and molecules in the tokamak edge plasma. The DEGAS2 routine allows detailed atomic physics and plasma-material interaction processes to be incorporated into these simulations. The spatial pro le of the neutral particle source used in the DEGAS2 routine is determined from the uxes of XGC0 ions to the material surfaces. The kinetic-kinetic plasma-neutral transport capability is demonstrated with example pedestal fueling simulations.

  19. Energy Conservation Tests of a Coupled Kinetic-kinetic Plasma-neutral Transport Code

    SciTech Connect

    Stotler, D. P.; Chang, C. S.; Ku, S. H.; Lang, J.; Park, G.

    2012-08-29

    A Monte Carlo neutral transport routine, based on DEGAS2, has been coupled to the guiding center ion-electron-neutral neoclassical PIC code XGC0 to provide a realistic treatment of neutral atoms and molecules in the tokamak edge plasma. The DEGAS2 routine allows detailed atomic physics and plasma-material interaction processes to be incorporated into these simulations. The spatial pro le of the neutral particle source used in the DEGAS2 routine is determined from the uxes of XGC0 ions to the material surfaces. The kinetic-kinetic plasma-neutral transport capability is demonstrated with example pedestal fueling simulations.

  20. A multilevel method for coupling the neutron kinetics and heat transfer equations

    SciTech Connect

    Tamang, A.; Anistratov, D. Y.

    2013-07-01

    We present a computational method for adequate and efficient coupling the neutron transport equation with the precursor and heat transfer equations. It is based on the multilevel nonlinear quasidiffusion (QD) method for solving the multigroup transport equation. The system of equations includes the time-dependent high-order transport equation and time-dependent multigroup and effective one-group low-order QD equations. We also formulate a method applying the {alpha}-approximation for the time-dependent high-order transport equation. This approach enables one to avoid storing the angular flux from the previous time step. Numerical results for a model transient problem are presented. (authors)

  1. Application of 3-dimensional radiation transport codes to the analysis of the CRBR prototypic coolant pipe chaseway neutron streaming experiment

    SciTech Connect

    Chatani, K. )

    1992-08-01

    This report summarizes the calculational results from analyses of a Clinch River Breeder Reactor (CRBR) prototypic coolant pipe chaseway neutron streaming experiment Comparisons of calculated and measured results are presented, major emphasis being placed on results at bends in the chaseway. Calculations were performed with three three-dimensional radiation transport codes: the discrete ordinates code TORT and the Monte Carlo code MORSE, both developed by the Oak Ridge National Laboratory (ORNL), and the discrete ordinates code ENSEMBLE, developed by Japan. The calculated results from the three codes are compared (1) with previously-calculated DOT3.5 two-dimensional results, (2) among themselves, and (3) with measured results. Calculations with TORT used both the weighted-difference and nodal methods. Only the weighted-difference method was used in ENSEMBLE. When the calculated results were compared to measured results, it was found that calculation-to-experiment (C/E) ratios were good in the regions of the chaseway where two-dimensional modeling might be difficult and where there were no significant discrete ordinates ray effects. Excellent agreement was observed for responses dominated by thermal neutron contributions. MORSE-calculated results and comparisons are described also, and detailed results are presented in an appendix.

  2. Surface code architecture for donors and dots in silicon with imprecise and nonuniform qubit couplings

    NASA Astrophysics Data System (ADS)

    Pica, G.; Lovett, B. W.; Bhatt, R. N.; Schenkel, T.; Lyon, S. A.

    2016-01-01

    A scaled quantum computer with donor spins in silicon would benefit from a viable semiconductor framework and a strong inherent decoupling of the qubits from the noisy environment. Coupling neighboring spins via the natural exchange interaction according to current designs requires gate control structures with extremely small length scales. We present a silicon architecture where bismuth donors with long coherence times are coupled to electrons that can shuttle between adjacent quantum dots, thus relaxing the pitch requirements and allowing space between donors for classical control devices. An adiabatic SWAP operation within each donor/dot pair solves the scalability issues intrinsic to exchange-based two-qubit gates, as it does not rely on subnanometer precision in donor placement and is robust against noise in the control fields. We use this SWAP together with well established global microwave Rabi pulses and parallel electron shuttling to construct a surface code that needs minimal, feasible local control.

  3. SOURCES 4C : a code for calculating ([alpha],n), spontaneous fission, and delayed neutron sources and spectra.

    SciTech Connect

    Wilson, W. B.; Perry, R. T.; Shores, E. F.; Charlton, W. S.; Parish, Theodore A.; Estes, G. P.; Brown, T. H.; Arthur, Edward D. ,; Bozoian, Michael; England, T. R.; Madland, D. G.; Stewart, J. E.

    2002-01-01

    SOURCES 4C is a computer code that determines neutron production rates and spectra from ({alpha},n) reactions, spontaneous fission, and delayed neutron emission due to radionuclide decay. The code is capable of calculating ({alpha},n) source rates and spectra in four types of problems: homogeneous media (i.e., an intimate mixture of a-emitting source material and low-Z target material), two-region interface problems (i.e., a slab of {alpha}-emitting source material in contact with a slab of low-Z target material), three-region interface problems (i.e., a thin slab of low-Z target material sandwiched between {alpha}-emitting source material and low-Z target material), and ({alpha},n) reactions induced by a monoenergetic beam of {alpha}-particles incident on a slab of target material. Spontaneous fission spectra are calculated with evaluated half-life, spontaneous fission branching, and Watt spectrum parameters for 44 actinides. The ({alpha},n) spectra are calculated using an assumed isotropic angular distribution in the center-of-mass system with a library of 107 nuclide decay {alpha}-particle spectra, 24 sets of measured and/or evaluated ({alpha},n) cross sections and product nuclide level branching fractions, and functional {alpha}-particle stopping cross sections for Z < 106. The delayed neutron spectra are taken from an evaluated library of 105 precursors. The code provides the magnitude and spectra, if desired, of the resultant neutron source in addition to an analysis of the'contributions by each nuclide in the problem. LASTCALL, a graphical user interface, is included in the code package.

  4. Optimal Dense Coding and Swap Operation Between Two Coupled Electronic Spins: Effects of Nuclear Field and Spin-Orbit Interaction

    NASA Astrophysics Data System (ADS)

    Jiang, Li; Zhang, Guo-Feng

    2016-08-01

    The effects of nuclear field and spin-orbit interaction on dense coding and swap operation are studied in detail for both the antiferromagnetic (AFM) and ferromagnetic (FM) coupling cases. The conditions for a valid dense coding and under which swap operation is feasible are given.

  5. MAC/GMC Code Enhanced for Coupled Electromagnetothermoelastic Analysis of Smart Composites

    NASA Technical Reports Server (NTRS)

    Bednarcyk, Brett A.; Arnold, Steven M.; Aboudi, Jacob

    2002-01-01

    Intelligent materials are those that exhibit coupling between their electromagnetic response and their thermomechanical response. This coupling allows smart materials to react mechanically (e.g., an induced displacement) to applied electrical or magnetic fields (for instance). These materials find many important applications in sensors, actuators, and transducers. Recently interest has arisen in the development of smart composites that are formed via the combination of two or more phases, one or more of which is a smart material. To design with and utilize smart composites, designers need theories that predict the coupled smart behavior of these materials from the electromagnetothermoelastic properties of the individual phases. The micromechanics model known as the generalized method of cells (GMC) has recently been extended to provide this important capability. This coupled electromagnetothermoelastic theory has recently been incorporated within NASA Glenn Research Center's Micromechanics Analysis Code with Generalized Method of Cells (MAC/GMC). This software package is user friendly and has many additional features that render it useful as a design and analysis tool for composite materials in general, and with its new capabilities, for smart composites as well.

  6. Simulating Coupling Complexity in Space Plasmas: First Results from a new code

    NASA Astrophysics Data System (ADS)

    Kryukov, I.; Zank, G. P.; Pogorelov, N. V.; Raeder, J.; Ciardo, G.; Florinski, V. A.; Heerikhuisen, J.; Li, G.; Petrini, F.; Shematovich, V. I.; Winske, D.; Shaikh, D.; Webb, G. M.; Yee, H. M.

    2005-12-01

    The development of codes that embrace 'coupling complexity' via the self-consistent incorporation of multiple physical scales and multiple physical processes in models has been identified by the NRC Decadal Survey in Solar and Space Physics as a crucial necessary development in simulation/modeling technology for the coming decade. The National Science Foundation, through its Information Technology Research (ITR) Program, is supporting our efforts to develop a new class of computational code for plasmas and neutral gases that integrates multiple scales and multiple physical processes and descriptions. We are developing a highly modular, parallelized, scalable code that incorporates multiple scales by synthesizing 3 simulation technologies: 1) Computational fluid dynamics (hydrodynamics or magneto-hydrodynamics-MHD) for the large-scale plasma; 2) direct Monte Carlo simulation of atoms/neutral gas, and 3) transport code solvers to model highly energetic particle distributions. We are constructing the code so that a fourth simulation technology, hybrid simulations for microscale structures and particle distributions, can be incorporated in future work, but for the present, this aspect will be addressed at a test-particle level. This synthesis we will provide a computational tool that will advance our understanding of the physics of neutral and charged gases enormously. Besides making major advances in basic plasma physics and neutral gas problems, this project will address 3 Grand Challenge space physics problems that reflect our research interests: 1) To develop a temporal global heliospheric model which includes the interaction of solar and interstellar plasma with neutral populations (hydrogen, helium, etc., and dust), test-particle kinetic pickup ion acceleration at the termination shock, anomalous cosmic ray production, interaction with galactic cosmic rays, while incorporating the time variability of the solar wind and the solar cycle. 2) To develop a coronal

  7. A Coupling Methodology for Mesoscale-informed Nuclear Fuel Performance Codes

    SciTech Connect

    Michael Tonks; Derek Gaston; Cody Permann; Paul Millett; Glen Hansen; Dieter Wolf

    2010-10-01

    This study proposes an approach for capturing the effect of microstructural evolution on reactor fuel performance by coupling a mesoscale irradiated microstructure model with a finite element fuel performance code. To achieve this, the macroscale system is solved in a parallel, fully coupled, fully-implicit manner using the preconditioned Jacobian-free Newton Krylov (JFNK) method. Within the JFNK solution algorithm, microstructure-influenced material parameters are calculated by the mesoscale model and passed back to the macroscale calculation. Due to the stochastic nature of the mesoscale model, a dynamic fitting technique is implemented to smooth roughness in the calculated material parameters. The proposed methodology is demonstrated on a simple model of a reactor fuel pellet. In the model, INL’s BISON fuel performance code calculates the steady-state temperature profile in a fuel pellet and the microstructure-influenced thermal conductivity is determined with a phase field model of irradiated microstructures. This simple multiscale model demonstrates good nonlinear convergence and near ideal parallel scalability. By capturing the formation of large mesoscale voids in the pellet interior, the multiscale model predicted the irradiation-induced reduction in the thermal conductivity commonly observed in reactors.

  8. Phase-amplitude coupling supports phase coding in human ECoG

    PubMed Central

    Watrous, Andrew J; Deuker, Lorena; Fell, Juergen; Axmacher, Nikolai

    2015-01-01

    Prior studies have shown that high-frequency activity (HFA) is modulated by the phase of low-frequency activity. This phenomenon of phase-amplitude coupling (PAC) is often interpreted as reflecting phase coding of neural representations, although evidence for this link is still lacking in humans. Here, we show that PAC indeed supports phase-dependent stimulus representations for categories. Six patients with medication-resistant epilepsy viewed images of faces, tools, houses, and scenes during simultaneous acquisition of intracranial recordings. Analyzing 167 electrodes, we observed PAC at 43% of electrodes. Further inspection of PAC revealed that category specific HFA modulations occurred at different phases and frequencies of the underlying low-frequency rhythm, permitting decoding of categorical information using the phase at which HFA events occurred. These results provide evidence for categorical phase-coded neural representations and are the first to show that PAC coincides with phase-dependent coding in the human brain. DOI: http://dx.doi.org/10.7554/eLife.07886.001 PMID:26308582

  9. Characterisation of the TRIUMF neutron facility using a Monte Carlo simulation code.

    PubMed

    Monk, S D; Abram, T; Joyce, M J

    2015-04-01

    Here, the characterisation of the high-energy neutron field at TRIUMF (The Tri Universities Meson Facility, Vancouver, British Columbia) with Monte Carlo simulation software is described. The package used is MCNPX version 2.6.0, with the neutron fluence rate determined at three locations within the TRIUMF Thermal Neutron Facility (TNF), including the exit of the neutron channel where users of the facility can test devices that may be susceptible to the effects of this form of radiation. The facility is often used to roughly emulate the field likely to be encountered at high altitudes due to radiation of galactic origin and thus the simulated information is compared with the energy spectrum calculated to be due to neutron radiation of cosmic origin at typical aircraft altitudes. The calculated values were also compared with neutron flux measurements that were estimated using the activation of various foils by the staff of the facility, showing agreement within an order of magnitude. PMID:25342608

  10. Cold Uniform Matter and Neutron Stars in the Quark-Meson-Coupling Model

    SciTech Connect

    J.R. Stone; P.A.M. Guichon; H.H. Matevosyan; A.W. Thomas

    2007-08-01

    A new density dependent effective baryon-baryon interaction has been recently derived from the quark-meson-coupling (QMC) model, offering impressive results in application to finite nuclei and dense baryon matter. This self-consistent, relativistic quark-level approach is used to construct the Equation of State (EoS) and to calculate key properties of high density matter and cold, slowly rotating neutron stars. The results include predictions for the maximum mass of neutron star models, together with the corresponding radius and central density, as well the properties of neutron stars with mass of order 1.4 M{sub {circle_dot}}. The cooling mechanism allowed by the QMC EoS is explored and the parameters relevant to slow rotation, namely the moment of inertia and the period of rotation investigated. The results of the calculation, which are found to be in good agreement with available observational data, are compared with the predictions of more traditional EoS, based on the A18+{delta}v+UIX* and modified Reid soft core potentials, the Skyrme SkM* interaction and two relativistic mean field (RMF) models for a hybrid stars including quark matter. The QMC EoS provides cold neutron star models with maximum mass 1.9-2.1 M{sub {circle_dot}}, with central density less than 6 times nuclear saturation density (n{sub 0} = 0.16 fm{sup -3}) and offers a consistent description of the stellar mass up to this density limit. In contrast with other models, QMC predicts no hyperon contribution at densities lower than 3n{sub 0}, for matter in {beta}-equilibrium. At higher densities, {Xi}{sup -,0} and {Lambda} hyperons are present. The absence of lighter {Sigma}{sup {+-},0} hyperons is understood as a consequence of antisymmetrization, together with the implementation of the color hyperfine interaction in the response of the quark bag to the nuclear scalar field.

  11. Verification and validation of the MCNPX-PoliMi code for simulations of neutron multiplicity counting systems

    NASA Astrophysics Data System (ADS)

    Clarke, S. D.; Miller, E. C.; Flaska, M.; Pozzi, S. A.; Oberer, R. B.; Chiang, L. G.

    2013-02-01

    Neutron coincidence counting is widely used in nuclear safeguards. Simulations of these systems can be performed using Monte Carlo codes such as MCNPX to aid in calibration or measurement design. However, the MCNPX coincidence-counting routines treat particle histories individually, therefore the dead time of the acquisition electronics is not treated. The MCNPX-PoliMi code provides the ability to model detailed effects such as data-acquisition electronics and system dead times. A specialized post-processing code has been developed to interpret the collision-log file and determine the response of a 3He multiplicity counter. The MCNPX-PoliMi simulation provides the full neutron multiplicity distribution measured by the 3He tubes. This distribution is used to compute the singles, doubles, and triples rates which are the quantities used to determine 235U mass. MCNPX-PoliMi has previously been validated with passive multiplicity measurements. In this study, a detailed analysis of the measurement system operating in active mode is presented for uranium-oxide standards ranging from 0.5 to 4.0 kg with a Canberra JCC-51 active well coincidence counter. MCNPX-PoliMi calculations are also compared with MCNPX. The two codes agree to within 1% for the cases with negligible dead times. The simulations are validated with measurements performed at the Y-12 National Security Complex.

  12. Evaluation of Nodal Reactor Physics Methods for Quasi-Static and Time-Dependent Coupled Neutronic Thermal - Analysis of Pressurized Water Reactor Transients

    NASA Astrophysics Data System (ADS)

    Feltus, Madeline Anne

    1990-01-01

    This thesis examines coupled time-dependent thermal -hydraulic (T/H) and neutronics solution methods for Pressurized Water Reactor (PWR) transient analysis. The degree of equivalence is evaluated between the typical quasi-static approach and a newly-developed iterative tandem method. Four specific PWR transients that exhibit a wide range of Reactor Coolant System (RCS) T/H response were investigated: (1) a Station Blackout Anticipated Transient Without Scram (ATWS), (2) a Loss of Feedwater ATWS, (3) a Total Loss of RCS Flow with Scram, and (4) a Main Steam Line Break (MSLB). Rather than using simplified RCS and core models, the theory and method in this thesis were applied practically by using realistic models for an actual four-loop Westinghouse PWR plant. The time-dependent STAR kinetics code, based on the QUANDRY Analytic Nodal Method, and the RETRAN and MCPWR T/H systems codes were used to develop a new, fully coupled, tandem STAR/MCPWRQ methodology that runs tandemly on an enhanced 386/387 IBM PC architecture. MCPWRQ uses externally calculated power input rather than point kinetics power level results. The tandem method was compared to quasi -static STAR and time-dependent STAR 2-D and 3-D kinetics results. The new STAR/MCPWRQ method uses RETRAN time-dependent T/H and point kinetics power input as a first estimate. STAR and MCPWRQ are used tandemly to couple STAR 3-D, time-dependent core power results with the MCPWRQ RCS T/H phenomena. This thesis shows that: (a) quasi-static and point kinetics methods are not able to describe severe PWR transient phenomena adequately; and (b) fully coupled, 3-D, time -dependent, tandem (or possibly parallel) analysis methods should be used for PWR reactor transients instead. By tandemly coupling the RCS response in terms of updated core inlet conditions with 3-D time-dependent core kinetics response, the core power response and T/H conditions are forced to be self-consistent during the entire transient. The transient analyses

  13. PHISICS multi-group transport neutronic capabilities for RELAP5

    SciTech Connect

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

    2012-07-01

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

  14. Coupling of Sph and Finite Element Codes for Multi-Layer Orbital Debris Shield Design

    NASA Technical Reports Server (NTRS)

    Fahrenthold, Eric P.

    1997-01-01

    Particle-based hydrodynamics models offer distinct advantages over Eulerian and Lagrangian hydrocodes in particular shock physics applications. Particle models are designed to avoid the mesh distortion and state variable diffusion problems which can hinder the effective use of Lagrangian and Eulerian codes respectively. However conventional particle-in-cell and smooth particle hydrodynamics methods employ particles which are actually moving interpolation points. A new particle-based modeling methodology, termed Hamiltonian particle hydrodynamics, was developed by Fahrenthold and Koo (1997) to provide an alternative, fully Lagrangian, energy-based approach to shock physics simulations. This alternative formulation avoids the tensile and boundary instabilities associated with standard smooth particle hydrodynamics formulations and the diffusive grid- to-particle mapping schemes characteristic of particle-in-cell methods. In the work described herein, the method of Fahrenthold and Koo has been extended, by coupling the aforementioned hydrodynamic particle model to a hexahedral finite element based description of the continuum dynamics. The resulting continuum model retains all of the features (including general contact-impact effects) of Hamiltonian particle hydrodynamics, while in addition accounting for tensile strength, plasticity, and damage effects important in the simulation of hypervelocity impact on orbital debris shielding. A three dimensional, vectorized, and autotasked implementation of the extended particle method described here has been coded for application to orbital debris shielding design. Source code for the pre-processor (PREP), analysis code (EXOS), post-processor (POST), and rezoner (ZONE), have been delivered separately, along with a User's Guide describing installation and application of the software.

  15. SOURCES 4A: A Code for Calculating (alpha,n), Spontaneous Fission, and Delayed Neutron Sources and Spectra

    SciTech Connect

    Madland, D.G.; Arthur, E.D.; Estes, G.P.; Stewart, J.E.; Bozoian, M.; Perry, R.T.; Parish, T.A.; Brown, T.H.; England, T.R.; Wilson, W.B.; Charlton, W.S.

    1999-09-01

    SOURCES 4A is a computer code that determines neutron production rates and spectra from ({alpha},n) reactions, spontaneous fission, and delayed neutron emission due to the decay of radionuclides. The code is capable of calculating ({alpha},n) source rates and spectra in four types of problems: homogeneous media (i.e., a mixture of {alpha}-emitting source material and low-Z target material), two-region interface problems (i.e., a slab of {alpha}-emitting source material in contact with a slab of low-Z target material), three-region interface problems (i.e., a thin slab of low-Z target material sandwiched between {alpha}-emitting source material and low-Z target material), and ({alpha},n) reactions induced by a monoenergetic beam of {alpha}-particles incident on a slab of target material. Spontaneous fission spectra are calculated with evaluated half-life, spontaneous fission branching, and Watt spectrum parameters for 43 actinides. The ({alpha},n) spectra are calculated using an assumed isotropic angular distribution in the center-of-mass system with a library of 89 nuclide decay {alpha}-particle spectra, 24 sets of measured and/or evaluated ({alpha},n) cross sections and product nuclide level branching fractions, and functional {alpha}-particle stopping cross sections for Z < 106. The delayed neutron spectra are taken from an evaluated library of 105 precursors. The code outputs the magnitude and spectra of the resultant neutron source. It also provides an analysis of the contributions to that source by each nuclide in the problem.

  16. Multigroup Time-Independent Neutron Transport Code System for Plane or Spherical Geometry.

    Energy Science and Technology Software Center (ESTSC)

    1986-12-01

    Version 00 PALLAS-PL/SP solves multigroup time-independent one-dimensional neutron transport problems in plane or spherical geometry. The problems solved are subject to a variety of boundary conditions or a distributed source. General anisotropic scattering problems are treated for solving deep-penetration problems in which angle-dependent neutron spectra are calculated in detail.

  17. A comparative study of the neutron flux spectra in the MNSR irradiation sites for the HEU and LEU cores using the MCNP4C code.

    PubMed

    Dawahra, S; Khattab, K; Saba, G

    2015-10-01

    A comparative study for fuel conversion from the HEU to LEU in the Miniature Neutron Source Reactor (MNSR) has been performed in this paper using the MCNP4C code. The neutron energy and lethargy flux spectra in the first inner and outer irradiation sites of the MNSR reactor for the existing HEU fuel (UAl4-Al, 90% enriched) and the potential LEU fuels (U3Si2-Al, U3Si-Al, U9Mo-Al, 19.75% enriched and UO2, 12.6% enriched) were investigated using the MCNP4C code. The neutron energy flux spectra for each group was calculated by dividing the neutron flux by the width of each energy group. The neutron flux spectra per unit lethargy was calculated by multiplying the neutron energy flux spectra for each energy group by the average energy of each group. The thermal neutron flux was calculated by summing the neutron fluxes from 0.0 to 0.625 eV, the fast neutron flux was calculated by summing the neutron fluxes from 0.5 MeV to 10 MeV for the existing HEU and potential LEU fuels. Good agreements have been noticed between the flux spectra for the potential LEU fuels and the existing HEU fuels with maximum relative differences less than 10% and 8% in the inner and outer irradiation sites. PMID:26142805

  18. Code System to Calculate Neutron and Gamma-Ray Skyshine Doses Using the Integral Line-Beam Method.

    Energy Science and Technology Software Center (ESTSC)

    2000-11-16

    Version 03 This package includes the SKYNEUT 1.1, SKYDOSE 2.3, MCSKY 2.3 and SKYCONES 1.1 codes plus the DLC-188/SKYDATA library to form a comprehensive system for calculating skyshine doses. See the author's web site for related information: http://athena.mne.ksu.edu/~jks/ SKYNEUT evaluates the neutron and neutron-induced secondary gamma-ray skyshine doses from an isotropic, point, neutron source collimated by three simple geometries: an open silo, a vertical black (perfectly absorbing) wall, and a rectangular building. The source maymore » emit monoenergetic neutrons or neutrons with an arbitrary multigroup spectrum of energies. SKYDOSE evaluates the gamma-ray skyshine dose from an isotropic, monoenergetic, point gamma-photon source collimated by three simple geometries: (1) a source in a silo, (2) a source behind an infinitely long, vertical, black wall, and (3) a source in a rectangular building. In all three geometries an optional overhead slab shield may be specified. MCSKY evaluates the gamma-ray skyshine dose from an isotropic, monoenergetic, point gamma-photon source collimated into either a vertical cone (i.e., silo geometry) or into a vertically oriented structure with an N-sided polygon cross section. An overhead laminate shield composed of two different materials is assumed, although shield thicknesses of zero may be specified to model an unshielded SKYSHINE source. SKYCONES evaluates the skyshine doses produced by a point neutron or gamma-photon source emitting, into the atmosphere, radiation that is collimated into an upward conical annulus between two arbitrary polar angles. The source is assumed to be axially (azimuthally) symmetric about a vertical axis through the source and can have an arbitrary polyenergetic spectrum. Nested contiguous annular cones can thus be used to represent the energy and polar-angle dependence of a skyshine source emitting radiation into the atmosphere.« less

  19. Application of power time-projection on the operator-splitting coupling scheme of the TRACE/S3K coupled code

    SciTech Connect

    Wicaksono, D.; Zerkak, O.; Nikitin, K.; Ferroukhi, H.; Chawla, R.

    2013-07-01

    This paper reports refinement studies on the temporal coupling scheme and time-stepping management of TRACE/S3K, a dynamically coupled code version of the thermal-hydraulics system code TRACE and the 3D core simulator Simulate-3K. The studies were carried out for two test cases, namely a PWR rod ejection accident and the Peach Bottom 2 Turbine Trip Test 2. The solution of the coupled calculation, especially the power peak, proves to be very sensitive to the time-step size with the currently employed conventional operator-splitting. Furthermore, a very small time-step size is necessary to achieve decent accuracy. This degrades the trade-off between accuracy and performance. A simple and computationally cheap implementation of time-projection of power has been shown to be able to improve the convergence of the coupled calculation. This scheme is able to achieve a prescribed accuracy with a larger time-step size. (authors)

  20. Peach Bottom 2 Turbine Trip Simulation Using TRAC-BF1/COS3D, a Best-Estimate Coupled 3-D Core and Thermal-Hydraulic Code System

    SciTech Connect

    Ui, Atsushi; Miyaji, Takamasa

    2004-10-15

    The best-estimate coupled three-dimensional (3-D) core and thermal-hydraulic code system TRAC-BF1/COS3D has been developed. COS3D, based on a modified one-group neutronic model, is a 3-D core simulator used for licensing analyses and core management of commercial boiling water reactor (BWR) plants in Japan. TRAC-BF1 is a plant simulator based on a two-fluid model. TRAC-BF1/COS3D is a coupled system of both codes, which are connected using a parallel computing tool. This code system was applied to the OECD/NRC BWR Turbine Trip Benchmark. Since the two-group cross-section tables are provided by the benchmark team, COS3D was modified to apply to this specification. Three best-estimate scenarios and four hypothetical scenarios were calculated using this code system. In the best-estimate scenario, the predicted core power with TRAC-BF1/COS3D is slightly underestimated compared with the measured data. The reason seems to be a slight difference in the core boundary conditions, that is, pressure changes and the core inlet flow distribution, because the peak in this analysis is sensitive to them. However, the results of this benchmark analysis show that TRAC-BF1/COS3D gives good precision for the prediction of the actual BWR transient behavior on the whole. Furthermore, the results with the modified one-group model and the two-group model were compared to verify the application of the modified one-group model to this benchmark. This comparison shows that the results of the modified one-group model are appropriate and sufficiently precise.

  1. Simulations of the Dynamics of the Coupled Energetic and Relativistic Electrons Using VERB Code

    NASA Astrophysics Data System (ADS)

    Shprits, Y.; Kellerman, A. C.; Drozdov, A.

    2015-12-01

    Modeling and understanding of ring current and radiation belt coupled system has been a grand challenge since the beginning of the space age. In this study we show long term simulations with a 3D VERB code of modeling the radiation belts with boundary conditions derived from observations around geosynchronous orbit. We also present 4D VERB simulations that include convective transport, radial diffusion, pitch angle scattering and local acceleration. VERB simulations show that the lower energy inward transport is dominated by the convection and higher energy transport is dominated by the diffusive radial transport. We also show that at energies of 100s of keV a number of processes work simultaneously including convective transport, radial diffusion, local acceleration, loss to the loss cone and loss to the magnetopause. The results of the simulaiton of March 2013 storm are compared with Van Allen Probes observations.

  2. Simulation of high-energy radiation belt electron fluxes using NARMAX-VERB coupled codes

    PubMed Central

    Pakhotin, I P; Drozdov, A Y; Shprits, Y Y; Boynton, R J; Subbotin, D A; Balikhin, M A

    2014-01-01

    This study presents a fusion of data-driven and physics-driven methodologies of energetic electron flux forecasting in the outer radiation belt. Data-driven NARMAX (Nonlinear AutoRegressive Moving Averages with eXogenous inputs) model predictions for geosynchronous orbit fluxes have been used as an outer boundary condition to drive the physics-based Versatile Electron Radiation Belt (VERB) code, to simulate energetic electron fluxes in the outer radiation belt environment. The coupled system has been tested for three extended time periods totalling several weeks of observations. The time periods involved periods of quiet, moderate, and strong geomagnetic activity and captured a range of dynamics typical of the radiation belts. The model has successfully simulated energetic electron fluxes for various magnetospheric conditions. Physical mechanisms that may be responsible for the discrepancies between the model results and observations are discussed. PMID:26167432

  3. Effects of tensor couplings of ω and ρ mesons on 1S0 nucleon superfluidity in neutron star matter

    NASA Astrophysics Data System (ADS)

    Huang, Xiu-Lin; Xu, Yan; Liu, Cheng-Zhi; Liu, Guang-Zhou

    2016-03-01

    The 1S 0 nucleon superfluidity in neutron star matter was investigated in the framework of relativistic σ-ω-π-ρ model with the tensor couplings of ω and ρ mesons using the relativistic Hartree-Fock (RHF) approximation. It was found that the tensor couplings of ω and ρ mesons lead to a clear growth of the 1S 0 neutron pairing gap in the density range where there exists 1S 0 neutron superfluidity. The 1S 0 pairing gap of proton with the tensor couplings of ω and ρ mesons in the density range of ρB = 0.0-0.079fm-3 is lower and then in the density range of ρB = 0.079-0.383fm-3 higher than the corresponding value without the tensor couplings of ω and ρ mesons. Our results provide a basic to understand the influence of the tensor couplings of ω and ρ mesons on the cooling properties of neutron star.

  4. Neutron scattering from coupled phonon-impurity modes in KCl1-c(KCN)c

    NASA Astrophysics Data System (ADS)

    Nicklow, R. M.; Crummett, W. P.; Mostoller, M.; Wood, R. F.

    1980-09-01

    The hybridization of host-lattice phonons with the internal-energy states of CN- impurities in KCl has been studied by inelastic neutron scattering as a function of temperature between 10 and 100 K for samples with impurity concentrations in the range c=0.4 to 6 at.%. A temperature- and concentration-dependent coupling between phonons with Eg symmetry and the Eg transitions of CN- ions is observed near a frequency of 0.5 THz, a value which is consistent with the energy-level spacings for CN- in KCl as deduced by Beyeler. However, an expected coupling of phonons and CN- transitions with T2g symmetry near the same frequency was not detected. A simple two-level model for the CN- impurity provides a rather good description of the data for the Eg coupled modes for c<2 at.%, but it deviates significantly for larger concentrations. Quasielastic scattering, which has a strong dependence on impurity concentration, temperature, and phonon wave vector, is also observed.

  5. Warthog: A MOOSE-Based Application for the Direct Code Coupling of BISON and PROTEUS

    SciTech Connect

    McCaskey, Alexander J.; Slattery, Stuart; Billings, Jay Jay

    2015-09-01

    The Nuclear Energy Advanced Modeling and Simulation (NEAMS) program from the Department of Energy s Office of Nuclear Energy provides a robust toolkit for the modeling and simulation of current and future advanced nuclear reactor designs. This toolkit provides these technologies organized across product lines: two divisions targeted at fuels and end-to-end reactor modeling, and a third for integration, coupling, and high-level workflow management. The Fuels Product Line and the Reactor Product line provide advanced computational technologies that serve each respective field well, however, their current lack of integration presents a major impediment to future improvements of simulation solution fidelity. There is a desire for the capability to mix and match tools across Product Lines in an effort to utilize the best from both to improve NEAMS modeling and simulation technologies. This report will detail a new effort to provide this Product Line interoperability through the development of a new application called Warthog. This application couples the BISON Fuel Performance application from the Fuels Product Line and the PROTEUS Core Neutronics application from the Reactors Product Line in an effort to utilize the best from all parts of the NEAMS toolkit and improve overall solution fidelity of nuclear fuel simulations. To acheive this, Warthog leverages as much prior work from the NEAMS program as possible, and in doing so, enables interoperability between the disparate MOOSE and SHARP frameworks, and the libMesh and MOAB mesh data formats. The remainder of this report will describe this work in full. We will begin with a detailed look at the individual NEAMS framework technologies used and developed in the various Product Lines, and the current status of their interoperability. We will then introduce the Warthog application: its overall architecture and the ways it leverages the best existing tools from accross the NEAMS toolkit to enable BISON-PROTEUS integration

  6. Comparison of codes and neutron IC data used in US and Russia for the Topaz-II nuclear reactor assessment

    SciTech Connect

    Glushkov, Y.S.; Ponomarev-Stepnoi, N.N.; Kompanietz, G.V.; Gomin, Y.A.; Maiorov, L.V.; Lobynstev, V.A.; Polyakov, D.N.; Sapir, J.; Streetman, J.R.

    1993-11-01

    Topaz-II is a heterogeneous, epithermal reactor, fueled with highly enriched uranium-dioxide, cooled with NaK, and moderated with zirconium-hydride. The reactor core contains 37 single-cell thermionic fuel elements, and is surrounded by a radial beryllium reflector that contains 12 rotatable control drums with poison segments. For the physics analysis of TOPAZ II it is necessary to use the Monte Carlo method. The United States (US) and Russia used two different Monte Carlo codes, namely MCNP and MCU-2, respectively. The work described in this paper was aimed at comparing the codes and neutronic data used in the US and Russia for verification of Topaz-II nuclear safety. For this purpose, the US and Russia developed a joint benchmark model of the Topaz-II reactor. The American and Russian teams performed independent computations for a series of variants representing potential water immersion accidents. Comparison of the MCNP and MCU-2 codes showed somewhat different results both for the absolute values of k{sub eff} and for reactivity effects. Future calculations will be performed to obtain a detailed understanding of the reasons for such discrepancies. For these analyses it will be necessary for the US and Russian teams to exchange neutronic data on Topaz-II physics calculations.

  7. Coupled Neutronics Thermal-Hydraulic Solution of a Full-Core PWR Using VERA-CS

    SciTech Connect

    Clarno, Kevin T; Palmtag, Scott; Davidson, Gregory G; Salko, Robert K; Evans, Thomas M; Turner, John A; Belcourt, Kenneth; Hooper, Russell; Schmidt, Rodney

    2014-01-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing a core simulator called VERA-CS to model operating PWR reactors with high resolution. This paper describes how the development of VERA-CS is being driven by a set of progression benchmark problems that specify the delivery of useful capability in discrete steps. As part of this development, this paper will describe the current capability of VERA-CS to perform a multiphysics simulation of an operating PWR at Hot Full Power (HFP) conditions using a set of existing computer codes coupled together in a novel method. Results for several single-assembly cases are shown that demonstrate coupling for different boron concentrations and power levels. Finally, high-resolution results are shown for a full-core PWR reactor modeled in quarter-symmetry.

  8. Determination of neutron flux distribution by using ANISN, a one-dimensional discrete S sub n ordinates transport code with anisotropic scattering

    NASA Technical Reports Server (NTRS)

    Ghorai, S. K.

    1983-01-01

    The purpose of this project was to use a one-dimensional discrete coordinates transport code called ANISN in order to determine the energy-angle-spatial distribution of neutrons in a 6-feet cube rock box which houses a D-T neutron generator at its center. The project was two-fold. The first phase of the project involved adaptation of the ANISN code written for an IBM 360/75/91 computer to the UNIVAC system at JSC. The second phase of the project was to use the code with proper geometry, source function and rock material composition in order to determine the neutron flux distribution around the rock box when a 14.1 MeV neutron generator placed at its center is activated.

  9. Determination of neutron flux distribution by using ANISN, a one-dimensional discrete S sub n ordinates transport code with anisotropic scattering

    NASA Astrophysics Data System (ADS)

    Ghorai, S. K.

    1983-09-01

    The purpose of this project was to use a one-dimensional discrete coordinates transport code called ANISN in order to determine the energy-angle-spatial distribution of neutrons in a 6-feet cube rock box which houses a D-T neutron generator at its center. The project was two-fold. The first phase of the project involved adaptation of the ANISN code written for an IBM 360/75/91 computer to the UNIVAC system at JSC. The second phase of the project was to use the code with proper geometry, source function and rock material composition in order to determine the neutron flux distribution around the rock box when a 14.1 MeV neutron generator placed at its center is activated.

  10. Code Coupling via Jacobian-Free Newton-Krylov Algorithms with Application to Magnetized Fluid Plasma and Kinetic Neutral Models

    SciTech Connect

    Joseph, Ilon

    2014-05-27

    Jacobian-free Newton-Krylov (JFNK) algorithms are a potentially powerful class of methods for solving the problem of coupling codes that address dfferent physics models. As communication capability between individual submodules varies, different choices of coupling algorithms are required. The more communication that is available, the more possible it becomes to exploit the simple sparsity pattern of the Jacobian, albeit of a large system. The less communication that is available, the more dense the Jacobian matrices become and new types of preconditioners must be sought to efficiently take large time steps. In general, methods that use constrained or reduced subsystems can offer a compromise in complexity. The specific problem of coupling a fluid plasma code to a kinetic neutrals code is discussed as an example.

  11. Role of nuclear couplings in the inelastic excitation of weakly-bound neutron-rich nuclei

    SciTech Connect

    Dasso, C.H.; Lenzi, S.M.; Vitturi, A.

    1996-12-31

    Much effort is presently devoted to the study of nuclear systems far from the stability line. Particular emphasis has been placed in light systems such as {sup 11}Li, {sup 8}B and others, where the very small binding energy of the last particles causes their density distribution to extend considerably outside of the remaining nuclear core. Some of the properties associated with this feature are expected to characterize also heavier systems in the vicinity of the proton or neutron drip lines. It is by now well established that low-lying concentrations of multipole strength arise from pure configurations in which a peculiar matching between the wavelength of the continuum wavefunction of the particles and the range of the weakly-bound hole states occurs. To this end the authors consider the break-up of a weakly-bound system in a heavy-ion collision and focus attention in the inelastic excitation of the low-lying part of the continuum. They make use of the fact that previous investigations have shown that the multipole response in this region is not of a collective nature and describe their excited states as pure particle-hole configurations. Since the relevant parameter determining the strength distributions is the binding energy of the last bound orbital they find it most convenient to use single-particle wavefunctions generated by a sperical square-well potential with characteristic nuclear dimensions and whose depth has been adjusted to give rise to a situation in which the last occupied neutron orbital is loosely-bound. Spin-orbit couplings are, for the present purpose, ignored. The results of this investigation clearly indicate that nuclear couplings have the predominant role in causing projectile dissociation in many circumstances, even at bombarding energies remarkably below the Coulomb barrier.

  12. Code System to Perform Neutronic and Thermal-Hydraulic Subchannel Analysis from Converged Coarse-Mesh Nodal Solutions.

    Energy Science and Technology Software Center (ESTSC)

    2000-06-14

    Version 00 NORMA-FP is an auxiliary program which can perform a neutronic and thermal-hydraulic subchannel analysis, starting from global core calculations carried out by both PSR-471/NORMA or PSR-492/QUARK codes. Detailed flux and power distributions inside homogenized nodes are computed by a two-stage bivariate interpolation method, upon separation of the axial variable for which an analytical solution is adopted. The actual heterogeneous structure of a node is accounted for by fuel rod power factors computed asmore » functions of burnup, burnup-weighted coolant density, and instantaneous coolant density.« less

  13. Description and assessment of RAMONA-3B Mod. 0 Cycle 4: a computer code with three-dimensional neutron kinetics for BWR system transients

    SciTech Connect

    Wulff, W; Cheng, H S; Diamond, D J; Khatib-Rahbar, M

    1984-01-01

    This report documents the physical models and the numerical methods employed in the BWR systems code RAMONA-3B. The RAMONA-3B code simulates three-dimensional neutron kinetics and multichannel core hydraulics of nonhomogeneous, nonequilibrium two-phase flows. RAMONA-3B is programmed to calculate the steady and transient conditions in the main steam supply system for normal and abnormal operational transients, including the performances of plant control and protection systems. Presented are code capabilities and limitations, models and solution techniques, the results of development code assessment and suggestions for improving the code in the future.

  14. Coupled-channel analysis of neutron scattering from /sup 12/C between 9 and 15 MeV

    SciTech Connect

    Hansen, L.F.; Meigooni, A.S.

    1986-07-01

    A deformed and energy dependent phenomenological optical model potential and coupled-channel formalism for deformed nuclei have been used in the analysis of elastic and inelastic (Q = 4.439 MeV) scattering, and analyzing power for neutrons scattered from /sup 12/C in the energy range of 9 to 15 MeV. 6 refs., 1 fig., 1 tab.

  15. Coupled protein domain motion in Taq polymerase revealed by neutron spin-echo spectroscopy

    PubMed Central

    Bu, Zimei; Biehl, Ralf; Monkenbusch, Michael; Richter, Dieter; Callaway, David J. E.

    2005-01-01

    Long-range conformational changes in proteins are ubiquitous in biology for the transmission and amplification of signals; such conformational changes can be triggered by small-amplitude, nanosecond protein domain motion. Understanding how conformational changes are initiated requires the characterization of protein domain motion on these timescales and on length scales comparable to protein dimensions. Using neutron spin-echo spectroscopy (NSE), normal mode analysis, and a statistical-mechanical framework, we reveal overdamped, coupled domain motion within DNA polymerase I from Thermus aquaticus (Taq polymerase). This protein utilizes correlated domain dynamics over 70 Å to coordinate nucleotide synthesis and cleavage during DNA synthesis and repair. We show that NSE spectroscopy can determine the domain mobility tensor, which determines the degree of dynamical coupling between domains. The mobility tensor defines the domain velocity response to a force applied to it or to another domain, just as the sails of a sailboat determine its velocity given the applied wind force. The NSE results provide insights into the nature of protein domain motion that are not appreciated by conventional biophysical techniques. PMID:16306270

  16. Testing universal relations of neutron stars with a nonlinear matter-gravity coupling theory

    SciTech Connect

    Sham, Y.-H.; Lin, L.-M.; Leung, P. T. E-mail: lmlin@phy.cuhk.edu.hk

    2014-02-01

    Due to our ignorance of the equation of state (EOS) beyond nuclear density, there is still no unique theoretical model for neutron stars (NSs). It is therefore surprising that universal EOS-independent relations connecting different physical quantities of NSs can exist. Lau et al. found that the frequency of the f-mode oscillation, the mass, and the moment of inertia are connected by universal relations. More recently, Yagi and Yunes discovered the I-Love-Q universal relations among the mass, the moment of inertia, the Love number, and the quadrupole moment. In this paper, we study these universal relations in the Eddington-inspired Born-Infeld (EiBI) gravity. This theory differs from general relativity (GR) significantly only at high densities due to the nonlinear coupling between matter and gravity. It thus provides us an ideal case to test how robust the universal relations of NSs are with respect to the change of the gravity theory. Due to the apparent EOS formulation of EiBI gravity developed recently by Delsate and Steinhoff, we are able to study the universal relations in EiBI gravity using the same techniques as those in GR. We find that the universal relations in EiBI gravity are essentially the same as those in GR. Our work shows that, within the currently viable coupling constant, there exists at least one modified gravity theory that is indistinguishable from GR in view of the unexpected universal relations.

  17. Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data.

    Energy Science and Technology Software Center (ESTSC)

    1998-05-13

    Version 00 The NJOY nuclear data processing system is a modular computer code used for converting evaluated nuclear data in the ENDF format into libraries useful for applications calculations. Because the Evaluated Nuclear Data File (ENDF) format is used all around the world (e.g., ENDF/B-VI in the US, JEF-2.2 in Europe, JENDL-3.2 in Japan, BROND-2.2 in Russia), NJOY gives its users access to a wide variety of the most up-to-date nuclear data. NJOY provides comprehensivemore » capabilities for processing evaluated data, and it can serve applications ranging from continuous-energy Monte Carlo (MCNP), through deterministic transport codes (DANT, ANISN, DORT), to reactor lattice codes (WIMS, EPRI). NJOY handles a wide variety of nuclear effects, including resonances, Doppler broadening, heating (KERMA), radiation damage, thermal scattering (even cold moderators), gas production, neutrons and charged particles, photoatomic interactions, self shielding, probability tables, photon production, and high-energy interactions (to 150 MeV). Output can include printed listings, special library files for applications, and Postscript graphics (plus color). More information on NJOY is available from the developer's home page at http://t2.lanl.gov. Follow the Tourbus section of the Tour area to find notes from the ICTP lectures held at Trieste in March 1998 on the ENDF format and on the NJOY code.« less

  18. Supercomputing with TOUGH2 family codes for coupled multi-physics simulations of geologic carbon sequestration

    NASA Astrophysics Data System (ADS)

    Yamamoto, H.; Nakajima, K.; Zhang, K.; Nanai, S.

    2015-12-01

    Powerful numerical codes that are capable of modeling complex coupled processes of physics and chemistry have been developed for predicting the fate of CO2 in reservoirs as well as its potential impacts on groundwater and subsurface environments. However, they are often computationally demanding for solving highly non-linear models in sufficient spatial and temporal resolutions. Geological heterogeneity and uncertainties further increase the challenges in modeling works. Two-phase flow simulations in heterogeneous media usually require much longer computational time than that in homogeneous media. Uncertainties in reservoir properties may necessitate stochastic simulations with multiple realizations. Recently, massively parallel supercomputers with more than thousands of processors become available in scientific and engineering communities. Such supercomputers may attract attentions from geoscientist and reservoir engineers for solving the large and non-linear models in higher resolutions within a reasonable time. However, for making it a useful tool, it is essential to tackle several practical obstacles to utilize large number of processors effectively for general-purpose reservoir simulators. We have implemented massively-parallel versions of two TOUGH2 family codes (a multi-phase flow simulator TOUGH2 and a chemically reactive transport simulator TOUGHREACT) on two different types (vector- and scalar-type) of supercomputers with a thousand to tens of thousands of processors. After completing implementation and extensive tune-up on the supercomputers, the computational performance was measured for three simulations with multi-million grid models, including a simulation of the dissolution-diffusion-convection process that requires high spatial and temporal resolutions to simulate the growth of small convective fingers of CO2-dissolved water to larger ones in a reservoir scale. The performance measurement confirmed that the both simulators exhibit excellent

  19. TACI: a code for interactive analysis of neutron data produced by a tissue equivalent proportional counter

    SciTech Connect

    Cummings, F.M.

    1984-06-01

    The TEPC analysis code (TACI) is a computer program designed to analyze pulse height data generated by a tissue equivalent proportional counter (TEPC). It is written in HP BASIC and is for use on an HP-87XM personal computer. The theory of TEPC analysis upon which this code is based is summarized.

  20. Validation and verification of RELAP5 for Advanced Neutron Source accident analysis: Part I, comparisons to ANSDM and PRSDYN codes

    SciTech Connect

    Chen, N.C.J.; Ibn-Khayat, M.; March-Leuba, J.A.; Wendel, M.W.

    1993-12-01

    As part of verification and validation, the Advanced Neutron Source reactor RELAP5 system model was benchmarked by the Advanced Neutron Source dynamic model (ANSDM) and PRSDYN models. RELAP5 is a one-dimensional, two-phase transient code, developed by the Idaho National Engineering Laboratory for reactor safety analysis. Both the ANSDM and PRSDYN models use a simplified single-phase equation set to predict transient thermal-hydraulic performance. Brief descriptions of each of the codes, models, and model limitations were included. Even though comparisons were limited to single-phase conditions, a broad spectrum of accidents was benchmarked: a small loss-of-coolant-accident (LOCA), a large LOCA, a station blackout, and a reactivity insertion accident. The overall conclusion is that the three models yield similar results if the input parameters are the same. However, ANSDM does not capture pressure wave propagation through the coolant system. This difference is significant in very rapid pipe break events. Recommendations are provided for further model improvements.

  1. Fan Noise Prediction System Development: Source/Radiation Field Coupling and Workstation Conversion for the Acoustic Radiation Code

    NASA Technical Reports Server (NTRS)

    Meyer, H. D.

    1993-01-01

    The Acoustic Radiation Code (ARC) is a finite element program used on the IBM mainframe to predict far-field acoustic radiation from a turbofan engine inlet. In this report, requirements for developers of internal aerodynamic codes regarding use of their program output an input for the ARC are discussed. More specifically, the particular input needed from the Bolt, Beranek and Newman/Pratt and Whitney (turbofan source noise generation) Code (BBN/PWC) is described. In a separate analysis, a method of coupling the source and radiation models, that recognizes waves crossing the interface in both directions, has been derived. A preliminary version of the coupled code has been developed and used for initial evaluation of coupling issues. Results thus far have shown that reflection from the inlet is sufficient to indicate that full coupling of the source and radiation fields is needed for accurate noise predictions ' Also, for this contract, the ARC has been modified for use on the Sun and Silicon Graphics Iris UNIX workstations. Changes and additions involved in this effort are described in an appendix.

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

    SciTech Connect

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

    2013-07-01

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

  3. Multigroup 3-Dimensional Neutron Diffusion Nodal Code System with Thermohydraulic Feedbacks.

    Energy Science and Technology Software Center (ESTSC)

    1994-02-07

    Version 01 GNOMER is a program which solves the multigroup neutron diffusion equation on coarse mesh in 1D, 2D, and 3D Cartesian geometry. The program is designed to calculate the global core power distributions (with thermohydraulic feedbacks) as well as power distributions and homogenized cross sections over a fuel assembly.

  4. Numerical simulation of gamma ray and neutron production on lunar surface using MCNPX code

    NASA Astrophysics Data System (ADS)

    Kim, Kyeong Ja

    The production of gamma ray and neutron on various planetary surfaces has been investigated for a few decades by remote sensing techniques. The production of these radiations is due to the nuclear reactions between incoming cosmic rays (i.e. galactic cosmic rays and solar cosmic rays) and planetary materials. In the case of the Moon, not like Mars, returned samples from the previous lunar missions provide us much more realistic parameters in understanding of lunar surface composition although it varies a lot. These constraints provide us restricted settings of parameters for model calculations toward the understanding radiation environment. Recent SELENE (KAGUYA) Gamma Ray spectrometer provides elemental lunar surface maps using major important gamma ray lines, such as Fe, Si, Al, U, Th, Ti, etc toward understanding of the evolution of the lunar surfaces. The effect of size or sub-layered soil deposit in the production of gamma rays and neutrons can be effectively understood using model calculations. Our Monte Carlo N-Particle eXtended (MCNPX) generated provides the gamma ray and neutron productions for various soil settings. Our study demonstrates the gamma ray and neutron production on lunar surfaces of various lunar surface soil settings using numerical simulation of MCNPX.

  5. Self characterization of a coded aperture array for neutron source imaging.

    PubMed

    Volegov, P L; Danly, C R; Fittinghoff, D N; Guler, N; Merrill, F E; Wilde, C H

    2014-12-01

    The neutron imaging system at the National Ignition Facility (NIF) is an important diagnostic tool for measuring the two-dimensional size and shape of the neutrons produced in the burning deuterium-tritium plasma during the stagnation stage of inertial confinement fusion implosions. Since the neutron source is small (∼100 μm) and neutrons are deeply penetrating (>3 cm) in all materials, the apertures used to achieve the desired 10-μm resolution are 20-cm long, triangular tapers machined in gold foils. These gold foils are stacked to form an array of 20 apertures for pinhole imaging and three apertures for penumbral imaging. These apertures must be precisely aligned to accurately place the field of view of each aperture at the design location, or the location of the field of view for each aperture must be measured. In this paper we present a new technique that has been developed for the measurement and characterization of the precise location of each aperture in the array. We present the detailed algorithms used for this characterization and the results of reconstructed sources from inertial confinement fusion implosion experiments at NIF. PMID:25554292

  6. Transmission Probability Code System for Calculating Neutron Flux Distributions in Hexagonal Geometry.

    Energy Science and Technology Software Center (ESTSC)

    1991-01-25

    Version 00 TPHEX calculates the multigroup neutron flux distribution in an assembly of hexagonal cells using a transmission probability (interface current) method. It is primarily intended for calculations on hexagonal LWR fuel assemblies but can be used for other purposes subject to the qualifications mentioned in Restrictions/Limitations.

  7. Self characterization of a coded aperture array for neutron source imaging

    NASA Astrophysics Data System (ADS)

    Volegov, P. L.; Danly, C. R.; Fittinghoff, D. N.; Guler, N.; Merrill, F. E.; Wilde, C. H.

    2014-12-01

    The neutron imaging system at the National Ignition Facility (NIF) is an important diagnostic tool for measuring the two-dimensional size and shape of the neutrons produced in the burning deuterium-tritium plasma during the stagnation stage of inertial confinement fusion implosions. Since the neutron source is small (˜100 μm) and neutrons are deeply penetrating (>3 cm) in all materials, the apertures used to achieve the desired 10-μm resolution are 20-cm long, triangular tapers machined in gold foils. These gold foils are stacked to form an array of 20 apertures for pinhole imaging and three apertures for penumbral imaging. These apertures must be precisely aligned to accurately place the field of view of each aperture at the design location, or the location of the field of view for each aperture must be measured. In this paper we present a new technique that has been developed for the measurement and characterization of the precise location of each aperture in the array. We present the detailed algorithms used for this characterization and the results of reconstructed sources from inertial confinement fusion implosion experiments at NIF.

  8. Self characterization of a coded aperture array for neutron source imaging

    SciTech Connect

    Volegov, P. L. Danly, C. R.; Guler, N.; Merrill, F. E.; Wilde, C. H.; Fittinghoff, D. N.

    2014-12-15

    The neutron imaging system at the National Ignition Facility (NIF) is an important diagnostic tool for measuring the two-dimensional size and shape of the neutrons produced in the burning deuterium-tritium plasma during the stagnation stage of inertial confinement fusion implosions. Since the neutron source is small (∼100 μm) and neutrons are deeply penetrating (>3 cm) in all materials, the apertures used to achieve the desired 10-μm resolution are 20-cm long, triangular tapers machined in gold foils. These gold foils are stacked to form an array of 20 apertures for pinhole imaging and three apertures for penumbral imaging. These apertures must be precisely aligned to accurately place the field of view of each aperture at the design location, or the location of the field of view for each aperture must be measured. In this paper we present a new technique that has been developed for the measurement and characterization of the precise location of each aperture in the array. We present the detailed algorithms used for this characterization and the results of reconstructed sources from inertial confinement fusion implosion experiments at NIF.

  9. ITS version 5.0 : the integrated TIGER series of coupled electron/photon Monte Carlo transport codes.

    SciTech Connect

    Franke, Brian Claude; Kensek, Ronald Patrick; Laub, Thomas William

    2004-06-01

    ITS is a powerful and user-friendly software package permitting state of the art Monte Carlo solution of linear time-independent couple electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 5.0, the latest version of ITS, contains (1) improvements to the ITS 3.0 continuous-energy codes, (2)multigroup codes with adjoint transport capabilities, and (3) parallel implementations of all ITS codes. Moreover the general user friendliness of the software has been enhanced through increased internal error checking and improved code portability.

  10. An overview of the ENEA activities in the field of coupled codes NPP simulation

    SciTech Connect

    Parisi, C.; Negrenti, E.; Sepielli, M.; Del Nevo, A.

    2012-07-01

    In the framework of the nuclear research activities in the fields of safety, training and education, ENEA (the Italian National Agency for New Technologies, Energy and the Sustainable Development) is in charge of defining and pursuing all the necessary steps for the development of a NPP engineering simulator at the 'Casaccia' Research Center near Rome. A summary of the activities in the field of the nuclear power plants simulation by coupled codes is here presented with the long term strategy for the engineering simulator development. Specifically, results from the participation in international benchmarking activities like the OECD/NEA 'Kalinin-3' benchmark and the 'AER-DYN-002' benchmark, together with simulations of relevant events like the Fukushima accident, are here reported. The ultimate goal of such activities performed using state-of-the-art technology is the re-establishment of top level competencies in the NPP simulation field in order to facilitate the development of Enhanced Engineering Simulators and to upgrade competencies for supporting national energy strategy decisions, the nuclear national safety authority, and the R and D activities on NPP designs. (authors)

  11. Present development of TRANSFUSION, a novel time-dependent three-dimensional finite-element neutron and gamma diffusion code for the analysis of pulsed-neutron oil-well logs

    SciTech Connect

    De Raedt, C.M.; Ruan, D.

    1994-12-31

    The novel code TRANSFUSION is being developed with a view to analysing borehole logs in the field of oil exploration and production. TRANSFUSION describes the time-dependent neutron and gamma populations in the logging tool and the surrounding formation resulting from bursts of high energy neutrons by providing first-order solutions to the Boltzmann transport equation. In particular, the gamma responses in the near and the far detectors housed in the logging tool are calculated. TRANSFUSION was developed from the ANSWERS steady-state (three-dimensional) finite-element neutron and gamma diffusion code FENDER by extending it to become time-dependent. The code is an attractive alternative to the generally used rigorous Monte Carlo methods, being as user friendly but less time-consuming while remaining sufficiently accurate.

  12. TREND code for reconstruction of neutron reaction cross sections and calculation of the Doppler broadening in the region of resolved resonances

    SciTech Connect

    Chichulin, N. L. Kulakov, A. S.

    2010-12-15

    The TREND code is developed for reconstruction of neutron cross sections in the region of resolved resonances from the data of the international evaluated nuclear data libraries using the Reich-Moore, Adler-Adler, and Breit-Wigner formalisms and for calculation of the Doppler broadening of the resulting cross sections within the classical approximation. The TREND code is incorporated into the MCU code package for Monte Carlo reactor calculations. The TREND and MCU codes have been used to evaluate the data of the ROSFOND data bank on uranium and plutonium isotopes in comparison with the data of the ICSBEP data bank.

  13. Three-Dimensional (X,Y,Z) Deterministic Analysis of the PCA-Replica Neutron Shielding Benchmark Experiment using the TORT-3.2 Code and Group Cross Section Libraries for LWR Shielding and Pressure Vessel Dosimetry

    NASA Astrophysics Data System (ADS)

    Pescarini, Massimo; Orsi, Roberto; Frisoni, Manuela

    2016-02-01

    The PCA-Replica 12/13 (H2O/Fe) neutron shielding benchmark experiment was analysed using the ORNL TORT-3.2 3D SN code. PCA-Replica, specifically conceived to test the accuracy of nuclear data and transport codes employed in LWR shielding and radiation damage calculations, reproduces a PWR ex-core radial geometry with alternate layers of water and steel including a PWR pressure vessel simulator. Three broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format with the same energy group structure (47 n + 20 γ) and based on different nuclear data were alternatively used: the ENEA BUGJEFF311.BOLIB (JEFF-3.1.1) and BUGENDF70.BOLIB (ENDF/B-VII.0) libraries and the ORNL BUGLE-96 (ENDF/B-VI.3) library. Dosimeter cross sections derived from the IAEA IRDF-2002 dosimetry file were employed. The calculated reaction rates for the Rh-103(n,n')Rh-103 m, In-115(n,n')In-115m and S-32(n,p)P-32 threshold activation dosimeters and the calculated neutron spectra are compared with the corresponding experimental results.

  14. Tandem Mirror Reactor Systems Code (Version I)

    SciTech Connect

    Reid, R.L.; Finn, P.A.; Gohar, M.Y.; Barrett, R.J.; Gorker, G.E.; Spampinaton, P.T.; Bulmer, R.H.; Dorn, D.W.; Perkins, L.J.; Ghose, S.

    1985-09-01

    A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost.

  15. Experimental investigation of neutronic characteristics of the IR-8 reactor to confirm the results of calculations by MCU-PTR code

    SciTech Connect

    Surkov, A. V. Kochkin, V. N.; Pesnya, Yu. E.; Nasonov, V. A.; Vihrov, V. I.; Erak, D. Yu.

    2015-12-15

    A comparison of measured and calculated neutronic characteristics (fast neutron flux and fission rate of {sup 235}U) in the core and reflector of the IR-8 reactor is presented. The irradiation devices equipped with neutron activation detectors were prepared. The determination of fast neutron flux was performed using the {sup 54}Fe (n, p) and {sup 58}Ni (n, p) reactions. The {sup 235}U fission rate was measured using uranium dioxide with 10% enrichment in {sup 235}U. The determination of specific activities of detectors was carried out by measuring the intensity of characteristic gamma peaks using the ORTEC gamma spectrometer. Neutron fields in the core and reflector of the IR-8 reactor were calculated using the MCU-PTR code.

  16. Experimental investigation of neutronic characteristics of the IR-8 reactor to confirm the results of calculations by MCU-PTR code

    NASA Astrophysics Data System (ADS)

    Surkov, A. V.; Kochkin, V. N.; Pesnya, Yu. E.; Nasonov, V. A.; Vihrov, V. I.; Erak, D. Yu.

    2015-12-01

    A comparison of measured and calculated neutronic characteristics (fast neutron flux and fission rate of 235U) in the core and reflector of the IR-8 reactor is presented. The irradiation devices equipped with neutron activation detectors were prepared. The determination of fast neutron flux was performed using the 54Fe ( n, p) and 58Ni ( n, p) reactions. The 235U fission rate was measured using uranium dioxide with 10% enrichment in 235U. The determination of specific activities of detectors was carried out by measuring the intensity of characteristic gamma peaks using the ORTEC gamma spectrometer. Neutron fields in the core and reflector of the IR-8 reactor were calculated using the MCU-PTR code.

  17. Typical BWR/4 MSIV closure ATWS analysis using RAMONA-3B code with space-time neutron kinetics

    SciTech Connect

    Neymotin, L.; Saha, P.

    1984-01-01

    A best-estimate analysis of a typical BWR/4 MSIV closure ATWS has been performed using the RAMONA-3B code with three-dimensional neutron kinetics. All safety features, namely, the safety and relief valves, recirculation pump trip, high pressure safety injections and the standby liquid control system (boron injection), were assumed to work as designed. No other operator action was assumed. The results show a strong spatial dependence of reactor power during the transient. After the initial peak of pressure and reactor power, the reactor vessel pressure oscillated between the relief valve set points, and the reactor power oscillated between 20 to 50% of the steady state power until the hot shutdown condition was reached at approximately 1400 seconds. The suppression pool bulk water temperature at this time was predicted to be approx. 96/sup 0/C (205/sup 0/F). In view of code performance and reasonable computer running time, the RAMONA-3B code is recommended for further best-estimate analyses of ATWS-type events in BWRs.

  18. Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data.

    Energy Science and Technology Software Center (ESTSC)

    2000-03-28

    Version 00 The NJOY nuclear data processing system is a modular computer code used for converting evaluated nuclear data in the ENDF format into libraries useful for applications calculations. Because the Evaluated Nuclear Data File (ENDF) format is used all around the world (e.g., ENDF/B‑VI in the US, JEF‑2.2 in Europe, JENDL‑3.2 in Japan, BROND‑2.2 in Russia), NJOY gives its users access to a wide variety of the most up‑to‑date nuclear data. NJOY provides comprehensivemore » capabilities for processing evaluated data, and it can serve applications ranging from continuous‑energy Monte Carlo (MCNP), through deterministic transport codes (DANT, ANISN, DORT), to reactor lattice codes (WIMS, EPRI). NJOY handles a wide variety of nuclear effects, including resonances, Doppler broadening, heating (KERMA), radiation damage, thermal scattering (even cold moderators), gas production, neutrons and charged particles, photoatomic interactions, self shielding, probability tables, photon production, and high‑energy interactions (to 150 MeV). Output can include printed listings, special library files for applications, and Postscript graphics (plus color).« less

  19. A preliminary neutronic evaluation of high temperature engineering test reactor using the SCALE6 code

    NASA Astrophysics Data System (ADS)

    Tanure, L. P. A. R.; Sousa, R. V.; Costa, D. F.; Cardoso, F.; Veloso, M. A. F.; Pereira, C.

    2014-02-01

    Neutronic parameters of some fourth generation nuclear reactors have been investigated at the Departamento de Engenharia Nuclear/UFMG. Previous studies show the possibility to increase the transmutation capabilities of these fourth generation systems to achieve significant reduction concerning transuranic elements in spent fuel. To validate the studies, a benchmark on core physics analysis, related to initial testing of the High Temperature Engineering Test Reactor and provided by International Atomic Energy Agency (IAEA) was simulated using the Standardized Computer Analysis for Licensing Evaluation (SCALE). The CSAS6/KENO-VI control sequence and the 44-group ENDF/B-V 0 cross-section neutron library were used to evaluate the keff (effective multiplication factor) and the result presents good agreement with experimental value.

  20. Ensemble Activation of G-Protein -Coupled Receptors Revealed by Small-Angle Neutron Scattering

    NASA Astrophysics Data System (ADS)

    Chu, Xiang-Qiang; Perera, Suchithranga; Shrestha, Utsab; Chawla, Udeep; Struts, Andrey; Qian, Shuo; Brown, Michael

    2014-03-01

    Rhodopsin is a G-protein -coupled receptor (GPCR) involved in visual light perception and occurs naturally in a membrane lipid environment. Rhodopsin photoactivation yields cis-trans isomerization of retinal giving equilibrium between inactive Meta-I and active Meta-II states. Does photoactivation lead to a single Meta-II conformation, or do substates exist as described by an ensemble-activation mechanism (EAM)? We use small-angle neutron scattering (SANS) to investigate conformational changes in rhodopsin-detergent and rhodopsin-lipid complexes upon photoactivation. Meta-I state is stabilized in CHAPS-solubilized rhodopsin, while Meta-II is trapped in DDM-solubilized rhodopsin. SANS data are acquired from 80% D2O solutions and at contrast-matching points for both DDM and CHAPS samples. Our experiments demonstrate that for detergent-solubilized rhodopsin, SANS with contrast variation can detect structural differences between the rhodopsin dark-state, Meta-I, Meta-II, and ligand-free opsin states. Dark-state rhodopsin has more conformational flexibility in DDM micelles compared to CHAPS, which is consistent with an ensemble of activated Meta-II states. Furthermore, time-resolved SANS enables study of the time-dependent structural transitions between Meta-I and Meta-II, which is crucial to understanding the ensemble-based activation.

  1. Unsteady Analysis of Inlet-Compressor Acoustic Interactions Using Coupled 3-D and 1-D CFD Codes

    NASA Technical Reports Server (NTRS)

    Suresh, A.; Cole, G. L.

    2000-01-01

    It is well known that the dynamic response of a mixed compression supersonic inlet is very sensitive to the boundary condition imposed at the subsonic exit (engine face) of the inlet. In previous work, a 3-D computational fluid dynamics (CFD) inlet code (NPARC) was coupled at the engine face to a 3-D turbomachinery code (ADPAC) simulating an isolated rotor and the coupled simulation used to study the unsteady response of the inlet. The main problem with this approach is that the high fidelity turbomachinery simulation becomes prohibitively expensive as more stages are included in the simulation. In this paper, an alternative approach is explored, wherein the inlet code is coupled to a lesser fidelity 1-D transient compressor code (DYNTECC) which simulates the whole compressor. The specific application chosen for this evaluation is the collapsing bump experiment performed at the University of Cincinnati, wherein reflections of a large-amplitude acoustic pulse from a compressor were measured. The metrics for comparison are the pulse strength (time integral of the pulse amplitude) and wave form (shape). When the compressor is modeled by stage characteristics the computed strength is about ten percent greater than that for the experiment, but the wave shapes are in poor agreement. An alternate approach that uses a fixed rise in duct total pressure and temperature (so-called 'lossy' duct) to simulate a compressor gives good pulse shapes but the strength is about 30 percent low.

  2. Development of a finite element code to solve thermo-hydro-mechanical coupling and simulate induced seismicity.

    NASA Astrophysics Data System (ADS)

    María Gómez Castro, Berta; De Simone, Silvia; Rossi, Riccardo; Larese De Tetto, Antonia; Carrera Ramírez, Jesús

    2015-04-01

    Coupled thermo-hydro-mechanical modeling is essential for CO2 storage because of (1) large amounts of CO2 will be injected, which will cause large pressure buildups and might compromise the mechanical stability of the caprock seal, (2) the most efficient technique to inject CO2 is the cold injection, which induces thermal stress changes in the reservoir and seal. These stress variations can cause mechanical failure in the caprock and can also trigger induced earthquakes. To properly assess these effects, numerical models that take into account the short and long-term thermo-hydro-mechanical coupling are an important tool. For this purpose, there is a growing need of codes that couple these processes efficiently and accurately. This work involves the development of an open-source, finite element code written in C ++ for correctly modeling the effects of thermo-hydro-mechanical coupling in the field of CO2 storage and in others fields related to these processes (geothermal energy systems, fracking, nuclear waste disposal, etc.), and capable to simulate induced seismicity. In order to be able to simulate earthquakes, a new lower dimensional interface element will be implemented in the code to represent preexisting fractures, where pressure continuity will be imposed across the fractures.

  3. Testing and Modeling of a 3-MW Wind Turbine Using Fully Coupled Simulation Codes (Poster)

    SciTech Connect

    LaCava, W.; Guo, Y.; Van Dam, J.; Bergua, R.; Casanovas, C.; Cugat, C.

    2012-06-01

    This poster describes the NREL/Alstom Wind testing and model verification of the Alstom 3-MW wind turbine located at NREL's National Wind Technology Center. NREL,in collaboration with ALSTOM Wind, is studying a 3-MW wind turbine installed at the National Wind Technology Center(NWTC). The project analyzes the turbine design using a state-of-the-art simulation code validated with detailed test data. This poster describes the testing and the model validation effort, and provides conclusions about the performance of the unique drive train configuration used in this wind turbine. The 3-MW machine has been operating at the NWTC since March 2011, and drive train measurements will be collected through the spring of 2012. The NWTC testing site has particularly turbulent wind patterns that allow for the measurement of large transient loads and the resulting turbine response. This poster describes the 3-MW turbine test project, the instrumentation installed, and the load cases captured. The design of a reliable wind turbine drive train increasingly relies on the use of advanced simulation to predict structural responses in a varying wind field. This poster presents a fully coupled, aero-elastic and dynamic model of the wind turbine. It also shows the methodology used to validate the model, including the use of measured tower modes, model-to-model comparisons of the power curve, and mainshaft bending predictions for various load cases. The drivetrain is designed to only transmit torque to the gearbox, eliminating non-torque moments that are known to cause gear misalignment. Preliminary results show that the drivetrain is able to divert bending loads in extreme loading cases, and that a significantly smaller bending moment is induced on the mainshaft compared to a three-point mounting design.

  4. Development of a novel neutron detection technique by using a boron layer coating a Charge Coupled Device

    SciTech Connect

    Blostein, Juan Jerónimo; Estrada, Juan; Tartaglione, Aureliano; Sofo haro, Miguel; Fernández Moroni, Guillermo; Cancelo, Gustavo

    2015-01-19

    This article describes the design features and the first test measurements obtained during the installation of a novel high resolution 2D neutron detection technique. The technique proposed in this work consists of a boron layer (enriched in ${^{10}}$B) placed on a scientific Charge Coupled Device (CCD). After the nuclear reaction ${^{10}}$B(n,$\\alpha$)${^{7}}$Li, the CCD detects the emitted charge particles thus obtaining information on the neutron absorption position. The above mentioned ionizing particles, with energies in the range 0.5-5.5 MeV, produce a plasma effect in the CCD which is recorded as a circular spot. This characteristic circular shape, as well as the relationship observed between the spot diameter and the charge collected, is used for the event recognition, allowing the discrimination of undesirable gamma events. We present the first results recently obtained with this technique, which has the potential to perform neutron tomography investigations with a spatial resolution better than that previously achieved. Numerical simulations indicate that the spatial resolution of this technique will be about 15 $\\mu$m, and the intrinsic detection efficiency for thermal neutrons will be about 3 %. We compare the proposed technique with other neutron detection techniques and analyze its advantages and disadvantages.

  5. Dyadic wavelet for image coding implementation on a Xilinx MicroBlaze processor: application to neutron radiography.

    PubMed

    Saadi, Slami; Touiza, Maamar; Kharfi, Fayçal; Guessoum, Abderrezak

    2013-12-01

    In this work, we present a mixed software/hardware implementation of 2-D signals encoder/decoder using dyadic discrete wavelet transform (DWT) based on quadrature mirror filters (QMF); using fast wavelet Mallat's algorithm. This work is designed and compiled on the embedded development kit EDK6.3i, and the synthesis software, ISE6.3i, which is available with Xilinx Virtex-IIV2MB1000 FPGA. Huffman coding scheme is used to encode the wavelet coefficients so that they can be transmitted progressively through an Ethernet TCP/IP based connection. The possible reconfiguration can be exploited to attain higher performance. The design will be integrated with the neutron radiography system that is used with the Es-Salem research reactor. PMID:24041807

  6. Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data.

    Energy Science and Technology Software Center (ESTSC)

    1995-06-01

    Version 04 The NJOY nuclear data processing system is a comprehensive computer code package for producing pointwise and multigroup neutron and photon cross sections from ENDF/B evaluated nuclear data. This is the last NJOY-91 series. It uses the same module structure as the earlier versions and its graphics options depend on DISSPLA. This new release, designated NJOY91.119, includes bug fixes, improvements in several modules, and some new capabilities. Information on the changes is included inmore » the README file. A new test problem was added to test some ENDF-6 features, including Reich-Moore resonance reconstruction, energy-angle matrices in GROUPR, and energy-angle distributions in ACER. The 91.119 release is basically configured for UNIX.« less

  7. Couple attachment and the quality of marital relationships: method and concept in the validation of the new couple attachment interview and coding system.

    PubMed

    Alexandrov, Elina O; Cowan, Philip A; Cowan, Carolyn Pape

    2005-06-01

    This study investigates links between adult attachment and marital quality in 73 married couples, using a new Couple Attachment Interview that was modeled after the Adult Attachment Interview but focuses on the relationship between the partners. A coding system (CAICS) comparing the interview protocol to prototypes for secure, dismissing, and preoccupied attachment styles yielded continuous ratings of all three styles, and categorical classifications of secure/insecure for each partner. The study found direct links between couple attachment and both self-reported and observed marital quality, with all three continuous scores contributing uniquely to the equations. In most cases, the continuous scores explained variation in marital quality after the categorical security scores were entered into the regressions, although categorical scores also contributed uniquely to the explanation of marital quality. Pairing of partners' scores explained significant variance in both self-reported and observed evaluations of the couple relationship. Security of couple attachment served as a mediator in the link between self-reported marital satisfaction and observed marital quality. The results illustrated the interconnection of methodological choices and theoretical advances in the study of attachment and couple relationship quality. PMID:16096190

  8. Three-Dimensional, Nodal, Neutron Diffusion Criticality Code System in Hex-Z Geometry.

    Energy Science and Technology Software Center (ESTSC)

    1992-07-27

    Version: 00 SIXTUS-3 is a 3D extention of SIXTUS-2 and is based on a response matrix nodal model. The code offers a fast and accurate analysis of critical systems in the regular hex-z geometry with the multigroup cross section representation including arbitrary upscattering.

  9. A Two-Dimensional Monte Carlo Code System for Linear Neutron Transport Calculations.

    Energy Science and Technology Software Center (ESTSC)

    1980-04-24

    Version 00 KIM (k-infinite-Monte Carlo) solves the steady-state linear neutron transport equation for a fixed source problem or, by successive fixed-source runs, for the eigenvalue problem, in a two-dimensional infinite thermal reactor lattice using the Monte Carlo method. In addition to the combinatorial description of domains, the program allows complex configurations to be represented by a discrete set of points whereby the calculation speed is greatly improved. Configurations are described as the result of overlaysmore » of elementary figures over a basic domain.« less

  10. Code System for Two-Dinensional Sn-Neutronics and Fluid Dynamics.

    Energy Science and Technology Software Center (ESTSC)

    2003-07-28

    Version 00 SIMMERII is designed to predict the neutronic and fluid-dynamic behavior of an LMFBR during a hypothetical core-disruptive accident. Cross sections depend on temperature and background cross sections. The structure, liquid, and vapor fields are modeled to predict the fluid-dynamic behavior of the reactor. Each field consists of density components to follow the material motion and energy components to predict the material temperatures. For typical accident calculations, the materials are fertile fuel, fissile fuel,more » stainless steel, sodium, control material, and fission gas. Heat, mass, and momentum transfer among the three fields and their components are calculated.« less

  11. Evaluation of the influence of the main plasma density parameters on antenna coupling and radio frequency potentials with TOPICA code

    NASA Astrophysics Data System (ADS)

    Milanesio, D.; Maggiora, R.

    2013-04-01

    The successful design of an ion cyclotron antenna mainly relies on the capability of accurately predicting its behavior both in terms of input parameters, and therefore power coupled to plasma, and radiated fields. All these features essentially depend on the antenna itself (its geometry, the matching and tuning systems) and, obviously, on the faced loading. In this paper a number of plasma profiles is analysed with the help of the TOPICA code, a predictive tool for the design and optimization of radio frequency (RF) launchers in front of a plasma, in order to understand which plasma parameters have the most significant influence on the coupling performances of a typical IC antenna.

  12. Development code for sensitivity and uncertainty analysis of input on the MCNPX for neutronic calculation in PWR core

    SciTech Connect

    Hartini, Entin Andiwijayakusuma, Dinan

    2014-09-30

    This research was carried out on the development of code for uncertainty analysis is based on a statistical approach for assessing the uncertainty input parameters. In the butn-up calculation of fuel, uncertainty analysis performed for input parameters fuel density, coolant density and fuel temperature. This calculation is performed during irradiation using Monte Carlo N-Particle Transport. The Uncertainty method based on the probabilities density function. Development code is made in python script to do coupling with MCNPX for criticality and burn-up calculations. Simulation is done by modeling the geometry of PWR terrace, with MCNPX on the power 54 MW with fuel type UO2 pellets. The calculation is done by using the data library continuous energy cross-sections ENDF / B-VI. MCNPX requires nuclear data in ACE format. Development of interfaces for obtaining nuclear data in the form of ACE format of ENDF through special process NJOY calculation to temperature changes in a certain range.

  13. User's guide for the computer code COLTS for calculating the coupled laminar and turbulent flow over a Jovian entry probe

    NASA Technical Reports Server (NTRS)

    Kumar, A.; Graeves, R. A.

    1980-01-01

    A user's guide for a computer code 'COLTS' (Coupled Laminar and Turbulent Solutions) is provided which calculates the laminar and turbulent hypersonic flows with radiation and coupled ablation injection past a Jovian entry probe. Time-dependent viscous-shock-layer equations are used to describe the flow field. These equations are solved by an explicit, two-step, time-asymptotic finite-difference method. Eddy viscosity in the turbulent flow is approximated by a two-layer model. In all, 19 chemical species are used to describe the injection of carbon-phenolic ablator in the hydrogen-helium gas mixture. The equilibrium composition of the mixture is determined by a free-energy minimization technique. A detailed frequency dependence of the absorption coefficient for various species is considered to obtain the radiative flux. The code is written for a CDC-CYBER-203 computer and is capable of providing solutions for ablated probe shapes also.

  14. Electromagnetic pulse (EMP) coupling codes for use with the vulnerability/lethality (VIL) taxonomy. Final report, June-October 1984

    SciTech Connect

    Mar, M.H.

    1995-07-01

    Based on the vulnerability Lethality (V/L) taxonomy developed by the Ballistic Vulnerability Lethality Division (BVLD) of the Survivability Lethality Analysis Directorate (SLAD), a nuclear electromagnetic pulse (EMP) coupling V/L analysis taxonomy has been developed. A nuclear EMP threat to a military system can be divided into two levels: (1) coupling to a system level through a cable, antenna, or aperture; and (2) the component level. This report will focus on the initial condition, which includes threat definition and target description, as well as the mapping process from the initial condition to damaged components state. EMP coupling analysis at a system level is used to accomplish this. This report introduces the nature of EMP threat, interaction between the threat and target, and how the output of EMP coupling analysis at a system level becomes the input to the component level analysis. Many different tools (EMP coupling codes) will be discussed for the mapping process, which correponds to the physics of phenomenology. This EMP coupling V/L taxonomy and the models identified in this report will provide the tools necessary to conduct basic V/L analysis of EMP coupling.

  15. A review on the relativistic effective field theory with parameterized couplings for nuclear matter and neutron stars

    NASA Astrophysics Data System (ADS)

    Vasconcellos, C. A. Zen

    2015-12-01

    Nuclear science has developed many excellent theoretical models for many-body systems in the domain of the baryon-meson strong interaction for the nucleus and nuclear matter at low, medium and high densities. However, a full microscopic understanding of nuclear systems in the extreme density domain of compact stars is still lacking. The aim of this contribution is to shed some light on open questions facing the nuclear many-body problem at the very high density domain. Here we focus our attention on the conceptual issue of naturalness and its role in shaping the baryon-meson phase space dynamics in the description of the equation of state (EoS) of nuclear matter and neutrons stars. In particular, in order to stimulate possible new directions of research, we discuss relevant aspects of a recently developed relativistic effective theory for nuclear matter within Quantum Hadrodynamics (QHD) with genuine many-body forces and derivative natural parametric couplings. Among other topics we discuss in this work the connection of this theory with other known effective QHD models of the literature and its potentiality in describing a new physics for dense matter. The model with parameterized couplings exhausts the whole fundamental baryon octet (n, p, Σ-, Σ0, Σ+, Λ, Ξ-, Ξ0) and simulates n-order corrections to the minimal Yukawa baryon couplings by considering nonlinear self-couplings of meson fields and meson-meson interaction terms coupled to the baryon fields involving scalar-isoscalar (σ, σ∗), vector-isoscalar (ω, ɸ), vector-isovector (ϱ) and scalar-isovector (δ) virtual sectors. Following recent experimental results, we consider in our calculations the extreme case where the Σ- experiences such a strong repulsion that its influence in the nuclear structure of a neutron star is excluded at all. A few examples of calculations of properties of neutron stars are shown and prospects for the future are discussed.

  16. European Pressurized water Reactor (EPR) SAR ATWS Accident Analyses by using 3D Code Internal Coupling Method

    SciTech Connect

    Gagner, Renata; Lafitte, Helene; Dormeau, Pascal; Stoudt, Roger H.

    2004-07-01

    Anticipated Transients Without Scram (ATWS) accident analyses make part of the Safety Analysis Report of the European Pressurized water Reactor (EPR), covering Risk Reduction Category A (Core Melt Prevention) events. This paper deals with three of the most penalizing RRC-A sequences of ATWS caused by mechanical blockage of the control/shutdown rods, regarding their consequences on the Reactor Coolant System (RCS) and core integrity. A new 3D code internal coupling calculation method has been introduced. (authors)

  17. Benchmark of Lower Hybrid coupling codes (Brambilla, GRILL3D-U, TOPLHA) with the FTU conventional grill experimental data

    NASA Astrophysics Data System (ADS)

    Ceccuzzi, Silvio; Maggiora, Riccardo; Milanesio, Daniele; Mirizzi, Francesco; Panaccione, Luigi

    2011-12-01

    The present work compares and experimentally validates the results coming out from the following three Lower Hybrid (LH) coupling codes: Brambilla code (M. Brambilla), GRILL3D-U (Mikhail Irzak, A. F. Ioffe Physico-Technical Institute, Russia) and TOPLHA (Politecnico di Torino, Italy). The conventional grill antenna, operating in FTU in different scenarios, is used as benchmark. The validation with experimental data is carried out with respect to the average reflection coefficients at the input of a row of the grill, considering two different phasings between adjacent waveguides: -90 ° and -75 °. A comparison between calculated power spectra is also presented. Good agreement can be observed for all the simulated plasma profiles and waveguide phasings between experimental data and codes, in particular for the most recent numerical tools, namely GRILL3D-U and TOPLHA.

  18. Coupling functions for lead and lead-free neutron monitors from the latitudinal measurements performed in 1982 in the research station Academician Kurchatov

    NASA Technical Reports Server (NTRS)

    Alekanyan, T. M.; Dorman, L. I.; Yanke, V. G.; Korotkov, V. K.

    1985-01-01

    The latitudinal behavior of intensities and multiplicities was registered by the neutron monitor 2 NM and the lead-free neutron monitor 3 SND (slow-neuron detector) in the equator-Kaliningrad line in the Atlantic Ocean. Coupling coefficients for 3 SND show the sensitivity of this detector to primary particles of cosmic rays of energies on the average lower than for 2 NM. As multiplicities increase, the coupling coefficients shift towards higher energies.

  19. Relativistic mean-field models with scaled hadron masses and couplings: Hyperons and maximum neutron star mass

    NASA Astrophysics Data System (ADS)

    Maslov, K. A.; Kolomeitsev, E. E.; Voskresensky, D. N.

    2016-06-01

    An equation of state of cold nuclear matter with an arbitrary isotopic composition is studied within a relativistic mean-field approach with hadron masses and coupling constants depending self-consistently on the scalar mean-field. All hadron masses decrease universally with the scalar field growth, whereas meson-nucleon coupling constants can vary differently. More specifically we focus on two modifications of the KVOR model studied previously. One extension of the model (KVORcut) demonstrates that the equation of state stiffens if the increase of the scalar-field magnitude with the density is bounded from above at some value for baryon densities above the saturation nuclear density. This can be realized if the nucleon vector-meson coupling constant changes rapidly as a function of the scalar field slightly above the desired value. The other version of the model (MKVOR) utilizes a smaller value of the nucleon effective mass at the nuclear saturation density and a saturation of the scalar field in the isospin asymmetric matter induced by a strong variation of the nucleon isovector-meson coupling constant as function of the scalar field. A possibility of hyperonization of the matter in neutron star interiors is incorporated. Our equations of state fulfill majority of known empirical constraints including the pressure-density constraint from heavy-ion collisions, direct Urca constraint, gravitational-baryon mass constraint for the pulsar J0737-3039B, and the constraint on the maximum mass of the neutron stars.

  20. Specification of the Advanced Burner Test Reactor Multi-Physics Coupling Demonstration Problem

    SciTech Connect

    Shemon, E. R.; Grudzinski, J. J.; Lee, C. H.; Thomas, J. W.; Yu, Y. Q.

    2015-12-21

    This document specifies the multi-physics nuclear reactor demonstration problem using the SHARP software package developed by NEAMS. The SHARP toolset simulates the key coupled physics phenomena inside a nuclear reactor. The PROTEUS neutronics code models the neutron transport within the system, the Nek5000 computational fluid dynamics code models the fluid flow and heat transfer, and the DIABLO structural mechanics code models structural and mechanical deformation. The three codes are coupled to the MOAB mesh framework which allows feedback from neutronics, fluid mechanics, and mechanical deformation in a compatible format.

  1. A model for stationary and dynamic impression of undercooled boiling in coupled thermohydraulic and neutron physics calculations of nuclear reactors

    NASA Astrophysics Data System (ADS)

    Mueller, Roland Guenther

    1987-06-01

    In order to account for subcooled boiling in calculations of neutron physics and thermal hydraulics of light water reactors (where vapor bubbles strongly influence the nuclear chain reaction), a dynamic model is derived from the time-dependent conservation equations. It contains methods for the time-dependent determination of evaporation and condensation heat flow and for the heat transfer coefficient in subcooled boiling. It enables the complete two-phase flow region to be treated consistently. The calculation model was verified using measured data of experiments covering a wide range of thermodynamic boundary conditions. In all cases very good agreement is reached. The results from the coupling of the new calculation model with a neutron kinetics program proves its suitability for the steady-state and transient calculation of reactor cores.

  2. Water dynamics in hectorite clays: influence of temperature studied by coupling neutron spin echo and molecular dynamics.

    PubMed

    Marry, Virginie; Dubois, Emmanuelle; Malikova, Natalie; Durand-Vidal, Serge; Longeville, Stéphane; Breu, Josef

    2011-04-01

    Within the wider context of water behavior in soils, and with a particular emphasis on clays surrounding underground radioactive waste packages, we present here the translational dynamics of water in clays in low hydrated states as studied by coupling molecular dynamics (MD) simulations and quasielastic neutron scattering experiments by neutron spin echo (NSE). A natural montmorillonite clay of interest is modeled by a synthetic clay which allows us to understand the determining parameters from MD simulations by comparison with the experimental values. We focus on temperatures between 300 and 350 K, i.e., the range relevant to the highlighted application. The activation energy Ea experimentally determined is 6.6 kJ/mol higher than that for bulk water. Simulations are in good agreement with experiments for the relevant set of conditions, and they give more insight into the origin of the observed dynamics. PMID:21381672

  3. A highly optimized code for calculating atomic data at neutron star magnetic field strengths using a doubly self-consistent Hartree-Fock-Roothaan method

    NASA Astrophysics Data System (ADS)

    Schimeczek, C.; Engel, D.; Wunner, G.

    2012-07-01

    Our previously published code for calculating energies and bound-bound transitions of medium-Z elements at neutron star magnetic field strengths [D. Engel, M. Klews, G. Wunner, Comput. Phys. Comm. 180 (2009) 302-311] was based on the adiabatic approximation. It assumes a complete decoupling of the (fast) gyration of the electrons under the action of the magnetic field and the (slow) bound motion along the field under the action of the Coulomb forces. For the single-particle orbitals this implied that each is a product of a Landau state and an (unknown) longitudinal wave function whose B-spline coefficients were determined self-consistently by solving the Hartree-Fock equations for the many-electron problem on a finite-element grid. In the present code we go beyond the adiabatic approximation, by allowing the transverse part of each orbital to be a superposition of Landau states, while assuming that the longitudinal part can be approximated by the same wave function in each Landau level. Inserting this ansatz into the energy variational principle leads to a system of coupled equations in which the B-spline coefficients depend on the weights of the individual Landau states, and vice versa, and which therefore has to be solved in a doubly self-consistent manner. The extended ansatz takes into account the back-reaction of the Coulomb motion of the electrons along the field direction on their motion in the plane perpendicular to the field, an effect which cannot be captured by the adiabatic approximation. The new code allows for the inclusion of up to 8 Landau levels. This reduces the relative error of energy values as compared to the adiabatic approximation results by typically a factor of three (1/3 of the original error), and yields accurate results also in regions of lower neutron star magnetic field strengths where the adiabatic approximation fails. Further improvements in the code are a more sophisticated choice of the initial wave functions, which takes into

  4. LCM: A new approach to parse XML documents through loose coupling between XML document design and the corresponding application code

    NASA Astrophysics Data System (ADS)

    Chandu, Vaddadi P.

    XML (eXtensible Markup Language) is a widely used language for communication between heterogeneous, homogeneous, and internet software applications. Currently available binding and parsing techniques for XML documents involve tight coupling of the document with the application code and require a redesign of the application code for any change in the document design. This paper proposes a 'Loosely Coupled Model (LCM)' for XML document parsing, which eliminates the need to redesign the application code for moderately large changes in the XML document design. LCM takes a token based approach, which is very different from the existing methodologies for XML parsing. LCM is implemented in Java using Sun's JDK 1.5. The dataset includes 24 XML documents categorized into two sets- (a) deeply nested and (b) non-deeply nested XMLs. Each set contains 12 documents ranging from 2 KB up to 4 MB. A performance comparison for LCM is obtained and is noticed to be comparable or better against this dataset with (a) Sun's Java XML Binding (JAXB), (b) Apache's XMLBeans, and (c) JDOM parser.

  5. Analysis of Inlet-Compressor Acoustic Interactions Using Coupled CFD Codes

    NASA Technical Reports Server (NTRS)

    Suresh, A.; Townsend, S. E.; Cole, G. L.; Slater, J. W.; Chima, R.

    1998-01-01

    A problem that arises in the numerical simulation of supersonic inlets is the lack of a suitable boundary condition at the engine face. In this paper, a coupled approach, in which the inlet computation is coupled dynamically to a turbomachinery computation, is proposed as a means to overcome this problem. The specific application chosen for validation of this approach is the collapsing bump experiment performed at the University of Cincinnati. The computed results are found to be in reasonable agreement with experimental results. The coupled simulation results could also be used to aid development of a simplified boundary condition.

  6. ANITA-2000 activation code package - updating of the decay data libraries and validation on the experimental data of the 14 MeV Frascati Neutron Generator

    NASA Astrophysics Data System (ADS)

    Frisoni, Manuela

    2016-03-01

    ANITA-2000 is a code package for the activation characterization of materials exposed to neutron irradiation released by ENEA to OECD-NEADB and ORNL-RSICC. The main component of the package is the activation code ANITA-4M that computes the radioactive inventory of a material exposed to neutron irradiation. The code requires the decay data library (file fl1) containing the quantities describing the decay properties of the unstable nuclides and the library (file fl2) containing the gamma ray spectra emitted by the radioactive nuclei. The fl1 and fl2 files of the ANITA-2000 code package, originally based on the evaluated nuclear data library FENDL/D-2.0, were recently updated on the basis of the JEFF-3.1.1 Radioactive Decay Data Library. This paper presents the results of the validation of the new fl1 decay data library through the comparison of the ANITA-4M calculated values with the measured electron and photon decay heats and activities of fusion material samples irradiated at the 14 MeV Frascati Neutron Generator (FNG) of the NEA-Frascati Research Centre. Twelve material samples were considered, namely: Mo, Cu, Hf, Mg, Ni, Cd, Sn, Re, Ti, W, Ag and Al. The ratios between calculated and experimental values (C/E) are shown and discussed in this paper.

  7. Neutron-Proton Coupling and the Lifetime of the First Excited State in ^16C

    NASA Astrophysics Data System (ADS)

    Fallon, Paul

    2008-04-01

    Nuclei near the valley of β-stability have strongly correlated proton and neutron spatial distributions. This need not be the case for nuclei with a large excess of one nucleon type and the search for new phenomena and structure effects due to the ``decoupling'' of neutrons and protons is of great interest in nuclear structure physics. Cited examples of decoupled behavior include neutron-halo nuclei with measurably different proton and neutron radial distributions, and low-energy dipole modes such as ``pygmy'' resonances where, simplistically, a core of equal numbers of protons and neutrons oscillates against the excess neutron ``skin'''. Recently, another example was suggested to occur in ^16C where the measurement of an anomalously quenched B(E2;2^+->^ 0^+) value of 0.63 e^2fm^4 combined with a large nuclear deformation led to the suggestion that the ^16C valence neutrons were decoupled from its near-spherical proton core (N.Imai et al., PRL 92 (2004) 062501; Z.Elekes et al., PLB 586 (2004) 34; H.J.Ong et al., PRC 73 (2006) 024610). In this talk I will discuss a new lifetime measurement for the first-excited 2^+ state in ^16C carried out at the LBNL 88-Inch Cyclotron using the Recoil Distance Method and ^9Be(^9Be,2p) fusion-evaporation reaction. The mean lifetime was found to be 11.7(20) ps corresponding to a B(E2) of 4.15(73) e^2fm^4, consistent with other even-even closed shell nuclei and neighboring systematics. Our result does not support the interpretation of decoupled protons and neutrons in ^16C. The revised value provides an important benchmark for theory. Time permitting I will present results on the neutron-rich nucleus ^30Ne produced in a 2p knockout reaction performed at the NSCL using the S800 spectrometer and SeGA gamma-ray detector. The measured (quenched) 2p knockout cross-section, when compared to theory, suggests a significant difference in the neutron intruder content between ^32Mg and ^30Ne, contrary to current shell models.

  8. A simulation of a capacitively coupled oxygen discharge using the oopd1 particle-in-cell Monte Carlo code

    NASA Astrophysics Data System (ADS)

    Gudmundsson, J. T.; Lieberman, M. A.; Wang, Ying; Verboncoeur, J. P.

    2009-10-01

    The oopd1 particle-in-cell Monte Carlo (PIC-MC) code is used to simulate a capacitively coupled discharge in oxygen. oopd1 is a one-dimensional object-oriented PIC-MC code [1] in which the model system has one spatial dimension and three velocity components. It contains models for planar, cylindrical, and spherical geometries and replaces the XPDx1 series [2], which is not object-oriented. The revised oxygen model includes, in addition to electrons, the oxygen molecule in ground state, the oxygen atom in ground state, the negative ion O^-, and the positive ions O^+ and O2^+. The cross sections for the collisions among the oxygen species have been significantly revised from earlier work using the xpdp1 code [3]. Here we explore the electron energy distribution function (EEDF), the ion energy distribution function (IEDF) and the density profiles for various pressures and driving frequencies. In particular we investigate the influence of the O^+ ion on the IEDF, we explore the influence of multiple driving frequencies, and we do comparisons to the previous xpdx1 codes. [1] J. P. Verboncoeur, A. B. Langdon, and N. T. Gladd, Comp. Phys. Comm. 87 (1995) 199 [2] J. P. Verboncoeur, M. V. Alves, V. Vahedi, and C. K. Birdsall, J. Comp. Physics 104 (1993) 321 [2] V. Vahedi and M. Surendra, Comp. Phys. Comm. 87 (1995) 179

  9. Validation of 3D Code KATRIN For Fast Neutron Fluence Calculation of VVER-1000 Reactor Pressure Vessel by Ex-Vessel Measurements and Surveillance Specimens Results

    NASA Astrophysics Data System (ADS)

    Dzhalandinov, A.; Tsofin, V.; Kochkin, V.; Panferov, P.; Timofeev, A.; Reshetnikov, A.; Makhotin, D.; Erak, D.; Voloschenko, A.

    2016-02-01

    Usually the synthesis of two-dimensional and one-dimensional discrete ordinate calculations is used to evaluate neutron fluence on VVER-1000 reactor pressure vessel (RPV) for prognosis of radiation embrittlement. But there are some cases when this approach is not applicable. For example the latest projects of VVER-1000 have upgraded surveillance program. Containers with surveillance specimens are located on the inner surface of RPV with fast neutron flux maximum. Therefore, the synthesis approach is not suitable enough for calculation of local disturbance of neutron field in RPV inner surface behind the surveillance specimens because of their complicated and heterogeneous structure. In some cases the VVER-1000 core loading consists of fuel assemblies with different fuel height and the applicability of synthesis approach is also ambiguous for these fuel cycles. Also, the synthesis approach is not enough correct for the neutron fluence estimation at the RPV area above core top. Because of these reasons only the 3D neutron transport codes seem to be satisfactory for calculation of neutron fluence on the VVER-1000 RPV. The direct 3D calculations are also recommended by modern regulations.

  10. Neutron-Diffraction Evidence for the Ferrimagnetic Ground State of a Molecule-Based Magnet with Weakly Coupled Sublattices

    SciTech Connect

    Fishman, Randy Scott; Campo, Javier; Vos, Thomas E.; Miller, Joel S.

    2012-01-01

    The diruthenium compound [Ru2(O2CMe)4]3[Cr(CN)6] contains two weakly coupled, ferrimag- netically ordered sublattices occupying the same volume. The magnetic field Hc 800 Oe required to align the two sublattice moments is proportional to the antiferromagnetic dipolar interaction Kc B Hc 5 10 3 meV between sublattices. Powder neutron-diffraction measurements on a deuterated sample reveal that the sublattice moments are restricted by the anisotropy of the diruthenium paddle-wheel complexes to the cubic diagonals. Those measurements also suggest that the quantum corrections to the ground state are significant.

  11. Stimulation at Desert Peak -modeling with the coupled THM code FEHM

    SciTech Connect

    kelkar, sharad

    2013-04-30

    Numerical modeling of the 2011 shear stimulation at the Desert Peak well 27-15. This submission contains the FEHM executable code for a 64-bit PC Windows-7 machine, and the input and output files for the results presented in the included paper from ARMA-213 meeting.

  12. Point Kernel Code System for Neutron and Gamma-Ray Shielding Calculations in Complex Geometry, Including a Graphical User Interface.

    Energy Science and Technology Software Center (ESTSC)

    2001-10-01

    Version 01 GUI2QAD is an aid in preparation of input for the included QAD-CGPIC program, which is based on CCC-493/QAD-CGGP and PICTURE. QAD-CGPIC is a Fortran code for fast neutron and gamma-ray shielding calculations through various shield configurations defined by combinatorial geometry specifications. Provision is available to interactively input the geometry and view the same in three dimensions with arbitrary rotations along x,y,z axis. The salient features of the present package include: a) Handles offmore » centered multiple identical sources b) Axis of cylindrical sources can be parallel to any of the axes. c) Provides plots of buildup factors (ANSI-1990) and material cross sections d) Estimates dose rate for point source-slab shield situations e) Interactive input of CG geometry with 3D view and rotation f) Fission product decay power computation and plots for source term calculations. g) Provision to read and graphical 1y display picture input file.« less

  13. Neutron production using a pyroelectric driven target coupled with a gated field ionization source

    SciTech Connect

    Ellsworth, J. L.; Tang, V.; Falabella, S.; Naranjo, B.; Putterman, S.

    2013-04-19

    A palm sized, portable neutron source would be useful for widespread implementation of detection systems for shielded, special nuclear material. We present progress towards the development of the components for an ultracompact neutron generator using a pulsed, meso-scale field ionization source, a deuterated (or tritiated) titanium target driven by a negative high voltage lithium tantalate crystal. Neutron production from integrated tests using an ion source with a single, biased tungsten tip and a 3 Multiplication-Sign 1 cm, vacuum insulated crystal with a plastic deuterated target are presented. Component testing of the ion source with a single tip produces up to 3 nA of current. Dielectric insulation of the lithium tantalate crystals appears to reduce flashover, which should improve the robustness. The field emission losses from a 3 cm diameter crystal with a plastic target and 6 cm diameter crystal with a metal target are compared.

  14. Near-barrier fusion of Sn+Ni and Te+Ni Systems: Examining the influence of neutron transfer couplings

    SciTech Connect

    Liang, J Felix; Kohley, Zachary W; Shapira, Dan; Varner Jr, Robert L; Gross, Carl J; Allmond, J M; Lagergren, Karin B; Mueller, Paul Edward

    2011-01-01

    The fusion excitation functions for radioactive 132Sn+58Ni and stable 130Te+58;64Ni were measured at energies near the Coulomb barrier. The role of transfer couplings in heavy-ion fusion was examined through a comparison of Sn+Ni and Te+Ni systems, which have large variations in the number of positive Q-value nucleon transfer channels. In contrast with previous comparisons, where increased sub-barrier fusion cross sections were observed in the systems with positive Q-value neutron transfer channels, the reduced excitation functions were equivalent for the different Sn+Ni and Te+Ni systems. The present results suggest a significant change in the influence of transfer couplings on the fusion process for the Sn+Ni and Te+Ni systems.

  15. A review on the relativistic effective field theory with parameterized couplings for nuclear matter and neutron stars

    SciTech Connect

    Vasconcellos, C. A. Zen

    2015-12-17

    Nuclear science has developed many excellent theoretical models for many-body systems in the domain of the baryon-meson strong interaction for the nucleus and nuclear matter at low, medium and high densities. However, a full microscopic understanding of nuclear systems in the extreme density domain of compact stars is still lacking. The aim of this contribution is to shed some light on open questions facing the nuclear many-body problem at the very high density domain. Here we focus our attention on the conceptual issue of naturalness and its role in shaping the baryon-meson phase space dynamics in the description of the equation of state (EoS) of nuclear matter and neutrons stars. In particular, in order to stimulate possible new directions of research, we discuss relevant aspects of a recently developed relativistic effective theory for nuclear matter within Quantum Hadrodynamics (QHD) with genuine many-body forces and derivative natural parametric couplings. Among other topics we discuss in this work the connection of this theory with other known effective QHD models of the literature and its potentiality in describing a new physics for dense matter. The model with parameterized couplings exhausts the whole fundamental baryon octet (n, p, Σ{sup −}, Σ{sup 0}, Σ{sup +}, Λ, Ξ{sup −}, Ξ{sup 0}) and simulates n-order corrections to the minimal Yukawa baryon couplings by considering nonlinear self-couplings of meson fields and meson-meson interaction terms coupled to the baryon fields involving scalar-isoscalar (σ, σ∗), vector-isoscalar (ω, Φ), vector-isovector (ϱ) and scalar-isovector (δ) virtual sectors. Following recent experimental results, we consider in our calculations the extreme case where the Σ{sup −} experiences such a strong repulsion that its influence in the nuclear structure of a neutron star is excluded at all. A few examples of calculations of properties of neutron stars are shown and prospects for the future are discussed.

  16. Finite element modeling of magnetic compression using coupled electromagnetic-structural codes

    SciTech Connect

    Hainsworth, G.; Leonard, P.J.; Rodger, D.; Leyden, C.

    1996-05-01

    A link between the electromagnetic code, MEGA, and the structural code, DYNA3D has been developed. Although the primary use of this is for modelling of Railgun components, it has recently been applied to a small experimental Coilgun at Bath. The performance of Coilguns is very dependent on projectile material conductivity, and so high purity aluminium was investigated. However, due to its low strength, it is crushed significantly by magnetic compression in the gun. Although impractical as a real projectile material, this provides useful benchmark experimental data on high strain rate plastic deformation caused by magnetic forces. This setup is equivalent to a large scale version of the classic jumping ring experiment, where the ring jumps with an acceleration of 40 kG.

  17. Comparison of the 3-D Deterministic Neutron Transport Code Attila® To Measure Data, MCNP And MCNPX For The Advanced Test Reactor

    SciTech Connect

    D. Scott Lucas; D. S. Lucas

    2005-09-01

    An LDRD (Laboratory Directed Research and Development) project is underway at the Idaho National Laboratory (INL) to apply the three-dimensional multi-group deterministic neutron transport code (Attila®) to criticality, flux and depletion calculations of the Advanced Test Reactor (ATR). This paper discusses the development of Attila models for ATR, capabilities of Attila, the generation and use of different cross-section libraries, and comparisons to ATR data, MCNP, MCNPX and future applications.

  18. Measurement of neutron dose with an organic liquid scintillator coupled with a spectrum weight function.

    PubMed

    Kim, E; Endo, A; Yamaguchi, Y; Yoshizawa, M; Nakamura, T

    2002-01-01

    A dose evaluation method for neutrons in the energy range of a few MeV to 100 MeV has been developed using a spectrum weight function (G-function), which is applied to an organic liquid scintillator of 12.7 cm in diameter and 12.7 cm in length. The G-function that converts the pulse height spectrum of the scintillator into the ambient dose equivalent, H*(10), was calculated by an unfolding method using successive approximation of the response function of the scintillator and the ambient dose equivalent per unit neutron fluence (H*(10) conversion coefficients) of ICRP 74. To verify the response function of the scintillator and the value of H*(10) evaluated by the G-function. pulse height spectra of the scintillator were measured in some different neutron fields, which have continuous energy, monoenergetic and quasi-monoenergetic spectra. Values of H*(10) estimated using the G-function and pulse height spectra of the scintillator were compared with those calculated using neutron energy spectra. These doses agreed with each other. From the results, it was concluded that H*(10) can be evaluated directly from the pulse height spectrum of the scintillator by applying the G-function proposed in this study. PMID:12212900

  19. ICECO-CEL: a coupled Eulerian-Lagrangian code for analyzing primary system response in fast reactors

    SciTech Connect

    Wang, C.Y.

    1981-02-01

    This report describes a coupled Eulerian-Lagrangian code, ICECO-CEL, for analyzing the response of the primary system during hypothetical core disruptive accidents. The implicit Eulerian method is used to calculate the fluid motion so that large fluid distortion, two-dimensional sliding interface, flow around corners, flow through coolant passageways, and out-flow boundary conditions can be treated. The explicit Lagrangian formulation is employed to compute the response of the containment vessel and other elastic-plastic solids inside the reactor containment. Large displacements, as well as geometrical and material nonlinearities are considered in the analysis. Marker particles are utilized to define the free surface or the material interface and to visualize the fluid motion. The basic equations and numerical techniques used in the Eulerian hydrodynamics and Lagrangian structural dynamics are described. Treatment of the above-core hydrodynamics, sodium spillage, fluid cavitation, free-surface boundary conditions and heat transfer are also presented. Examples are given to illustrate the capabilities of the computer code. Comparisons of the code predictions with available experimental data are also made.

  20. ICANT, a code for the self-consistent computation of ICRH antenna coupling

    SciTech Connect

    Pecoul, S.; Heuraux, S.; Koch, R.; Leclert, G.

    1996-02-01

    The code deals with 3D antenna structures (finite length antennae) that are used to launch electromagnetic waves into tokamak plasmas. The antenna radiation problem is solved using a finite boundary element technique combined with a spectral solution of the interior problem. The slab approximation is used, and periodicity in {ital y} and {ital z} directions is introduced to account for toroidal geometry. We present results for various types of antennae radiating in vacuum: antenna with a finite Faraday screen and ideal Faraday screen, antenna with side limiters and phased antenna arrays. The results (radiated power, current profile) obtained are very close to analytical solutions when available. {copyright} {ital 1996 American Institute of Physics.}

  1. Application of Gamma code coupled with turbomachinery models for high temperature gas-cooled reactors

    SciTech Connect

    Chang Oh

    2008-02-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-ofcoolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of a toxic gas, CO, and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. GAMMA code is being developed to implement turbomachinery models in the power conversion unit (PCU) and ultimately models associated with the hydrogen plant. Some preliminary results will be described in this paper.

  2. Neutron scattering from coupled phonon-impurity modes in KCl/sub 1-c/(KCN)/sub c/

    SciTech Connect

    Nicklow, R.M.; Crummett, W.P.; Mostoller, M.; Wood, R.F.

    1980-09-15

    The hybridization of host-lattice phonons with the internal-energy states of CN/sup -/ impurities in KCl has been studied by inelastic neutron scattering as a function of temperature between 10 and 100 K for samples with impurity concentrations in the range c=0.4 to 6 at. %. A temperature- and concentration-dependent coupling between phonons with E/sub g/ symmetry and the E/sub g/ transitions of CN/sup -/ ions is observed near a frequency of 0.5 THz, a value which is consistent with the energy-level spacings for CN/sup -/ in KCl as deduced by Beyeler. However, an expected coupling of phonons and CN/sup -/ transitions with T/sub 2g/ symmetry near the same frequency was not detected. A simple two-level model for the CN/sup -/ impurity provides a rather good description of the data for the E/sub g/ coupled modes for c<2 at. %, but it deviates significantly for larger concentrations. Quasielastic scattering, which has a strong dependence on impurity concentration, temperature, and phonon wave vector, is also observed.

  3. A Code for the Generation of Group Constants for Reactor Calculations from Neutron Nuclear Data in KEDAK Format.

    Energy Science and Technology Software Center (ESTSC)

    1988-09-15

    Version 00 Group averaged neutron cross sections, energy resonance self shielding factors, elastic transfer elements up to P5 approximation, the inelastic, (n,2n) and (n,3n) transfer elements, fission spectra, etc., for coarse groups (26 groups in the standard case) in the fast neutron energy range are calculated.

  4. Deciphering the Code to Aminergic G-Protein Coupled Receptor Drug Design

    PubMed Central

    Tan, Edwin S.; Groban, Eli S.; Jacobson, Matthew P.; Scanlan, Thomas S.

    2009-01-01

    Summary The trace amine-associated receptor 1 (TAAR1) is a biogenic amine G-protein coupled receptor (GPCR) that is potently activated by 3-iodothyronamine (1, T1AM) in vitro. Compound 1 is an endogenous derivative of the thyroid hormone thyroxine that rapidly induces hypothermia, anergia, and bradycardia when administered to mice. To explore the role of TAAR1 in mediating the effects of 1, we rationally designed and synthesized rat TAAR1 superagonists and lead antagonists using the rotamer toggle switch model of aminergic GPCR activation. The functional activity of a ligand was found to be correlated to the nature of its interactions with the rotamer switch residues. Allowing the rotamer switch residues to toggle to their active conformation was associated with agonism while interfering with this conformational transition resulted in antagonism. These agonist and antagonist design principles provide a conceptual model for understanding the relationship between the molecular structure of a drug and its pharmacological properties. PMID:18420141

  5. Approximate message-passing with spatially coupled structured operators, with applications to compressed sensing and sparse superposition codes

    NASA Astrophysics Data System (ADS)

    Barbier, Jean; Schülke, Christophe; Krzakala, Florent

    2015-05-01

    We study the behavior of approximate message-passing (AMP), a solver for linear sparse estimation problems such as compressed sensing, when the i.i.d matrices—for which it has been specifically designed—are replaced by structured operators, such as Fourier and Hadamard ones. We show empirically that after proper randomization, the structure of the operators does not significantly affect the performances of the solver. Furthermore, for some specially designed spatially coupled operators, this allows a computationally fast and memory efficient reconstruction in compressed sensing up to the information-theoretical limit. We also show how this approach can be applied to sparse superposition codes, allowing the AMP decoder to perform at large rates for moderate block length.

  6. Application of wavelet filtering and Barker-coded pulse compression hybrid method to air-coupled ultrasonic testing

    NASA Astrophysics Data System (ADS)

    Zhou, Zhenggan; Ma, Baoquan; Jiang, Jingtao; Yu, Guang; Liu, Kui; Zhang, Dongmei; Liu, Weiping

    2014-10-01

    Air-coupled ultrasonic testing (ACUT) technique has been viewed as a viable solution in defect detection of advanced composites used in aerospace and aviation industries. However, the giant mismatch of acoustic impedance in air-solid interface makes the transmission efficiency of ultrasound low, and leads to poor signal-to-noise (SNR) ratio of received signal. The utilisation of signal-processing techniques in non-destructive testing is highly appreciated. This paper presents a wavelet filtering and phase-coded pulse compression hybrid method to improve the SNR and output power of received signal. The wavelet transform is utilised to filter insignificant components from noisy ultrasonic signal, and pulse compression process is used to improve the power of correlated signal based on cross-correction algorithm. For the purpose of reasonable parameter selection, different families of wavelets (Daubechies, Symlet and Coiflet) and decomposition level in discrete wavelet transform are analysed, different Barker codes (5-13 bits) are also analysed to acquire higher main-to-side lobe ratio. The performance of the hybrid method was verified in a honeycomb composite sample. Experimental results demonstrated that the proposed method is very efficient in improving the SNR and signal strength. The applicability of the proposed method seems to be a very promising tool to evaluate the integrity of high ultrasound attenuation composite materials using the ACUT.

  7. Development of a GPU-based Monte Carlo dose calculation code for coupled electron-photon transport.

    PubMed

    Jia, Xun; Gu, Xuejun; Sempau, Josep; Choi, Dongju; Majumdar, Amitava; Jiang, Steve B

    2010-06-01

    Monte Carlo simulation is the most accurate method for absorbed dose calculations in radiotherapy. Its efficiency still requires improvement for routine clinical applications, especially for online adaptive radiotherapy. In this paper, we report our recent development on a GPU-based Monte Carlo dose calculation code for coupled electron-photon transport. We have implemented the dose planning method (DPM) Monte Carlo dose calculation package (Sempau et al 2000 Phys. Med. Biol. 45 2263-91) on the GPU architecture under the CUDA platform. The implementation has been tested with respect to the original sequential DPM code on the CPU in phantoms with water-lung-water or water-bone-water slab geometry. A 20 MeV mono-energetic electron point source or a 6 MV photon point source is used in our validation. The results demonstrate adequate accuracy of our GPU implementation for both electron and photon beams in the radiotherapy energy range. Speed-up factors of about 5.0-6.6 times have been observed, using an NVIDIA Tesla C1060 GPU card against a 2.27 GHz Intel Xeon CPU processor. PMID:20463376

  8. COUPLING

    DOEpatents

    Hawke, B.C.

    1963-02-26

    This patent relates to a releasable coupling connecting a control rod to a control rod drive. This remotely operable coupling mechanism can connect two elements which are laterally and angviarly misaligned, and provides a means for sensing the locked condition of the elements. The coupling utilizes a spherical bayonet joint which is locked against rotation by a ball detent lock. (AEC)

  9. Development of a technique using MCNPX code for determination of nitrogen content of explosive materials using prompt gamma neutron activation analysis method

    NASA Astrophysics Data System (ADS)

    Nasrabadi, M. N.; Bakhshi, F.; Jalali, M.; Mohammadi, A.

    2011-12-01

    Nuclear-based explosive detection methods can detect explosives by identifying their elemental components, especially nitrogen. Thermal neutron capture reactions have been used for detecting prompt gamma 10.8 MeV following radioactive neutron capture by 14N nuclei. We aimed to study the feasibility of using field-portable prompt gamma neutron activation analysis (PGNAA) along with improved nuclear equipment to detect and identify explosives, illicit substances or landmines. A 252Cf radio-isotopic source was embedded in a cylinder made of high-density polyethylene (HDPE) and the cylinder was then placed in another cylindrical container filled with water. Measurements were performed on high nitrogen content compounds such as melamine (C3H6N6). Melamine powder in a HDPE bottle was placed underneath the vessel containing water and the neutron source. Gamma rays were detected using two NaI(Tl) crystals. The results were simulated with MCNP4c code calculations. The theoretical calculations and experimental measurements were in good agreement indicating that this method can be used for detection of explosives and illicit drugs.

  10. Analysis of neutron and photon response of a TLD-ALBEDO personal dosemeter on an ISO slab phantom using TRIPOLI-4.3 Monte Carlo code.

    PubMed

    Lee, Y K

    2005-01-01

    TRIPOLI-4.3 Monte Carlo transport code has been used to evaluate the QUADOS (Quality Assurance of Computational Tools for Dosimetry) problem P4, neutron and photon response of an albedo-type thermoluminescence personal dosemeter (TLD) located on an ISO slab phantom. Two enriched 6LiF and two 7LiF TLD chips were used and they were protected, in front or behind, with a boron-loaded dosemeter-holder. Neutron response of the four chips was determined by counting 6Li(n,t)4He events using ENDF/B-VI.4 library and photon response by estimating absorbed dose (MeV g(-1)). Ten neutron energies from thermal to 20 MeV and six photon energies from 33 keV to 1.25 MeV were used to study the energy dependence. The fraction of the neutron and photon response owing to phantom backscatter has also been investigated. Detailed TRIPOLI-4.3 solutions are presented and compared with MCNP-4C calculations. PMID:16381740

  11. Integrated multi-scale simulations of drift-wave turbulence: coupling of two kinetic codes XGC1 and XGCa

    NASA Astrophysics Data System (ADS)

    Janhunen, Salomon; Hager, Robert; Ku, Seung-Hoe; Chang, Choong-Seock; Hesthaven, Jan; Choi, Jong; Zhang, Fan; Parashar, Manish

    2015-11-01

    A novel technique for acceleration of gyrokinetic total f particle simulations in diverted geometry has been developed, based on the XGC1 code and its axisymmetric version XGCa. Both XGC1 and XGCa calculate particle motion in a 5-dimensional (5D) phase space, but while XGC1 is equipped with a full turbulence solver, XGCa has an axisymmetric Poisson solver and is generally used for the simulation of neoclassical transport. Here, acceleration in transport calculations is achieved through relaxed constraints on numerical requirements in XGCa, such as mesh resolution and total number of markers. Coupled simulations have been performed for ITG turbulence, where long-term evolution is obtained by periodically calling XGC1 to obtain turbulence-driven transport while evolving the neoclassical equilibrium with XGCa. We present results from simulations with long-term evolution of the microscopic plasma state while using this technique in the presence of sources and sinks. We also introduce in-memory techniques used in the coupling between the fine-scale and coarse models, applicable for massively parallel simulations of long term evolution of kinetic plasma equilibria in the presence of turbulent and neoclassical transport processes. SciDAC Center for Edge Physics Simulation.

  12. Validation of the U.S. NRC coupled code system TRITON/TRACE/PARCS with the special power excursion reactor test III (SPERT III)

    SciTech Connect

    Wang, R. C.; Xu, Y.; Downar, T.; Hudson, N.

    2012-07-01

    The Special Power Excursion Reactor Test III (SPERT III) was a series of reactivity insertion experiments conducted in the 1950's. This paper describes the validation of the U.S. NRC Coupled Code system TRITON/PARCS/TRACE to simulate reactivity insertion accidents (RIA) by using several of the SPERT III tests. The work here used the SPERT III E-core configuration tests in which the RIA was initiated by ejecting a control rod. The resulting super-prompt reactivity excursion and negative reactivity feedback produced the familiar bell shaped power increase and decrease. The energy deposition during such a power peak has important safety consequences and provides validation basis for core coupled multi-physics codes. The transients of five separate tests are used to benchmark the PARCS/TRACE coupled code. The models were thoroughly validated using the original experiment documentation. (authors)

  13. A system for environmental model coupling and code reuse: The Great Rivers Project

    NASA Astrophysics Data System (ADS)

    Eckman, B.; Rice, J.; Treinish, L.; Barford, C.

    2008-12-01

    As part of the Great Rivers Project, IBM is collaborating with The Nature Conservancy and the Center for Sustainability and the Global Environment (SAGE) at the University of Wisconsin, Madison to build a Modeling Framework and Decision Support System (DSS) designed to help policy makers and a variety of stakeholders (farmers, fish & wildlife managers, hydropower operators, et al.) to assess, come to consensus, and act on land use decisions representing effective compromises between human use and ecosystem preservation/restoration. Initially focused on Brazil's Paraguay-Parana, China's Yangtze, and the Mississippi Basin in the US, the DSS integrates data and models from a wide variety of environmental sectors, including water balance, water quality, carbon balance, crop production, hydropower, and biodiversity. In this presentation we focus on the modeling framework aspect of this project. In our approach to these and other environmental modeling projects, we see a flexible, extensible modeling framework infrastructure for defining and running multi-step analytic simulations as critical. In this framework, we divide monolithic models into atomic components with clearly defined semantics encoded via rich metadata representation. Once models and their semantics and composition rules have been registered with the system by their authors or other experts, non-expert users may construct simulations as workflows of these atomic model components. A model composition engine enforces rules/constraints for composing model components into simulations, to avoid the creation of Frankenmodels, models that execute but produce scientifically invalid results. A common software environment and common representations of data and models are required, as well as an adapter strategy for code written in e.g., Fortran or python, that still enables efficient simulation runs, including parallelization. Since each new simulation, as a new composition of model components, requires calibration

  14. Galileogenesis: A new cosmophenomenological zip code for reheating through R-parity violating coupling

    NASA Astrophysics Data System (ADS)

    Choudhury, Sayantan; Dasgupta, Arnab

    2014-05-01

    In this paper we introduce an idea of leptogenesis scenario in higher derivative gravity induced DBI Galileon framework aka Galileogenesis in presence of one-loop R-parity violating couplings in the background of a low energy effective supergravity setup derived from higher dimensional string theory framework. We have studied extensively the detailed feature of reheating constraints and the cosmophenomenological consequences of thermal gravitino dark matter in light of PLANCK and PDG data. Finally, we have also established a direct cosmophenomenological connection among dark matter relic abundance, reheating temperature and tensor-to-scalar ratio in the context of DBI Galileon inflation. Higher order correction terms in the gravity sector are introduced in the effective action as a perturbative correction to the Einstein-Hilbert counterpart coming from the computation of Conformal Field Theory disk amplitude at the two loop level [34-36]. The matter sector encounters the effect of N=1, D=4 supergravity motivated DBI Galileon interaction which is embedded in the D3 brane. Additionally, we have considered the effect of R-parity violating interactions [37-40] in the matter sector which provide a convenient framework for quantifying quark and lepton-flavor violating effects. The low energy UV protective effective action for the proposed cosmophenomenological model is described by [31,32]: S=∫d4x √{-g}[K(Φ,X)-G(Φ,X)□Φ+B1R+(B2RRαβγδ-4B3RRαβ+B4R2)+B5] where the model dependent characteristic functions K(Φ,X) and G(Φ,X) are the implicit functions of Galileon and its kinetic counterpart is X=-1/2 >g∂μΦ∂νΦ. Additionally, Bi∀i are the self-coupling constants of graviton degrees of freedom appearing via dimensional reduction from higher dimensional string theory. Specifically B5 be the effective four dimensional cosmological constant. In general, B2≠B3≠B4 which implies that the quadratic curvature terms originated from two loop correction to the

  15. Neutronic/Thermalhydraulic Coupling Technigues for Sodium Cooled Fast Reactor Simulations

    SciTech Connect

    Jean Ragusa; Andrew Siegel; Jean-Michel Ruggieri

    2010-09-28

    The objective of this project was to test new coupling algorithms and enable efficient and scalable multi-physics simulations of advanced nuclear reactors, with considerations regarding the implementation of such algorithms in massively parallel environments. Numerical tests were carried out to verify the proposed approach and the examples included some reactor transients. The project was directly related to the Sodium Fast Reactor program element of the Generation IV Nuclear Energy Systems Initiative and the Advanced Fuel cycle Initiative, and, supported the requirement of high-fidelity simulation as a mean of achieving the goals of the presidential Global Nuclear Energy Partnership (GNEP) vision.

  16. Coupled high fidelity thermal hydraulics and neutronics for reactor safety simulations

    SciTech Connect

    Vincent A. Mousseau; Hongbin Zhang; Haihua Zhao

    2008-09-01

    This work is a continuation of previous work on the importance of accuracy in the simulation of nuclear reactor safety transients. This work is qualitative in nature and future work will be more quantitative. The focus of this work will be on a simplified single phase nuclear reactor primary. The transient of interest investigates the importance of accuracy related to passive (inherent) safety systems. The transient run here will be an Unprotected Loss of Flow (ULOF) transient. Here the coolant pump is turned off and the un’SCRAM’ed reactor transitions from forced to free convection (Natural circulation). Results will be presented that show the difference that the first order in time truncation physics makes on the transient. The purpose of this document is to illuminate a possible problem in traditional reactor simulation approaches. Detailed studies need to be done on each simulation code for each transient analyzed to determine if the first order truncation physics plays an important role.

  17. Coupling of the hydration water dynamics and the internal dynamics of actin detected by quasielastic neutron scattering

    SciTech Connect

    Fujiwara, Satoru; Plazanet, Marie; Oda, Toshiro

    2013-02-15

    Highlights: ► Quasielastic neutron scattering spectra of F-actin and G-actin were measured. ► Analysis of the samples in D{sub 2}O and H{sub 2}O provided the spectra of hydration water. ► The first layer hydration water around F-actin is less mobile than around G-actin. ► This difference in hydration water is in concert with the internal dynamics of actin. ► Water outside the first layer behaves bulk-like but influenced by the first layer. -- Abstract: In order to characterize dynamics of water molecules around F-actin and G-actin, quasielastic neutron scattering experiments were performed on powder samples of F-actin and G-actin, hydrated either with D{sub 2}O or H{sub 2}O, at hydration ratios of 0.4 and 1.0. By combined analysis of the quasielastic neutron scattering spectra, the parameter values characterizing the dynamics of the water molecules in the first hydration layer and those of the water molecules outside of the first layer were obtained. The translational diffusion coefficients (D{sub T}) of the hydration water in the first layer were found to be 1.2 × 10{sup −5} cm{sup 2}/s and 1.7 × 10{sup −5} cm{sup 2}/s for F-actin and G-actin, respectively, while that for bulk water was 2.8 × 10{sup −5} cm{sup 2}/s. The residence times were 6.6 ps and 5.0 ps for F-actin and G-actin, respectively, while that for bulk water was 0.62 ps. These differences between F-actin and G-actin, indicating that the hydration water around G-actin is more mobile than that around F-actin, are in concert with the results of the internal dynamics of F-actin and G-actin, showing that G-actin fluctuates more rapidly than F-actin. This implies that the dynamics of the hydration water is coupled to the internal dynamics of the actin molecules. The D{sub T} values of the water molecules outside of the first hydration layer were found to be similar to that of bulk water though the residence times are strongly affected by the first hydration layer. This supports the

  18. Operator-splitting errors in coupled reactive transport codes for flow and transport under atmospheric boundary conditions or layered soil profiles

    Technology Transfer Automated Retrieval System (TEKTRAN)

    One possible way of integrating subsurface flow and transport processes with (bio)geochemical reactions is to couple by means of an operator-splitting approach two completely separate codes, one for variably-saturated flow and solute transport and one for equilibrium and kinetic biogeochemical react...

  19. Innovations and enhancements in neutronic analysis of the Big-10 university research and training reactors based on the AGENT code system

    SciTech Connect

    Hursin, M.; Shanjie, X.; Burns, A.; Hopkins, J.; Satvat, N.; Gert, G.; Tsoukalas, L. H.; Jevremovic, T.

    2006-07-01

    Introduction. This paper summarizes salient aspects of the 'virtual' reactor system developed at Purdue Univ. emphasizing efficient neutronic modeling through AGENT (Arbitrary Geometry Neutron Transport) a deterministic neutron transport code. DOE's Big-10 Innovations in Nuclear Infrastructure and Education (INIE) Consortium was launched in 2002 to enhance scholarship activities pertaining to university research and training reactors (URTRs). Existing and next generation URTRs are powerful campus tools for nuclear engineering as well as a number of disciplines that include, but are not limited to, medicine, biology, material science, and food science. Advancing new computational environments for the analysis and configuration of URTRs is an important Big-10 INIE aim. Specifically, Big-10 INIE has pursued development of a 'virtual' reactor, an advanced computational environment to serve as a platform on which to build operations, utilization (research and education), and systemic analysis of URTRs physics. The 'virtual' reactor computational system will integrate computational tools addressing the URTR core and near core physics (transport, dynamics, fuel management and fuel configuration); thermal-hydraulics; beam line, in-core and near-core experiments; instrumentation and controls; confinement/containment and security issues. Such integrated computational environment does not currently exist. The 'virtual' reactor is designed to allow researchers and educators to configure and analyze their systems to optimize experiments, fuel locations for flux shaping, as well as detector selection and configuration. (authors)

  20. COUPLING

    DOEpatents

    Frisch, E.; Johnson, C.G.

    1962-05-15

    A detachable coupling arrangement is described which provides for varying the length of the handle of a tool used in relatively narrow channels. The arrangement consists of mating the key and keyhole formations in the cooperating handle sections. (AEC)

  1. Code System for NE-213 Unfolding of Neutron Spectra up to 100 MeV with Response Function Error Propagation.

    Energy Science and Technology Software Center (ESTSC)

    1987-09-30

    Version 00 The REFERDOU system can be used to calculate the response function of a NE-213 scintillation detector for energies up to 100 MeV, to interpolate and spread (Gaussian) the response function, and unfold the measured spectrum of neutrons while propagating errors from the response functions to the unfolded spectrum.

  2. Disc-Jet Coupling in the Terzan 5 Neutron Star X-ray Binary EXO 1745-248

    NASA Astrophysics Data System (ADS)

    Tetarenko, A. J.; Bahramian, A.; Sivakoff, G. R.; Tremou, E.; Linares, M.; Tudor, V.; Miller-Jones, J. C. A.; Heinke, C. O.; Chomiuk, L.; Strader, J.; Altamirano, D.; Degenaar, N.; Maccarone, T.; Patruno, A.; Sanna, A.; Wijnands, R.

    2016-04-01

    We present the results of VLA, ATCA, and Swift XRT observations of the 2015 outburst of the transient neutron star X-ray binary (NSXB), EXO 1745-248, located in the globular cluster Terzan 5. Combining (near-) simultaneous radio and X-ray measurements we measure a correlation between the radio and X-ray luminosities of L_R∝ L_X^β with β =1.68^{+0.10}_{-0.09}, linking the accretion flow (probed by X-ray luminosity) and the compact jet (probed by radio luminosity). While such a relationship has been studied in multiple black hole X-ray binaries (BHXBs), this work marks only the third NSXB with such a measurement. Constraints on this relationship in NSXBs are strongly needed, as comparing this correlation between different classes of XB systems is key in understanding the properties that affect the jet production process in accreting objects. Our best fit disc-jet coupling index for EXO 1745-248 is consistent with the measured correlation in NSXB 4U 1728-34 (β = 1.5 ± 0.2) but inconsistent with the correlation we fit using the most recent measurements from the literature of NSXB Aql X-1 (β =0.76^{+0.14}_{-0.15}). While a similar disc-jet coupling index appears to hold across multiple BHXBs in the hard accretion state, this does not appear to be the case with the three NSXBs measured so far. Additionally, the normalization of the EXO 1745-248 correlation is lower than the other two NSXBs, making it one of the most radio faint XBs ever detected in the hard state. We also report the detection of a type-I X-ray burst during this outburst, where the decay timescale is consistent with hydrogen burning.

  3. Disc-jet coupling in the Terzan 5 neutron star X-ray binary EXO 1745-248

    NASA Astrophysics Data System (ADS)

    Tetarenko, A. J.; Bahramian, A.; Sivakoff, G. R.; Tremou, E.; Linares, M.; Tudor, V.; Miller-Jones, J. C. A.; Heinke, C. O.; Chomiuk, L.; Strader, J.; Altamirano, D.; Degenaar, N.; Maccarone, T.; Patruno, A.; Sanna, A.; Wijnands, R.

    2016-07-01

    We present the results of Very Large Array, Australia Telescope Compact Array, and Swift X-ray Telescope observations of the 2015 outburst of the transient neutron star X-ray binary (NSXB), EXO 1745-248, located in the globular cluster Terzan 5. Combining (near-) simultaneous radio and X-ray measurements, we measure a correlation between the radio and X-ray luminosities of L_R∝ L_X^β with β =1.68^{+0.10}_{-0.09}, linking the accretion flow (probed by X-ray luminosity) and the compact jet (probed by radio luminosity). While such a relationship has been studied in multiple black hole X-ray binaries (BHXBs), this work marks only the third NSXB with such a measurement. Constraints on this relationship in NSXBs are strongly needed, as comparing this correlation between different classes of XB systems is key in understanding the properties that affect the jet production process in accreting objects. Our best-fitting disc-jet coupling index for EXO 1745-248 is consistent with the measured correlation in NSXB 4U 1728-34 (β = 1.5 ± 0.2) but inconsistent with the correlation we fit using the most recent measurements from the literature of NSXB Aql X-1 (β =0.76^{+0.14}_{-0.15}). While a similar disc-jet coupling index appears to hold across multiple BHXBs in the hard accretion state, this does not appear to be the case with the three NSXBs measured so far. Additionally, the normalization of the EXO 1745-248 correlation is lower than the other two NSXBs, making it one of the most radio faint XBs ever detected in the hard state. We also report the detection of a type-I X-ray burst during this outburst, where the decay time-scale is consistent with hydrogen burning.

  4. Atomic motions in poly(vinyl methyl ether): A combined study by quasielastic neutron scattering and molecular dynamics simulations in the light of the mode coupling theory.

    PubMed

    Capponi, S; Arbe, A; Alvarez, F; Colmenero, J; Frick, B; Embs, J P

    2009-11-28

    Quasielastic neutron scattering experiments (time-of-flight, neutron spin echo, and backscattering) on protonated poly(vinyl methyl ether) (PVME) have revealed the hydrogen dynamics above the glass-transition temperature. Fully atomistic molecular dynamics simulations properly validated with the neutron scattering results have allowed further characterization of the atomic motions accessing the correlation functions directly in real space. Deviations from Gaussian behavior are found in the high-momentum transfer range, which are compatible with the predictions of mode coupling theory (MCT). We have applied the MCT phenomenological version to the self-correlation functions of PVME atoms calculated from our simulation data, obtaining consistent results. The unusually large value found for the lambda-exponent parameter is close to that recently reported for polybutadiene and simple polymer models with intramolecular barriers. PMID:19947703

  5. 90° magnetic coupling in a NiFe/FeMn/biased NiFe multilayer spin valve component investigated by polarized neutron reflectometry

    SciTech Connect

    Callori, S. J. Bertinshaw, J.; Cortie, D. L.; Cai, J. W. Zhu, T.; Le Brun, A. P.; Klose, F.

    2014-07-21

    We have observed 90° magnetic coupling in a NiFe/FeMn/biased NiFe multilayer system using polarized neutron reflectometry. Magnetometry results show magnetic switching for both the biased and free NiFe layers, the latter of which reverses at low applied fields. As these measurements are only capable of providing information about the total magnetization within a sample, polarized neutron reflectometry was used to investigate the reversal behavior of the NiFe layers individually. Both the non-spin-flip and spin-flip neutron reflectometry signals were tracked around the free NiFe layer hysteresis loop and were used to detail the evolution of the magnetization during reversal. At low magnetic fields near the free NiFe coercive field, a large spin-flip signal was observed, indicating magnetization aligned perpendicular to both the applied field and pinned layer.

  6. Coupled thermohydraulic-neutronic instabilities in boiling water nuclear reactors: A review of the state of the art

    SciTech Connect

    March-Leuba, J. ); Rey, J.M. )

    1992-01-01

    This paper provides a review of the current state of the art on the topic of coupled neutronic-thermohydraulic instabilities in boiling water nuclear reactors (BWRs). The topic of BWR instabilities is of great current relevance since it affects the operation of a large number of commercial nuclear reactors. The recent trends towards introduction of high efficiency fuels that permit reactor operation at higher power densities with increased void reactivity feedback and decreased response times, has resulted in a decrease of the stability margin in the low-flow, high-power region of the operating map. This trend has resulted in a number of unexpected'' instability events. For instance, United States plants have experienced two instability events recently, one of them resulted in an automatic reactor scram; in Spain, two BWR plants have experienced unstable limit cycle oscillations that required operator action to suppress. Similar events have been experienced in other European countries. In recent years, BWR instabilities have been one of the more exciting topics of work in the area of transient thermohydraulics. As a result, significant advances in understanding the physics behind these events have occurred, and a new and improved'' state of the art has emerged recently.

  7. Coupled thermohydraulic-neutronic instabilities in boiling water nuclear reactors: A review of the state of the art

    SciTech Connect

    March-Leuba, J.; Rey, J.M.

    1992-05-01

    This paper provides a review of the current state of the art on the topic of coupled neutronic-thermohydraulic instabilities in boiling water nuclear reactors (BWRs). The topic of BWR instabilities is of great current relevance since it affects the operation of a large number of commercial nuclear reactors. The recent trends towards introduction of high efficiency fuels that permit reactor operation at higher power densities with increased void reactivity feedback and decreased response times, has resulted in a decrease of the stability margin in the low-flow, high-power region of the operating map. This trend has resulted in a number of ``unexpected`` instability events. For instance, United States plants have experienced two instability events recently, one of them resulted in an automatic reactor scram; in Spain, two BWR plants have experienced unstable limit cycle oscillations that required operator action to suppress. Similar events have been experienced in other European countries. In recent years, BWR instabilities have been one of the more exciting topics of work in the area of transient thermohydraulics. As a result, significant advances in understanding the physics behind these events have occurred, and a ``new and improved`` state of the art has emerged recently.

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

    SciTech Connect

    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.

  9. Status report on SHARP coupling framework.

    SciTech Connect

    Caceres, A.; Tautges, T. J.; Lottes, J.; Fischer, P.; Rabiti, C.; Smith, M. A.; Siegel, A.; Yang, W. S.; Palmiotti, G.

    2008-05-30

    This report presents the software engineering effort under way at ANL towards a comprehensive integrated computational framework (SHARP) for high fidelity simulations of sodium cooled fast reactors. The primary objective of this framework is to provide accurate and flexible analysis tools to nuclear reactor designers by simulating multiphysics phenomena happening in complex reactor geometries. Ideally, the coupling among different physics modules (such as neutronics, thermal-hydraulics, and structural mechanics) needs to be tight to preserve the accuracy achieved in each module. However, fast reactor cores in steady state mode represent a special case where weak coupling between neutronics and thermal-hydraulics is usually adequate. Our framework design allows for both options. Another requirement for SHARP framework has been to implement various coupling algorithms that are parallel and scalable to large scale since nuclear reactor core simulations are among the most memory and computationally intensive, requiring the use of leadership-class petascale platforms. This report details our progress toward achieving these goals. Specifically, we demonstrate coupling independently developed parallel codes in a manner that does not compromise performance or portability, while minimizing the impact on individual developers. This year, our focus has been on developing a lightweight and loosely coupled framework targeted at UNIC (our neutronics code) and Nek (our thermal hydraulics code). However, the framework design is not limited to just using these two codes.

  10. {lambda}-{sigma} coupling effect in the neutron-rich {lambda} hypernucleus {sub {lambda}}{sup 10}Li in a microscopic shell-model calculation

    SciTech Connect

    Umeya, A.; Harada, T.

    2009-02-15

    We investigate the structure of the neutron-rich {lambda}-hypernucleus {sub {lambda}}{sup 10}Li by using microscopic shell-model calculations considering a {lambda}-{sigma} coupling effect. The calculated {sigma}-mixing probability in the {sub {lambda}}{sup 10}Li ground state is found to be about 0.34% that is coherently enhanced by the {lambda}-{sigma} coupling configurations, leading to the energy shift 0.28 MeV that is about 3 times larger than that in {sub {lambda}}{sup 7}Li. The importance of the {sigma} configuration obtained by the {sigma}N interaction and the potentiality of the neutron-rich environment are discussed.

  11. Neutron metrology laboratory facility simulation.

    PubMed

    Pereira, Mariana; Salgado, Ana P; Filho, Aidano S; Pereira, Walsan W; Patrão, Karla C S; Fonseca, Evaldo S

    2014-10-01

    The Neutron Low Scattering Laboratory in Brazil has been completely rebuilt. Evaluation of air attenuation parameters and neutron component scattering in the room was done using Monte Carlo simulation code. Neutron fields produced by referenced neutron source were used to calculate neutron scattering and air attenuation. PMID:24864318

  12. Coupled 2-dimensional cascade theory for noise an d unsteady aerodynamics of blade row interaction in turbofans. Volume 2: Documentation for computer code CUP2D

    NASA Technical Reports Server (NTRS)

    Hanson, Donald B.

    1994-01-01

    A two dimensional linear aeroacoustic theory for rotor/stator interaction with unsteady coupling was derived and explored in Volume 1 of this report. Computer program CUP2D has been written in FORTRAN embodying the theoretical equations. This volume (Volume 2) describes the structure of the code, installation and running, preparation of the input file, and interpretation of the output. A sample case is provided with printouts of the input and output. The source code is included with comments linking it closely to the theoretical equations in Volume 1.

  13. ITER Generic Diagnostic Upper Port Plug Nuclear Heating and Personnel Dose Rate Assesment Neutronics Analysis using the ATTILA Discrete Ordinates Code

    SciTech Connect

    Russell Feder and Mahmoud Z. Yousef

    2009-05-29

    Neutronics analysis to find nuclear heating rates and personnel dose rates were conducted in support of the integration of diagnostics in to the ITER Upper Port Plugs. Simplified shielding models of the Visible-Infrared diagnostic and of the ECH heating system were incorporated in to the ITER global CAD model. Results for these systems are representative of typical designs with maximum shielding and a small aperture (Vis-IR) and minimal shielding with a large aperture (ECH). The neutronics discrete-ordinates code ATTILA® and SEVERIAN® (the ATTILA parallel processing version) was used. Material properties and the 500 MW D-T volume source were taken from the ITER “Brand Model” MCNP benchmark model. A biased quadrature set equivelant to Sn=32 and a scattering degree of Pn=3 were used along with a 46-neutron and 21-gamma FENDL energy subgrouping. Total nuclear heating (neutron plug gamma heating) in the upper port plugs ranged between 380 and 350 kW for the Vis-IR and ECH cases. The ECH or Large Aperture model exhibited lower total heating but much higher peak volumetric heating on the upper port plug structure. Personnel dose rates are calculated in a three step process involving a neutron-only transport calculation, the generation of activation volume sources at pre-defined time steps and finally gamma transport analyses are run for selected time steps. ANSI-ANS 6.1.1 1977 Flux-to-Dose conversion factors were used. Dose rates were evaluated for 1 full year of 500 MW DT operation which is comprised of 3000 1800-second pulses. After one year the machine is shut down for maintenance and personnel are permitted to access the diagnostic interspace after 2-weeks if dose rates are below 100 μSv/hr. Dose rates in the Visible-IR diagnostic model after one day of shutdown were 130 μSv/hr but fell below the limit to 90 μSv/hr 2-weeks later. The Large Aperture or ECH style shielding model exhibited higher and more persistent dose rates. After 1-day the dose rate was 230

  14. Calculation of pre-equilibrium effects in neutron-induced cross section on 32,34S isotopes using the EMPIRE 3.2 code

    NASA Astrophysics Data System (ADS)

    Yettou, Leila; Belgaid, Mohamed

    2015-07-01

    In this study, a new version EMPIRE 3.2 code was used in the cross section calculations of (n,p) reactions and in the calculation of proton emission spectra produced by (n,xp) reactions. Exciton model predictions combined with the Kalbach angular distribution systematics were used and some parameters such as those of mean free path, cluster emission in terms of Iwamoto-Harada model, optical model potentials of Morillon for neutrons and protons in the energy range up to 20 MeV, level density for spherical nuclei of Gilbert-Cameron model and width fluctuation correction in terms of compound nucleus have been investigated our calculations. The excitation functions and the proton emission spectra for 32,34S nuclei were calculated, discussed and found in good agreement with available experimental data.

  15. Code System to Solve the Few-Group Neutron Diffusion Equation Utilizing the Nodal Expansion Method (NEM) for Eigenvalue, Adjoint, and Fixed-Source

    Energy Science and Technology Software Center (ESTSC)

    2004-04-21

    Version 04 NESTLE solves the few-group neutron diffusion equation utilizing the NEM. The NESTLE code can solve the eigenvalue (criticality), eigenvalue adjoint, external fixed-source steady-state, and external fixed-source or eigenvalue initiated transient problems. The eigenvalue problem allows criticality searches to be completed, and the external fixed-source steady-state problem can search to achieve a specified power level. Transient problems model delayed neutrons via precursor groups. Several core properties can be input as time dependent. Two- ormore » four-energy groups can be utilized, with all energy groups being thermal groups (i.e., upscatter exits) if desired. Core geometries modeled include Cartesian and hexagonal. Three-, two-, and one-dimensional models can be utilized with various symmetries. The thermal conditions predicted by the thermal-hydraulic model of the core are used to correct cross sections for temperature and density effects. Cross sections are parameterized by color, control rod state (i.e., in or out), and burnup, allowing fuel depletion to be modeled. Either a macroscopic or microscopic model may be employed.« less

  16. A Monte Carlo neutron transport code for eigenvalue calculations on a dual-GPU system and CUDA environment

    SciTech Connect

    Liu, T.; Ding, A.; Ji, W.; Xu, X. G.; Carothers, C. D.; Brown, F. B.

    2012-07-01

    Monte Carlo (MC) method is able to accurately calculate eigenvalues in reactor analysis. Its lengthy computation time can be reduced by general-purpose computing on Graphics Processing Units (GPU), one of the latest parallel computing techniques under development. The method of porting a regular transport code to GPU is usually very straightforward due to the 'embarrassingly parallel' nature of MC code. However, the situation becomes different for eigenvalue calculation in that it will be performed on a generation-by-generation basis and the thread coordination should be explicitly taken care of. This paper presents our effort to develop such a GPU-based MC code in Compute Unified Device Architecture (CUDA) environment. The code is able to perform eigenvalue calculation under simple geometries on a multi-GPU system. The specifics of algorithm design, including thread organization and memory management were described in detail. The original CPU version of the code was tested on an Intel Xeon X5660 2.8 GHz CPU, and the adapted GPU version was tested on NVIDIA Tesla M2090 GPUs. Double-precision floating point format was used throughout the calculation. The result showed that a speedup of 7.0 and 33.3 were obtained for a bare spherical core and a binary slab system respectively. The speedup factor was further increased by a factor of {approx}2 on a dual GPU system. The upper limit of device-level parallelism was analyzed, and a possible method to enhance the thread-level parallelism was proposed. (authors)

  17. Rapidly rotating neutron star progenitors

    NASA Astrophysics Data System (ADS)

    Postnov, K. A.; Kuranov, A. G.; Kolesnikov, D. A.; Popov, S. B.; Porayko, N. K.

    2016-08-01

    Rotating proto-neutron stars can be important sources of gravitational waves to be searched for by present-day and future interferometric detectors. It was demonstrated by Imshennik that in extreme cases the rapid rotation of a collapsing stellar core may lead to fission and formation of a binary proto-neutron star which subsequently merges due to gravitational wave emission. In the present paper, we show that such dynamically unstable collapsing stellar cores may be the product of a former merger process of two stellar cores in a common envelope. We applied population synthesis calculations to assess the expected fraction of such rapidly rotating stellar cores which may lead to fission and formation of a pair of proto-neutron stars. We have used the BSE population synthesis code supplemented with a new treatment of stellar core rotation during the evolution via effective core-envelope coupling, characterized by the coupling time, τc. The validity of this approach is checked by direct MESA calculations of the evolution of a rotating 15 M⊙ star. From comparison of the calculated spin distribution of young neutron stars with the observed one, reported by Popov and Turolla, we infer the value τc ≃ 5 × 105 years. We show that merging of stellar cores in common envelopes can lead to collapses with dynamically unstable proto-neutron stars, with their formation rate being ˜0.1 - 1% of the total core collapses, depending on the common envelope efficiency.

  18. Spectrometry and dosimetry of fast neutrons using pin diode detectors

    NASA Astrophysics Data System (ADS)

    Zaki Dizaji, H.; Kakavand, T.; Abbasi Davani, F.

    2014-03-01

    Elastic scattering of light nuclei, especially hydrogen, is widely used for detection of fast neutrons. Semiconductor devices based on silicon detectors are frequently used for different radiation detections. In this work, a neutron spectrometer consisting of a pin diode coupled with a polyethylene converter and aluminum degrader layers has been developed. Aluminum layers are used as discriminators of different neutron energies for detectors. The response of the converter-degrader-pin diode configuration, the optimum thickness of the converter and the degrader layers have been extracted using MCNP and SRIM simulation codes. The possibility of using this type of detector for fast neutron spectrometry and dosimetry has been investigated. A fairly good agreement was seen between neutron energy spectrum and dose obtained from our configurations and these specifications from an 241Am-Be neutron source.

  19. ITS version 5.0 :the integrated TIGER series of coupled electron/Photon monte carlo transport codes with CAD geometry.

    SciTech Connect

    Franke, Brian Claude; Kensek, Ronald Patrick; Laub, Thomas William

    2005-09-01

    ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of linear time-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 5.0, the latest version of ITS, contains (1) improvements to the ITS 3.0 continuous-energy codes, (2) multigroup codes with adjoint transport capabilities, (3) parallel implementations of all ITS codes, (4) a general purpose geometry engine for linking with CAD or other geometry formats, and (5) the Cholla facet geometry library. Moreover, the general user friendliness of the software has been enhanced through increased internal error checking and improved code portability.

  20. PRELIMINARY COUPLING OF THE MONTE CARLO CODE OPENMC AND THE MULTIPHYSICS OBJECT-ORIENTED SIMULATION ENVIRONMENT (MOOSE) FOR ANALYZING DOPPLER FEEDBACK IN MONTE CARLO SIMULATIONS

    SciTech Connect

    Matthew Ellis; Derek Gaston; Benoit Forget; Kord Smith

    2011-07-01

    In recent years the use of Monte Carlo methods for modeling reactors has become feasible due to the increasing availability of massively parallel computer systems. One of the primary challenges yet to be fully resolved, however, is the efficient and accurate inclusion of multiphysics feedback in Monte Carlo simulations. The research in this paper presents a preliminary coupling of the open source Monte Carlo code OpenMC with the open source Multiphysics Object-Oriented Simulation Environment (MOOSE). The coupling of OpenMC and MOOSE will be used to investigate efficient and accurate numerical methods needed to include multiphysics feedback in Monte Carlo codes. An investigation into the sensitivity of Doppler feedback to fuel temperature approximations using a two dimensional 17x17 PWR fuel assembly is presented in this paper. The results show a functioning multiphysics coupling between OpenMC and MOOSE. The coupling utilizes Functional Expansion Tallies to accurately and efficiently transfer pin power distributions tallied in OpenMC to unstructured finite element meshes used in MOOSE. The two dimensional PWR fuel assembly case also demonstrates that for a simplified model the pin-by-pin doppler feedback can be adequately replicated by scaling a representative pin based on pin relative powers.

  1. Predicting fissile content of spent nuclear fuel assemblies with the passive neutron Albedo reactivity technique and Monte Carlo code emulation

    SciTech Connect

    Conlin, Jeremy Lloyd; Tobin, Stephen J

    2010-10-13

    There is a great need in the safeguards community to be able to nondestructively quantify the mass of plutonium of a spent nuclear fuel assembly. As part of the Next Generation of Safeguards Initiative, we are investigating several techniques, or detector systems, which, when integrated, will be capable of quantifying the plutonium mass of a spent fuel assembly without dismantling the assembly. This paper reports on the simulation of one of these techniques, the Passive Neutron Albedo Reactivity with Fission Chambers (PNAR-FC) system. The response of this system over a wide range of spent fuel assemblies with different burnup, initial enrichment, and cooling time characteristics is shown. A Monte Carlo method of using these modeled results to estimate the fissile content of a spent fuel assembly has been developed. A few numerical simulations of using this method are shown. Finally, additional developments still needed and being worked on are discussed.

  2. CHIC - Coupling Habitability, Interior and Crust: A new Code for Modeling the Thermal Evolution of Planets and Moons

    NASA Astrophysics Data System (ADS)

    Noack, Lena; Rivoldini, Attilio; Van Hoolst, Tim

    2015-04-01

    We present a new numerical code (CHIC) for the simulation of the thermal evolution of terrestrial planets. The code consists of both a 1d parameterised model to evaluate the temperature profile in the planet's interior and a 2d/3d convection model for the silicate mantle - the latter uses either a Cartesian box, a 2d cylindrical sphere or a 2d spherical annulus. The code is modular and can be easily extended (for example to include an atmosphere module). In the convection model next to the energy equation the conservation equations of mass and momentum are solved, as well. We apply either a Boussinesq approximation or an extended Boussinesq approximation for mantle convection; compressible treatment is planned for the future. The code provides information on the temperature field in the mantle, convective velocities and convective stresses. Simulations can be run under steady-state or thermal evolution conditions. The CHIC code handles surface volcanism, crustal development, and different regimes of surface mobilization like plate tectonics. It is therefore well suited for studying scenarios related to the habitability of terrestrial planets. The code provides a user updatable library of thermodynamic properties of iron and common mantle silicates as well as associated equations of state that allow to compute material properties at high pressure and temperature. Furthermore, the interior structure of a planet for given composition and mass can be determined, yielding the core and planet radius that can then be automatically used for the thermal evolution simulation. CHIC does also accommodate a module for computing a simple parameterised thermal evolution model of a planet's core that includes the formation of an inner core. This module can be combined with either the 1d parameterised thermal evolution model or the 2d/3d mantle convection model. The code has been benchmarked with different convection codes, and compared to published interior-structure models and 1d

  3. Intercomparison of inductively coupled plasma mass spectrometry, quantitative neutron capture radiography, and prompt gamma activation analysis for the determination of boron in biological samples.

    PubMed

    Schütz, C L; Brochhausen, C; Hampel, G; Iffland, D; Kuczewski, B; Otto, G; Schmitz, T; Stieghorst, C; Kratz, J V

    2012-10-01

    Boron determination in blood and tissue samples is a crucial task especially for treatment planning, preclinical research, and clinical application of boron neutron capture therapy (BNCT). Comparison of clinical findings remains difficult due to a variety of analytical methods, protocols, and standard reference materials in use. This paper addresses the comparability of inductively coupled plasma mass spectrometry, quantitative neutron capture radiography, and prompt gamma activation analysis for the determination of boron in biological samples. It was possible to demonstrate that three different methods relying on three different principles of sample preparation and boron detection can be validated against each other and yield consistent results for both blood and tissue samples. The samples were obtained during a clinical study for the application of BNCT for liver malignancies and therefore represent a realistic situation for boron analysis. PMID:22918535

  4. Black Hole - Neutron Star Binary Simulations at Georgia Tech

    NASA Astrophysics Data System (ADS)

    Haas, Roland

    2009-05-01

    Mixed compact object binaries consisting of a black hole and a neutron star are expected to be not only one of the primary sources of gravitational radiation to be observed by interferometric detectors but also the central engine of short gamma-ray bursts. We report on the status of our effort at Georgia Tech to model these mixed binary systems using the moving puncture method. The results are obtained with an enhanced version our vacuum MayaKranc code coupled to the hydrodynamics Whisky code. We present preliminary results of gravitational waveforms and the disruption of the neutron star for simple polytropic equations of state.

  5. Coupling Legacy and Contemporary Deterministic Codes to Goldsim for Probabilistic Assessments of Potential Low-Level Waste Repository Sites

    NASA Astrophysics Data System (ADS)

    Mattie, P. D.; Knowlton, R. G.; Arnold, B. W.; Tien, N.; Kuo, M.

    2006-12-01

    Sandia National Laboratories (Sandia), a U.S. Department of Energy National Laboratory, has over 30 years experience in radioactive waste disposal and is providing assistance internationally in a number of areas relevant to the safety assessment of radioactive waste disposal systems. International technology transfer efforts are often hampered by small budgets, time schedule constraints, and a lack of experienced personnel in countries with small radioactive waste disposal programs. In an effort to surmount these difficulties, Sandia has developed a system that utilizes a combination of commercially available codes and existing legacy codes for probabilistic safety assessment modeling that facilitates the technology transfer and maximizes limited available funding. Numerous codes developed and endorsed by the United States Nuclear Regulatory Commission and codes developed and maintained by United States Department of Energy are generally available to foreign countries after addressing import/export control and copyright requirements. From a programmatic view, it is easier to utilize existing codes than to develop new codes. From an economic perspective, it is not possible for most countries with small radioactive waste disposal programs to maintain complex software, which meets the rigors of both domestic regulatory requirements and international peer review. Therefore, re-vitalization of deterministic legacy codes, as well as an adaptation of contemporary deterministic codes, provides a creditable and solid computational platform for constructing probabilistic safety assessment models. External model linkage capabilities in Goldsim and the techniques applied to facilitate this process will be presented using example applications, including Breach, Leach, and Transport-Multiple Species (BLT-MS), a U.S. NRC sponsored code simulating release and transport of contaminants from a subsurface low-level waste disposal facility used in a cooperative technology transfer

  6. Simulation of the full-core pin-model by JMCT Monte Carlo neutron-photon transport code

    SciTech Connect

    Li, D.; Li, G.; Zhang, B.; Shu, L.; Shangguan, D.; Ma, Y.; Hu, Z.

    2013-07-01

    Since the large numbers of cells over a million, the tallies over a hundred million and the particle histories over ten billion, the simulation of the full-core pin-by-pin model has become a real challenge for the computers and the computational methods. On the other hand, the basic memory of the model has exceeded the limit of a single CPU, so the spatial domain and data decomposition must be considered. JMCT (J Monte Carlo Transport code) has successful fulfilled the simulation of the full-core pin-by-pin model by the domain decomposition and the nested parallel computation. The k{sub eff} and flux of each cell are obtained. (authors)

  7. Fullwave coupling to a 3D antenna code using Green's function formulation of wave-particle response

    NASA Astrophysics Data System (ADS)

    Wright, John; Bonoli, P. T.; Bilato, R.; Brambilla, M.; Maggiora, R.; Lancellotti, V.

    2006-10-01

    Using the fullwave code, TORIC, and the 3D antenna code, TOPICA, we construct a complete linear system for the RF driven plasma. The 3D finite element antenna code, TOPICA, requires an admittance, Y, for the plasma, where B=YE. In this work, TORIC was modified to allow excitation of the (Eη, Eζ) electric field components at the plasma surface, corresponding to a single poloidal and toroidal mode number combination (m,n). This leads to the tensor response: Yn= ( ll Yηη& YηζYζη& Yζζ), where each of the Yn submatrices is Nm in size. It is shown that the admittance matrix is equivalent to a Green's function calculation for the fullwave system and the net work done is less than twice a single fullwave calculation. The admittance calculation is used with loading calculation from TOPICA to construct self consistent plasma and antenna currents.

  8. Fullwave coupling to a 3D antenna code using Green's function formulation of wave-particle response.

    NASA Astrophysics Data System (ADS)

    Wright, John; Bonoli, Paul; Brambilla, Marco; Lancelloti, Vito; Maggiora, Riccardo; Carter, Mark

    2006-04-01

    Using the fullwave code, TORIC ,and the 3D antenna code, TOPICA, we construct a complete linear system for the RF driven plasma. The 3D finite element antenna code, TOPICA, requires an admittance, Y, for the plasma, where B=YE. In this work TORIC was modified to allow excitation of the (Eη, Eζ) electric field components at the plasma surface, corresponding to a single poloidal and toroidal mode number combination (m,n). This leads the tensor response: Y=( *20c Yηη & Yηζ Yζη & Yζζ ), where each of the Yn submatrices is Nm in size. It is shown that the admittance matrix is equivalent to a Greens function calculation for the fullwave system and in addition, the net work done in the calculation is on the order of twice a single fullwave calculation. After the admittance calculation is done, the response of a plasma to an antenna driven at a given frequency can be calculated by only running the TOPICA code for a new antenna geometry. In tests of loading, TOPICA has been able reproduce loading of the Alcator D antenna (S12 coefficient accurately.).

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

    SciTech Connect

    Demaziere, C.; Larsson, V.

    2012-07-01

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

  10. Using a multiphase flow code to model the coupled effects of repository consolidation and multiphase brine and gas flow at the Waste Isolation Pilot Plant

    SciTech Connect

    Freeze, G.A.; Larson, K.W.; Davies, P.B.; Webb, S.W.

    1995-10-01

    Long-term repository assessment must consider the processes of (1) gas generation, (2) room closure and expansions due to salt creep, and (3) multiphase (brine and gas) fluid flow, as well as the complex coupling between these three processes. The mechanical creep closure code SANCHO was used to simulate the closure of a single, perfectly sealed disposal room filled with water and backfill. SANCHO uses constitutive models to describe salt creep, waste consolidation, and backfill consolidation, Five different gas-generation rate histories were simulated, differentiated by a rate multiplier, f, which ranged from 0.0 (no gas generation) to 1.0 (expected gas generation under brine-dominated conditions). The results of the SANCHO f-series simulations provide a relationship between gas generation, room closure, and room pressure for a perfectly sealed room. Several methods for coupling this relationship with multiphase fluid flow into and out of a room were examined. Two of the methods are described.

  11. An instability due to the nonlinear coupling of p-modes to g-modes: Implications for coalescing neutron star binaries

    SciTech Connect

    Weinberg, Nevin N.; Arras, Phil; Burkart, Joshua

    2013-06-01

    A weakly nonlinear fluid wave propagating within a star can be unstable to three-wave interactions. The resonant parametric instability is a well-known form of three-wave interaction in which a primary wave of frequency ω {sub a} excites a pair of secondary waves of frequency ω {sub b} + ω {sub c} ≅ ω {sub a}. Here we consider a nonresonant form of three-wave interaction in which a low-frequency primary wave excites a high-frequency p-mode and a low-frequency g-mode such that ω {sub b} + ω {sub c} >> ω {sub a}. We show that a p-mode can couple so strongly to a g-mode of similar radial wavelength that this type of nonresonant interaction is unstable even if the primary wave amplitude is small. As an application, we analyze the stability of the tide in coalescing neutron star binaries to p-g mode coupling. We find that the equilibrium tide and dynamical tide are both p-g unstable at gravitational wave frequencies f {sub gw} ≳ 20 Hz and drive short wavelength p-g mode pairs to significant energies on very short timescales (much less than the orbital decay time due to gravitational radiation). Resonant parametric coupling to the tide is, by contrast, either stable or drives modes at a much smaller rate. We do not solve for the saturation of the p-g instability and therefore we cannot say precisely how it influences the evolution of neutron star binaries. However, we show that if even a single daughter mode saturates near its wave breaking amplitude, the p-g instability of the equilibrium tide will (1) induce significant orbital phase errors (Δφ ≳ 1 radian) that accumulate primarily at low frequencies (f {sub gw} ≲ 50 Hz) and (2) heat the neutron star core to a temperature of T ∼ 10{sup 10} K. Since there are at least ∼100 unstable p-g daughter pairs, Δφ and T are potentially much larger than these values. Tides might therefore significantly influence the gravitational wave signal and electromagnetic emission from coalescing neutron star binaries

  12. An Instability due to the Nonlinear Coupling of p-modes to g-modes: Implications for Coalescing Neutron Star Binaries

    NASA Astrophysics Data System (ADS)

    Weinberg, Nevin N.; Arras, Phil; Burkart, Joshua

    2013-06-01

    A weakly nonlinear fluid wave propagating within a star can be unstable to three-wave interactions. The resonant parametric instability is a well-known form of three-wave interaction in which a primary wave of frequency ω a excites a pair of secondary waves of frequency ω b + ω c ~= ω a . Here we consider a nonresonant form of three-wave interaction in which a low-frequency primary wave excites a high-frequency p-mode and a low-frequency g-mode such that ω b + ω c Gt ω a . We show that a p-mode can couple so strongly to a g-mode of similar radial wavelength that this type of nonresonant interaction is unstable even if the primary wave amplitude is small. As an application, we analyze the stability of the tide in coalescing neutron star binaries to p-g mode coupling. We find that the equilibrium tide and dynamical tide are both p-g unstable at gravitational wave frequencies f gw >~ 20 Hz and drive short wavelength p-g mode pairs to significant energies on very short timescales (much less than the orbital decay time due to gravitational radiation). Resonant parametric coupling to the tide is, by contrast, either stable or drives modes at a much smaller rate. We do not solve for the saturation of the p-g instability and therefore we cannot say precisely how it influences the evolution of neutron star binaries. However, we show that if even a single daughter mode saturates near its wave breaking amplitude, the p-g instability of the equilibrium tide will (1) induce significant orbital phase errors (Δphi >~ 1 radian) that accumulate primarily at low frequencies (f gw <~ 50 Hz) and (2) heat the neutron star core to a temperature of T ~ 1010 K. Since there are at least ~100 unstable p-g daughter pairs, Δphi and T are potentially much larger than these values. Tides might therefore significantly influence the gravitational wave signal and electromagnetic emission from coalescing neutron star binaries at much larger orbital separations than previously thought.

  13. An integrated radiation physics computer code system.

    NASA Technical Reports Server (NTRS)

    Steyn, J. J.; Harris, D. W.

    1972-01-01

    An integrated computer code system for the semi-automatic and rapid analysis of experimental and analytic problems in gamma photon and fast neutron radiation physics is presented. Such problems as the design of optimum radiation shields and radioisotope power source configurations may be studied. The system codes allow for the unfolding of complex neutron and gamma photon experimental spectra. Monte Carlo and analytic techniques are used for the theoretical prediction of radiation transport. The system includes a multichannel pulse-height analyzer scintillation and semiconductor spectrometer coupled to an on-line digital computer with appropriate peripheral equipment. The system is geometry generalized as well as self-contained with respect to material nuclear cross sections and the determination of the spectrometer response functions. Input data may be either analytic or experimental.

  14. Advanced neutron absorber materials

    DOEpatents

    Branagan, Daniel J.; Smolik, Galen R.

    2000-01-01

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

  15. Simulation of decay heat removal by natural convection in a pool type fast reactor model-ramona-with coupled 1D/2D thermal hydraulic code system

    SciTech Connect

    Kasinathan, N.; Rajakumar, A.; Vaidyanathan, G.; Chetal, S.C.

    1995-09-01

    Post shutdown decay heat removal is an important safety requirement in any nuclear system. In order to improve the reliability of this function, Liquid metal (sodium) cooled fast breeder reactors (LMFBR) are equipped with redundant hot pool dipped immersion coolers connected to natural draught air cooled heat exchangers through intermediate sodium circuits. During decay heat removal, flow through the core, immersion cooler primary side and in the intermediate sodium circuits are also through natural convection. In order to establish the viability and validate computer codes used in making predictions, a 1:20 scale experimental model called RAMONA with water as coolant has been built and experimental simulation of decay heat removal situation has been performed at KfK Karlsruhe. Results of two such experiments have been compiled and published as benchmarks. This paper brings out the results of the numerical simulation of one of the benchmark case through a 1D/2D coupled code system, DHDYN-1D/THYC-2D and the salient features of the comparisons. Brief description of the formulations of the codes are also included.

  16. A user's manual for MASH 1. 0: A Monte Carlo Adjoint Shielding Code System

    SciTech Connect

    Johnson, J.O.

    1992-03-01

    The Monte Carlo Adjoint Shielding Code System, MASH, calculates neutron and gamma-ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air-over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system include the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. MASH is the successor to the Vehicle Code System (VCS) initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the dose importance'' of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response a a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user's manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem (input data and selected output edits) for each code.

  17. Code System for the Analysis of Material Test Reactor (MTR) Cores.

    Energy Science and Technology Software Center (ESTSC)

    1995-03-24

    Version 00 The RETRAC code uses a set of coupled neutron point-kinetics equations and thermal-hydraulic conservation laws to simulate nuclear reactor core behavior under transient or accident conditions. The reactor core is represented by a single equivalent unit cell composed of three regions: fuel, clad, and moderator (coolant).

  18. The Hypothesis that the Genetic Code Originated in Coupled Synthesis of Proteins and the Evolutionary Predecessors of Nucleic Acids in Primitive Cells

    PubMed Central

    Francis, Brian R.

    2015-01-01

    Although analysis of the genetic code has allowed explanations for its evolution to be proposed, little evidence exists in biochemistry and molecular biology to offer an explanation for the origin of the genetic code. In particular, two features of biology make the origin of the genetic code difficult to understand. First, nucleic acids are highly complicated polymers requiring numerous enzymes for biosynthesis. Secondly, proteins have a simple backbone with a set of 20 different amino acid side chains synthesized by a highly complicated ribosomal process in which mRNA sequences are read in triplets. Apparently, both nucleic acid and protein syntheses have extensive evolutionary histories. Supporting these processes is a complex metabolism and at the hub of metabolism are the carboxylic acid cycles. This paper advances the hypothesis that the earliest predecessor of the nucleic acids was a β-linked polyester made from malic acid, a highly conserved metabolite in the carboxylic acid cycles. In the β-linked polyester, the side chains are carboxylic acid groups capable of forming interstrand double hydrogen bonds. Evolution of the nucleic acids involved changes to the backbone and side chain of poly(β-d-malic acid). Conversion of the side chain carboxylic acid into a carboxamide or a longer side chain bearing a carboxamide group, allowed information polymers to form amide pairs between polyester chains. Aminoacylation of the hydroxyl groups of malic acid and its derivatives with simple amino acids such as glycine and alanine allowed coupling of polyester synthesis and protein synthesis. Use of polypeptides containing glycine and l-alanine for activation of two different monomers with either glycine or l-alanine allowed simple coded autocatalytic synthesis of polyesters and polypeptides and established the first genetic code. A primitive cell capable of supporting electron transport, thioester synthesis, reduction reactions, and synthesis of polyesters and

  19. The Hypothesis that the Genetic Code Originated in Coupled Synthesis of Proteins and the Evolutionary Predecessors of Nucleic Acids in Primitive Cells.

    PubMed

    Francis, Brian R

    2015-01-01

    Although analysis of the genetic code has allowed explanations for its evolution to be proposed, little evidence exists in biochemistry and molecular biology to offer an explanation for the origin of the genetic code. In particular, two features of biology make the origin of the genetic code difficult to understand. First, nucleic acids are highly complicated polymers requiring numerous enzymes for biosynthesis. Secondly, proteins have a simple backbone with a set of 20 different amino acid side chains synthesized by a highly complicated ribosomal process in which mRNA sequences are read in triplets. Apparently, both nucleic acid and protein syntheses have extensive evolutionary histories. Supporting these processes is a complex metabolism and at the hub of metabolism are the carboxylic acid cycles. This paper advances the hypothesis that the earliest predecessor of the nucleic acids was a β-linked polyester made from malic acid, a highly conserved metabolite in the carboxylic acid cycles. In the β-linked polyester, the side chains are carboxylic acid groups capable of forming interstrand double hydrogen bonds. Evolution of the nucleic acids involved changes to the backbone and side chain of poly(β-d-malic acid). Conversion of the side chain carboxylic acid into a carboxamide or a longer side chain bearing a carboxamide group, allowed information polymers to form amide pairs between polyester chains. Aminoacylation of the hydroxyl groups of malic acid and its derivatives with simple amino acids such as glycine and alanine allowed coupling of polyester synthesis and protein synthesis. Use of polypeptides containing glycine and l-alanine for activation of two different monomers with either glycine or l-alanine allowed simple coded autocatalytic synthesis of polyesters and polypeptides and established the first genetic code. A primitive cell capable of supporting electron transport, thioester synthesis, reduction reactions, and synthesis of polyesters and

  20. Gamma scintillator system using boron carbide for neutron detection

    NASA Astrophysics Data System (ADS)

    Ben-Galim, Y.; Wengrowicz, U.; Raveh, A.; Orion, I.

    2014-08-01

    A new approach for neutron detection enhancement to scintillator gamma-ray detectors is suggested. By using a scintillator coupled with a boron carbide (B4C) disc, the 478 keV gamma-photon emitted from the excited Li in 94% of the 10B(n,α)7Li interactions was detected. This suggests that the performance of existing gamma detection systems in Homeland security applications can be improved. In this study, a B4C disc (2 in. diameter, 0.125 in. thick) with ~19.8% 10B was used and coupled with a scintillator gamma-ray detector. In addition, the neutron thermalization moderator was studied in order to be able to increase the neutron sensitivity. An improvement in the detector which is easy to assemble, affordable and efficient was demonstrated. Furthermore, a tailored Monte-Carlo code written in MATLAB was developed for validation of the proposed application through efficiency estimation for thermal neutrons. Validation of the code was accomplished by showing that the MATLAB code results were well correlated to a Monte-Carlo MCNP code results. The measured efficiency of the assembled experimental model was observed to be in agreement with both models calculations.

  1. A fast parallel code for calculating energies and oscillator strengths of many-electron atoms at neutron star magnetic field strengths in adiabatic approximation

    NASA Astrophysics Data System (ADS)

    Engel, D.; Klews, M.; Wunner, G.

    2009-02-01

    We have developed a new method for the fast computation of wavelengths and oscillator strengths for medium-Z atoms and ions, up to iron, at neutron star magnetic field strengths. The method is a parallelized Hartree-Fock approach in adiabatic approximation based on finite-element and B-spline techniques. It turns out that typically 15-20 finite elements are sufficient to calculate energies to within a relative accuracy of 10-5 in 4 or 5 iteration steps using B-splines of 6th order, with parallelization speed-ups of 20 on a 26-processor machine. Results have been obtained for the energies of the ground states and excited levels and for the transition strengths of astrophysically relevant atoms and ions in the range Z=2…26 in different ionization stages. Catalogue identifier: AECC_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AECC_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 3845 No. of bytes in distributed program, including test data, etc.: 27 989 Distribution format: tar.gz Programming language: MPI/Fortran 95 and Python Computer: Cluster of 1-26 HP Compaq dc5750 Operating system: Fedora 7 Has the code been vectorised or parallelized?: Yes RAM: 1 GByte Classification: 2.1 External routines: MPI/GFortran, LAPACK, PyLab/Matplotlib Nature of problem: Calculations of synthetic spectra [1] of strongly magnetized neutron stars are bedevilled by the lack of data for atoms in intense magnetic fields. While the behaviour of hydrogen and helium has been investigated in detail (see, e.g., [2]), complete and reliable data for heavier elements, in particular iron, are still missing. Since neutron stars are formed by the collapse of the iron cores of massive stars, it may be assumed that their atmospheres contain an iron plasma. Our objective is to fill the gap

  2. Multiple-code benchmark simulation study of coupled THMC processesin the excavation disturbed zone associated with geological nuclear wasterepositories

    SciTech Connect

    Rutqvist, J.; Feng, X-T.; Hudson, J.; Jing, L.; Kobayashi, A.; Koyama, T.; Pan, P-Z.; Lee, H-S.; Rinne, M.; Sonnenthal, E.; Yamamoto, Y.

    2006-05-10

    An international, multiple-code benchmark test (BMT) studyis being conducted within the international DECOVALEX project to analysecoupled thermal, hydrological, mechanical and chemical (THMC) processesin the excavation disturbed zone (EDZ) around emplacement drifts of anuclear waste repository. This BMT focuses on mechanical responses andlong-term chemo-mechanical effects that may lead to changes in mechanicaland hydrological properties in the EDZ. This includes time-de-pendentprocesses such as creep, and subcritical crack, or healing of fracturesthat might cause "weakening" or "hardening" of the rock over the longterm. Five research teams are studying this BMT using a wide range ofmodel approaches, including boundary element, finite element, and finitedifference, particle mechanics, and elasto-plastic cellular automatamethods. This paper describes the definition of the problem andpreliminary simulation results for the initial model inception part, inwhich time dependent effects are not yet included.

  3. Multigroup neutron dose calculations for proton therapy

    SciTech Connect

    Kelsey Iv, Charles T; Prinja, Anil K

    2009-01-01

    We have developed tools for the preparation of coupled multigroup proton/neutron cross section libraries. Our method is to use NJOY to process evaluated nuclear data files for incident particles below 150 MeV and MCNPX to produce data for higher energies. We modified the XSEX3 program of the MCNPX code system to produce Legendre expansions of scattering matrices generated by sampling the physics models that are comparable to the output of the GROUPR routine of NJOY. Our code combines the low and high energy scattering data with user input stopping powers and energy deposition cross sections that we also calculated using MCNPX. Our code also calculates momentum transfer coefficients for the library and optionally applies an energy straggling model to the scattering cross sections and stopping powers. The motivation was initially for deterministic solution of space radiation shielding calculations using Attila, but noting that proton therapy treatment planning may neglect secondary neutron dose assessments because of difficulty and expense, we have also investigated the feasibility of multi group methods for this application. We have shown that multigroup MCNPX solutions for secondary neutron dose compare well with continuous energy solutions and are obtainable with less than half computational cost. This efficiency comparison neglects the cost of preparing the library data, but this becomes negligible when distributed over many multi group calculations. Our deterministic calculations illustrate recognized obstacles that may have to be overcome before discrete ordinates methods can be efficient alternatives for proton therapy neutron dose calculations.

  4. A User's Manual for MASH V1.5 - A Monte Carlo Adjoint Shielding Code System

    SciTech Connect

    C. O. Slater; J. M. Barnes; J. O. Johnson; J.D. Drischler

    1998-10-01

    The Monte Carlo ~djoint ~ielding Code System, MASH, calculates neutron and gamma- ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air- over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system includes the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. The current version, MASH v 1.5, is the successor to the original MASH v 1.0 code system initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the "dose importance" of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response as a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user's manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem.

  5. Inductively coupled plasma mass spectrometry in comparison with neutron activation and ion chromatography with UV/VIS detection for the determination of lanthanides in plant materials.

    PubMed

    Bulska, Ewa; Danko, Bożena; Dybczyński, Rajmund S; Krata, Agnieszka; Kulisa, Krzysztof; Samczyński, Zbigniew; Wojciechowski, Marcin

    2012-08-15

    Analytical performance of inductively coupled plasma mass spectrometry (ICP-MS) for determination of lanthanides in plant materials was investigated and compared with neutron activation analysis (NAA) as well as ion chromatography (IC) with UV-VIS detection. Two sample preparation protocols were tested: (i) microwave assisted digestion by concentrated nitric acid; (ii) microwave digestion involving silica and fluoride removal, followed by the selective and quantitative lanthanides group separation from the plant matrix. Several Certified Reference Materials (CRM) of plant origin were used for the evaluation of the accuracy of the applied analytical procedures. The consistency of results, obtained by various methods, enabled to establish the tentative recommended values (TRV) for several missing elements in one of CRMs. The ICP-MS, due to its very high sensitivity, has the potential to contribute to this aim. The discrepancy of the results obtained by various methods was discussed in a view of possible matrix effects related to the composition of investigated materials. PMID:22841084

  6. Neutron generator power supply modeling in EMMA

    SciTech Connect

    Robinson, A.C.; Farnsworth, A.V.; Montgomery, S.T.; Peery, J.S; Merewether, K.O.

    1996-12-01

    Sandia National Laboratories has prime responsibility for neutron generator design and manufacturing, and is committed to developing predictive tools for modeling neutron generator performance. An important aspect of understanding component performance is explosively driven ferroelectric power supply modeling. EMMA (ElectroMechanical Modeling in ALEGRA) is a three dimensional compile time version of Sandia`s ALEGRA code. The code is built on top of the general ALEGRA framework for parallel shock-physics computations but also includes additional capability for modeling the electric potential field in dielectrics. The overall package includes shock propagation due to explosive detonation, depoling of ferroelectric ceramics, electric field calculation and coupling with a general lumped element circuit equation system. The AZTEC parallel iterative solver is used to solve for the electric potential. The DASPK differential algebraic equation package is used to solve the circuit equation system. Sample calculations are described.

  7. Finite element analysis and frequency shift studies for the bridge coupler of the coupled cavity linear accelerator of the spallation neutron source.

    SciTech Connect

    Chen, Z.

    2001-01-01

    The Spallation Neutron Source (SNS) is an accelerator-based neutron scattering research facility. The linear accelerator (linac) is the principal accelerating structure and divided into a room-temperature linac and a superconducting linac. The normal conducting linac system that consists of a Drift Tube Linac (DTL) and a Coupled Cavity Linac (CCL) is to be built by Los Alamos National Laboratory. The CCL structure is 55.36-meters long. It accelerates H- beam from 86.8 Mev to 185.6 Mev at operating frequency of 805 MHz. This side coupled cavity structure has 8 cells per segment, 12 segments and 11 bridge couplers per module, and 4 modules total. A 5-MW klystron powers each module. The number 3 and number 9 bridge coupler of each module are connected to the 5-MW RF power supply. The bridge coupler with length of 2.5 {beta}{gamma} is a three-cell structure and located between the segments and allows power flow through the module. The center cell of each bridge coupler is excited during normal operation. To obtain a uniform electromagnetic filed and meet the resonant frequency shift, the RF induced heat must be removed. Thus, the thermal deformation and frequency shift studies are performed via numerical simulations in order to have an appropriate cooling design and predict the frequency shift under operation. The center cell of the bridge coupler also contains a large 4-inch slug tuner and a tuning post that used to provide bulk frequency adjustment and field intensity adjustment, so that produce the proper total field distribution in the module assembly.

  8. Multiple-Code BenchMaek Simulation Stidy of Coupled THMC Processes IN the EXCAVATION DISTURBED ZONE Associated with Geological Nuclear Waste Repositories

    SciTech Connect

    J. Rutqvist; X. Feng; J. Hudson; L. Jing; A. Kobayashi; T. Koyama; P.Pan; H. Lee; M. Rinne; E. Sonnenthal; Y. Yamamoto

    2006-05-08

    An international, multiple-code benchmark test (BMT) study is being conducted within the international DECOVALEX project to analyze coupled thermal, hydrological, mechanical and chemical (THMC) processes in the excavation disturbed zone (EDZ) around emplacement drifts of a nuclear waste repository. This BMT focuses on mechanical responses and long-term chemo-mechanical effects that may lead to changes in mechanical and hydrological properties in the EDZ. This includes time-dependent processes such as creep, and subcritical crack, or healing of fractures that might cause ''weakening'' or ''hardening'' of the rock over the long term. Five research teams are studying this BMT using a wide range of model approaches, including boundary element, finite element, and finite difference, particle mechanics, and elasto-plastic cellular automata methods. This paper describes the definition of the problem and preliminary simulation results for the initial model inception part, in which time dependent effects are not yet included.

  9. Isotope Coded Protein Labeling Coupled Immunoprecipitation (ICPL-IP): A Novel Approach for Quantitative Protein Complex Analysis From Native Tissue*

    PubMed Central

    Vogt, Andreas; Fuerholzner, Bettina; Kinkl, Norbert; Boldt, Karsten; Ueffing, Marius

    2013-01-01

    High confidence definition of protein interactions is an important objective toward the understanding of biological systems. Isotope labeling in combination with affinity-based isolation of protein complexes has increased in accuracy and reproducibility, yet, larger organisms—including humans—are hardly accessible to metabolic labeling and thus, a major limitation has been its restriction to small animals, cell lines, and yeast. As composition as well as the stoichiometry of protein complexes can significantly differ in primary tissues, there is a great demand for methods capable to combine the selectivity of affinity-based isolation as well as the accuracy and reproducibility of isotope-based labeling with its application toward analysis of protein interactions from intact tissue. Toward this goal, we combined isotope coded protein labeling (ICPL)1 with immunoprecipitation (IP) and quantitative mass spectrometry (MS). ICPL-IP allows sensitive and accurate analysis of protein interactions from primary tissue. We applied ICPL-IP to immuno-isolate protein complexes from bovine retinal tissue. Protein complexes of immunoprecipitated β-tubulin, a highly abundant protein with known interactors as well as the lowly expressed small GTPase RhoA were analyzed. The results of both analyses demonstrate sensitive and selective identification of known as well as new protein interactions by our method. PMID:23268931

  10. Isotope coded protein labeling coupled immunoprecipitation (ICPL-IP): a novel approach for quantitative protein complex analysis from native tissue.

    PubMed

    Vogt, Andreas; Fuerholzner, Bettina; Kinkl, Norbert; Boldt, Karsten; Ueffing, Marius

    2013-05-01

    High confidence definition of protein interactions is an important objective toward the understanding of biological systems. Isotope labeling in combination with affinity-based isolation of protein complexes has increased in accuracy and reproducibility, yet, larger organisms--including humans--are hardly accessible to metabolic labeling and thus, a major limitation has been its restriction to small animals, cell lines, and yeast. As composition as well as the stoichiometry of protein complexes can significantly differ in primary tissues, there is a great demand for methods capable to combine the selectivity of affinity-based isolation as well as the accuracy and reproducibility of isotope-based labeling with its application toward analysis of protein interactions from intact tissue. Toward this goal, we combined isotope coded protein labeling (ICPL)(1) with immunoprecipitation (IP) and quantitative mass spectrometry (MS). ICPL-IP allows sensitive and accurate analysis of protein interactions from primary tissue. We applied ICPL-IP to immuno-isolate protein complexes from bovine retinal tissue. Protein complexes of immunoprecipitated β-tubulin, a highly abundant protein with known interactors as well as the lowly expressed small GTPase RhoA were analyzed. The results of both analyses demonstrate sensitive and selective identification of known as well as new protein interactions by our method. PMID:23268931

  11. Development of a new code to solve hydro-mechanical coupling, shear failure and tensile failure due to hydraulic fracturing operations.

    NASA Astrophysics Data System (ADS)

    María Gómez Castro, Berta; De Simone, Silvia; Carrera, Jesús

    2016-04-01

    Nowadays, there are still some unsolved relevant questions which must be faced if we want to proceed to the hydraulic fracturing in a safe way. How much will the fracture propagate? This is one of the most important questions that have to be solved in order to avoid the formation of pathways leading to aquifer targets and atmospheric release. Will the fracture failure provoke a microseismic event? Probably this is the biggest fear that people have in fracking. The aim of this work (developed as a part of the EU - FracRisk project) is to understand the hydro-mechanical coupling that controls the shear of existing fractures and their propagation during a hydraulic fracturing operation, in order to identify the key parameters that dominate these processes and answer the mentioned questions. This investigation focuses on the development of a new C++ code which simulates hydro-mechanical coupling, shear movement and propagation of a fracture. The framework employed, called Kratos, uses the Finite Element Method and the fractures are represented with an interface element which is zero thickness. This means that both sides of the element lie together in the initial configuration (it seems a 1D element in a 2D domain, and a 2D element in a 3D domain) and separate as the adjacent matrix elements deform. Since we are working in hard, fragile rocks, we can assume an elastic matrix and impose irreversible displacements in fractures when rock failure occurs. The formulation used to simulate shear and tensile failures is based on the analytical solution proposed by Okada, 1992 and it is part of an iterative process. In conclusion, the objective of this work is to employ the new code developed to analyze the main uncertainties related with the hydro-mechanical behavior of fractures derived from the hydraulic fracturing operations.

  12. Initial Coupling of the RELAP-7 and PRONGHORN Applications

    SciTech Connect

    J. Ortensi; D. Andrs; A.A. Bingham; R.C. Martineau; J.W. Peterson

    2012-10-01

    Modern nuclear reactor safety codes require the ability to solve detailed coupled neutronic- thermal fluids problems. For larger cores, this implies fully coupled higher dimensionality spatial dynamics with appropriate feedback models that can provide enough resolution to accurately compute core heat generation and removal during steady and unsteady conditions. The reactor analysis code PRONGHORN is being coupled to RELAP-7 as a first step to extend RELAP’s current capabilities. This report details the mathematical models, the type of coupling, and the testing results from the integrated system. RELAP-7 is a MOOSE-based application that solves the continuity, momentum, and energy equations in 1-D for a compressible fluid. The pipe and joint capabilities enable it to model parts of the power conversion unit. The PRONGHORN application, also developed on the MOOSE infrastructure, solves the coupled equations that define the neutron diffusion, fluid flow, and heat transfer in a full core model. The two systems are loosely coupled to simplify the transition towards a more complex infrastructure. The integration is tested on a simplified version of the OECD/NEA MHTGR-350 Coupled Neutronics-Thermal Fluids benchmark model.

  13. Coupling of mRNA Structure Rearrangement to Ribosome Movement during Bypassing of Non-coding Regions.

    PubMed

    Chen, Jin; Coakley, Arthur; O'Connor, Michelle; Petrov, Alexey; O'Leary, Seán E; Atkins, John F; Puglisi, Joseph D

    2015-11-19

    Nearly half of the ribosomes translating a particular bacteriophage T4 mRNA bypass a region of 50 nt, resuming translation 3' of this gap. How this large-scale, specific hop occurs and what determines whether a ribosome bypasses remain unclear. We apply single-molecule fluorescence with zero-mode waveguides to track individual Escherichia coli ribosomes during translation of T4's gene 60 mRNA. Ribosomes that bypass are characterized by a 10- to 20-fold longer pause in a non-canonical rotated state at the take-off codon. During the pause, mRNA secondary structure rearrangements are coupled to ribosome forward movement, facilitated by nascent peptide interactions that disengage the ribosome anticodon-codon interactions for slippage. Close to the landing site, the ribosome then scans mRNA in search of optimal base-pairing interactions. Our results provide a mechanistic and conformational framework for bypassing, highlighting a non-canonical ribosomal state to allow for mRNA structure refolding to drive large-scale ribosome movements. PMID:26590426

  14. A fast parallel code for calculating energies and oscillator strengths of many-electron atoms at neutron star magnetic field strengths in adiabatic approximation

    NASA Astrophysics Data System (ADS)

    Engel, D.; Klews, M.; Wunner, G.

    2009-02-01

    We have developed a new method for the fast computation of wavelengths and oscillator strengths for medium-Z atoms and ions, up to iron, at neutron star magnetic field strengths. The method is a parallelized Hartree-Fock approach in adiabatic approximation based on finite-element and B-spline techniques. It turns out that typically 15-20 finite elements are sufficient to calculate energies to within a relative accuracy of 10-5 in 4 or 5 iteration steps using B-splines of 6th order, with parallelization speed-ups of 20 on a 26-processor machine. Results have been obtained for the energies of the ground states and excited levels and for the transition strengths of astrophysically relevant atoms and ions in the range Z=2…26 in different ionization stages. Catalogue identifier: AECC_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AECC_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 3845 No. of bytes in distributed program, including test data, etc.: 27 989 Distribution format: tar.gz Programming language: MPI/Fortran 95 and Python Computer: Cluster of 1-26 HP Compaq dc5750 Operating system: Fedora 7 Has the code been vectorised or parallelized?: Yes RAM: 1 GByte Classification: 2.1 External routines: MPI/GFortran, LAPACK, PyLab/Matplotlib Nature of problem: Calculations of synthetic spectra [1] of strongly magnetized neutron stars are bedevilled by the lack of data for atoms in intense magnetic fields. While the behaviour of hydrogen and helium has been investigated in detail (see, e.g., [2]), complete and reliable data for heavier elements, in particular iron, are still missing. Since neutron stars are formed by the collapse of the iron cores of massive stars, it may be assumed that their atmospheres contain an iron plasma. Our objective is to fill the gap

  15. Magnetized Neutron Stars

    NASA Astrophysics Data System (ADS)

    Liebling, Steven; Anderson, Matthew; Hirschmann, Eric; Lehner, Luis; Motl, Patrick; Neilsen, David; Palenzuela, Carlos; Tohline, Joel

    2008-04-01

    Magnetized neutron stars, whether considered individually or within compact binary systems, demonstrate a number of interesting dynamical effects. Using a distributed adaptive mesh refinement (AMR) code, we evolve such stars and study their dynamics.

  16. FOREWORD: Neutron metrology Neutron metrology

    NASA Astrophysics Data System (ADS)

    Thomas, David J.; Nolte, Ralf; Gressier, Vincent

    2011-12-01

    industry, from the initial fuel enrichment and fabrication processes right through to storage or reprocessing, and neutron metrology is clearly important in this area. Neutron fields do, however, occur in other areas, for example where neutron sources are used in oil well logging and moisture measurements. They also occur around high energy accelerators, including photon linear accelerators used for cancer therapy, and are expected to be a more serious problem around the new hadron radiation therapy facilities. Roughly 50% of the cosmic ray doses experienced by fliers at the flight altitudes of commercial aircraft are due to neutrons. Current research on fusion presents neutron metrology with a whole new range of challenges because of the very high fluences expected. One of the most significant features of neutron fields is the very wide range of possible neutron energies. In the nuclear industry, for example, neutrons occur with energies from those of thermal neutrons at a few meV to the upper end of the fission spectrum at perhaps 10 MeV. For cosmic ray dosimetry the energy range extends into the GeV region. This enormous range sets a challenge for designing measuring devices and a parallel challenge of developing measurement standards for characterizing these devices. One of the major considerations when deciding on topics for this special issue was agreeing on what not to include. Modelling, i.e. the use of radiation transport codes, is now a very important aspect of neutron measurements. These calculations are vital for shielding and for instrument design; nevertheless, the topic has only been included here where it has a direct bearing on metrology and the development of standards. Neutron spectrometry is an increasingly important technique for unravelling some of the problems of dose equivalent measurements and for plasma diagnostics in fusion research. However, this topic is at least one step removed from primary metrology and so it was felt that it should not be

  17. Ex-vessel neutron dosimetry analysis for westinghouse 4-loop XL pressurized water reactor plant using the RadTrack{sup TM} Code System with the 3D parallel discrete ordinates code RAPTOR-M3G

    SciTech Connect

    Chen, J.; Alpan, F. A.; Fischer, G.A.; Fero, A.H.

    2011-07-01

    Traditional two-dimensional (2D)/one-dimensional (1D) SYNTHESIS methodology has been widely used to calculate fast neutron (>1.0 MeV) fluence exposure to reactor pressure vessel in the belt-line region. However, it is expected that this methodology cannot provide accurate fast neutron fluence calculation at elevations far above or below the active core region. A three-dimensional (3D) parallel discrete ordinates calculation for ex-vessel neutron dosimetry on a Westinghouse 4-Loop XL Pressurized Water Reactor has been done. It shows good agreement between the calculated results and measured results. Furthermore, the results show very different fast neutron flux values at some of the former plate locations and elevations above and below an active core than those calculated by a 2D/1D SYNTHESIS method. This indicates that for certain irregular reactor internal structures, where the fast neutron flux has a very strong local effect, it is required to use a 3D transport method to calculate accurate fast neutron exposure. (authors)

  18. Pulsed-neutron monochromator

    DOEpatents

    Mook, H.A. Jr.

    1984-01-01

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

  19. Pulsed-neutron monochromator

    DOEpatents

    Mook, Jr., Herbert A.

    1985-01-01

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

  20. A user`s manual for MASH 1.0: A Monte Carlo Adjoint Shielding Code System

    SciTech Connect

    Johnson, J.O.

    1992-03-01

    The Monte Carlo Adjoint Shielding Code System, MASH, calculates neutron and gamma-ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air-over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system include the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. MASH is the successor to the Vehicle Code System (VCS) initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the ``dose importance`` of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response a a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user`s manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem (input data and selected output edits) for each code.

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

    SciTech Connect

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

    SciTech Connect

    Biondo, Elliott D; Ibrahim, Ahmad M; Mosher, Scott W; Grove, Robert E

    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).

  3. Fusion and neutron transfer reactions with weakly bound nuclei within time-dependent and coupled channel approaches

    NASA Astrophysics Data System (ADS)

    Samarin, V. V.

    2016-05-01

    The time-dependent Schrödinger equation and the coupled channel approach based on the method of perturbed stationary two-center states are used to describe nucleon transfers and fusion in low-energy nuclear reactions. Results of the cross sections calculation for the formation of the 198Au and fusion in the 6He+197Au reaction and for the formation of the 65Zn in 6He+64Zn reaction agree satisfactorily with the experimental data near the barrier. The Feynman's continual integrals calculations for a few-body systems were used for the proposal of the new form of the shell model mean field for helium isotopes.

  4. Certification of Total Arsenic in Blood and Urine Standard Reference Materials by Radiochemical Neutron Activation Analysis and Inductively Coupled Plasma - Mass Spectrometry

    PubMed Central

    Paul, Rick L.; Davis, W. Clay; Yu, Lee; Murphy, Karen E.; Guthrie, William F.; Leber, Dennis D.; Bryan, Colleen E.; Vetter, Thomas W.; Shakirova, Gulchekhra; Mitchell, Graylin; Kyle, David J.; Jarrett, Jeffery M.; Caldwell, Kathleen L.; Jones, Robert L.; Eckdahl, Steven; Wermers, Michelle; Maras, Melissa; Palmer, C. D.; Verostek, M.F.; Geraghty, C. M.; Steuerwald, Amy J.; Parsons, Patrick J.

    2015-01-01

    A newly developed procedure for determination of arsenic by radiochemical neutron activation analysis (RNAA) was used to measure arsenic at four levels in SRM 955c Toxic Elements in Caprine Blood and at two levels in SRM 2668 Toxic Elements in Frozen Human Urine for the purpose of providing mass concentration values for certification. Samples were freeze-dried prior to analysis followed by neutron irradiation for 3 h at a fluence rate of 1×1014cm−2s−1. After sample dissolution in perchloric and nitric acids, arsenic was separated from the matrix by extraction into zinc diethyldithiocarbamate in chloroform, and 76As quantified by gamma-ray spectroscopy. Differences in chemical yield and counting geometry between samples and standards were monitored by measuring the count rate of a 77As tracer added before sample dissolution. RNAA results were combined with inductively coupled plasma – mass spectrometry (ICP-MS) values from NIST and collaborating laboratories to provide certified values of (10.81 ± 0.54) μg/kg and (213.1 ± 0.73) μg/kg for SRM 2668 Levels I and II, and certified values of (21.66 ± 0.73) μg/kg, (52.7 ± 1.1) μg/kg, and (78.8 ± 4.9) μg/kg for SRM 955c Levels 2, 3, and 4 respectively. Because of discrepancies between values obtained by different methods for SRM 955c Level 1, an information value of < 5 μg/kg was assigned for this material. PMID:26300575

  5. FOREWORD: Neutron metrology Neutron metrology

    NASA Astrophysics Data System (ADS)

    Thomas, David J.; Nolte, Ralf; Gressier, Vincent

    2011-12-01

    industry, from the initial fuel enrichment and fabrication processes right through to storage or reprocessing, and neutron metrology is clearly important in this area. Neutron fields do, however, occur in other areas, for example where neutron sources are used in oil well logging and moisture measurements. They also occur around high energy accelerators, including photon linear accelerators used for cancer therapy, and are expected to be a more serious problem around the new hadron radiation therapy facilities. Roughly 50% of the cosmic ray doses experienced by fliers at the flight altitudes of commercial aircraft are due to neutrons. Current research on fusion presents neutron metrology with a whole new range of challenges because of the very high fluences expected. One of the most significant features of neutron fields is the very wide range of possible neutron energies. In the nuclear industry, for example, neutrons occur with energies from those of thermal neutrons at a few meV to the upper end of the fission spectrum at perhaps 10 MeV. For cosmic ray dosimetry the energy range extends into the GeV region. This enormous range sets a challenge for designing measuring devices and a parallel challenge of developing measurement standards for characterizing these devices. One of the major considerations when deciding on topics for this special issue was agreeing on what not to include. Modelling, i.e. the use of radiation transport codes, is now a very important aspect of neutron measurements. These calculations are vital for shielding and for instrument design; nevertheless, the topic has only been included here where it has a direct bearing on metrology and the development of standards. Neutron spectrometry is an increasingly important technique for unravelling some of the problems of dose equivalent measurements and for plasma diagnostics in fusion research. However, this topic is at least one step removed from primary metrology and so it was felt that it should not be

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

    SciTech Connect

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

    2012-07-01

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

  7. Magnetic excitations in an exchange-coupled tetramer cluster of cobalt (II): a study by inelastic neutron scattering

    NASA Astrophysics Data System (ADS)

    Gómez-García, C. J.; Coronado, E.; Borrás-Almenar, J. J.; Aebersold, M.; Güdel, H. U.; Mutka, H.

    1992-06-01

    The polyoxometalate K 10[Co 4(H 2O) 2(PW 9O 34) 2].20H 2O contains a ferromagnetically exchange-coupled tetramer of Co 11 encapsulated in between two diamagnetic molecules (PW 9O 34-9. At 2.5K several inelastic peaks are observed in the energy range 1.5-7 meV, which are assigned to magnetic excitations in the cluster. A tentative interpretation of these data from an anisotropic exchange model yields a cobalt-cobalt interaction of 3meV (24cm -1) and an amount of anisotropy Jxy/ Jz≌0.6. These values are consistent with the magnetic susceptibility measurements.

  8. Interfaces MATXS Cross-Section Libraries to Nuclear Transport Codes for Fusion Systems Analysis.

    Energy Science and Technology Software Center (ESTSC)

    1985-04-10

    Version: 00 TRANSX-CTR is a computer code that reads nuclear data from a library in MATXS format and produces transport tables with many discrete-ordinates (Sn) and diffusion codes. Tables can be produced for neutron, photon, or coupled transport. Options include adjoint tables, mixtures, self-shielding, group collapse, homogenization, thermal upscatter, prompt or steady-state fission, transport corrections, elastic removal corrections, and flexible response-function edits. The ability to prepare coupled tables and response edits for heating, damage, gasmore » production, and delayed activity makes TRANSX-CTR especially useful for fusion reactor studies.« less

  9. Methodology, status and plans for development and assessment of Cathare code

    SciTech Connect

    Bestion, D.; Barre, F.; Faydide, B.

    1997-07-01

    This paper presents the methodology, status and plans for the development, assessment and uncertainty evaluation of the Cathare code. Cathare is a thermalhydraulic code developed by CEA (DRN), IPSN, EDF and FRAMATOME for PWR safety analysis. First, the status of the code development and assessment is presented. The general strategy used for the development and the assessment of the code is presented. Analytical experiments with separate effect tests, and component tests are used for the development and the validation of closure laws. Successive Revisions of constitutive laws are implemented in successive Versions of the code and assessed. System tests or integral tests are used to validate the general consistency of the Revision. Each delivery of a code Version + Revision is fully assessed and documented. A methodology is being developed to determine the uncertainty on all constitutive laws of the code using calculations of many analytical tests and applying the Discrete Adjoint Sensitivity Method (DASM). At last, the plans for the future developments of the code are presented. They concern the optimization of the code performance through parallel computing - the code will be used for real time full scope plant simulators - the coupling with many other codes (neutronic codes, severe accident codes), the application of the code for containment thermalhydraulics. Also, physical improvements are required in the field of low pressure transients and in the modeling for the 3-D model.

  10. Neutronics Modeling of the High Flux Isotope Reactor using COMSOL

    SciTech Connect

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

    2011-01-01

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

  11. Inelastic neutron scattering spectra of a hydrogen molecule in a nanocavity: Methodology for quantum calculations incorporating the coupled five-dimensional translation-rotation eigenstates

    NASA Astrophysics Data System (ADS)

    Xu, Minzhong; Bačić, Zlatko

    2011-11-01

    We present an in-depth description of the methodology for accurate quantum calculation of the inelastic neutron scattering (INS) spectra of an H2 molecule confined inside a nanosize cavity of an arbitrary shape. This methodology was introduced in a recent work [M. Xu, L. Ulivi, M. Celli, D. Colognesi, and Z. Bačić, Phys. Rev. BPRBMDO1098-012110.1103/PhysRevB.83.241403 83, 241403(R) (2011)], where the INS spectra of para- and ortho-H2 in the small cage of the structure II clathrate hydrate were simulated and compared with the measured spectra. The key distinctive feature of our approach, and its main strength and advantage, is the use of the coupled quantum 5D translation-rotation (TR) energy levels and wave functions of the entrapped H2 molecule, rigorously calculated on the 5D intermolecular potential energy surface (PES), as the initial and the final states of the INS transitions. In this work, we describe the implementation of the 5D TR wave functions within the quantum INS formalism, and obtain the working expressions for the matrix elements required to compute the INS spectra of the nanoconfined H2 molecule. The computational approach devised for efficient calculation of the 5D TR eigenstates in the compact contracted basis, indispensable for our quantum simulation of the INS spectra, is presented as well. Since the TR coupling is fully taken into account, the computed INS spectra exhibit a uniquely high degree of realism and faithfully reflect the quantum dynamics of H2 on the PES of the host environment.

  12. Development and implementation of a finite element solution of the coupled neutron transport and thermoelastic equations governing the behavior of small nuclear assemblies

    NASA Astrophysics Data System (ADS)

    Wilson, Stephen Christian

    Small, highly enriched reactors designed for weapons effects simulations undergo extreme thermal transients during pulsed operations. The primary shutdown mechanism of these reactors---thermal expansion of fuel material---experiences an inertial delay resulting in a different value for the fuel temperature coefficient of reactivity during pulse operation as compared to the value appropriate for steady-state operation. The value appropriate for pulsed operation may further vary as a function of initial reactivity addition. Here we design and implement a finite element numerical method to predict the pulse operation behavior of Sandia Pulsed Reactor (SPR) II, SPR III, and a hypothetical spherical assembly with identical fuel properties without using operationally observed data in our model. These numerical results are compared to available SPR II and SPR III operational data. The numerical methods employed herein may be modified and expanded in functionality to provide both accurate characterization of the behavior of fast burst reactors of any common geometry or isotropic fuel material in the design phase, as well as a computational tool for general coupled thermomechanical-neutronics behavior in the solid state for any reactor type.

  13. A Hybrid Reflective/Refractive/Diffractive Achromatic Fiber-Coupled Radiation Resistant Imaging System for Use in the Spallation Neutron Source (SNS)

    SciTech Connect

    Maxey, L Curt; Ally, Tanya R; Brunson, Aly; Garcia, Frances; Goetz, Kathleen C; Hasse, Katelyn E; McManamy, Thomas J; Shea, Thomas J; Simpson, Marc Livingstone

    2011-01-01

    A fiber-coupled imaging system for monitoring the proton beam profile on the target of the Spallation Neutron Source was developed using reflective, refractive and diffractive optics to focus an image onto a fiber optic imaging bundle. The imaging system monitors the light output from a chromium-doped aluminum oxide (Al{sub 2}0{sub 3}:Cr) scintillator on the nose of the target. Metal optics are used to relay the image to the lenses that focus the image onto the fiber. The material choices for the lenses and fiber were limited to high-purity fused silica, due to the anticipated radiation dose of 10{sup 8} R. In the first generation system (which had no diffractive elements), radiation damage to the scintillator on the nose of the target significantly broadened the normally monochromatic (694 nm) spectrum. This created the need for an achromatic design in the second generation system. This was achieved through the addition of a diffractive optic for chromatic correction. An overview of the target imaging system and its performance, with particular emphasis on the design and testing of a hybrid refractive/diffractive high-purity fused silica imaging triplet, is presented.

  14. Binary neutron-star mergers with Whisky and SACRA: First quantitative comparison of results from independent general-relativistic hydrodynamics codes

    NASA Astrophysics Data System (ADS)

    Baiotti, Luca; Shibata, Masaru; Yamamoto, Tetsuro

    2010-09-01

    We present the first quantitative comparison of two independent general-relativistic hydrodynamics codes, the whisky code and the sacra code. We compare the output of simulations starting from the same initial data and carried out with the configuration (numerical methods, grid setup, resolution, gauges) which for each code has been found to give consistent and sufficiently accurate results, in particular, in terms of cleanness of gravitational waveforms. We focus on the quantities that should be conserved during the evolution (rest mass, total mass energy, and total angular momentum) and on the gravitational-wave amplitude and frequency. We find that the results produced by the two codes agree at a reasonable level, with variations in the different quantities but always at better than about 10%.

  15. Boron neutron capture enhancement (BNCE) of fast neutron irradiation for glioblastoma: increase of thermal neutron flux with heavy material collimation, a theoretical evaluation.

    PubMed

    Paquis, P; Pignol, J P; Lonjon, M; Brassart, N; Courdi, A; Chauvel, P; Grellier, P; Chatel, M

    1999-01-01

    Despite the fact that fast neutron irradiation of glioblastoma has shown on autopsies an ability to sterilize tumors, no therapeutic windows have been found for these particles due to their toxicity toward normal brain. Therefore, the Boron Neutron Capture Enhancement (BNCE) of fast neutron beam has been suggested. This paper addresses the problem of fast neutron beam collimation, which induces a dramatic decrease of the thermal neutron flux in the depth of the tissues when smaller irradiation fields are used. Thermoluminescent dosimeter TLD-600 and TLD-700 were used to determine the thermal neutron flux within a Plexiglas phantom irradiated under the Nice Biomedical Cyclotron p(60)+Be(32) fast neutron beam. A BNCE of 4.6% in physical dose was determined for a 10 x 10 cm2 field, and of 10.4% for a 20 x 20 cm2 one. A Dose Modification Factor of 1.19 was calculated for CAL 58 glioblastoma cells irradiated thanks to the larger field. In order to increase the thermal flux in depth while shaping the beam, heavy material collimation was studied with Monte Carlo simulations using coupled FLUKA and MCNP-4A codes. The use of 20 cm width lead blocks allowed a 2 fold thermal neutron flux increase in the depth of the phantom, while shielding the fast neutron beam with a fast neutron dose transmission of 23%. Using the DMF of 1.19, a BNCE of 40% was calculated in the beam axis. This enhancement might be sufficient to open, at least theoretically, a therapeutic window. PMID:10222419

  16. MSTor version 2013: A new version of the computer code for the multi-structural torsional anharmonicity, now with a coupled torsional potential

    NASA Astrophysics Data System (ADS)

    Zheng, Jingjing; Meana-Pañeda, Rubén; Truhlar, Donald G.

    2013-08-01

    We present an improved version of the MSTor program package, which calculates partition functions and thermodynamic functions of complex molecules involving multiple torsions; the method is based on either a coupled torsional potential or an uncoupled torsional potential. The program can also carry out calculations in the multiple-structure local harmonic approximation. The program package also includes seven utility codes that can be used as stand-alone programs to calculate reduced moment of inertia matrices by the method of Kilpatrick and Pitzer, to generate conformational structures, to calculate, either analytically or by Monte Carlo sampling, volumes for torsional subdomains defined by Voronoi tessellation of the conformational subspace, to generate template input files for the MSTor calculation and Voronoi calculation, and to calculate one-dimensional torsional partition functions using the torsional eigenvalue summation method. Restrictions: There is no limit on the number of torsions that can be included in either the Voronoi calculation or the full MS-T calculation. In practice, the range of problems that can be addressed with the present method consists of all multitorsional problems for which one can afford to calculate all the conformational structures and their frequencies. Unusual features: The method can be applied to transition states as well as stable molecules. The program package also includes the hull program for the calculation of Voronoi volumes, the symmetry program for determining point group symmetry of a molecule, and seven utility codes that can be used as stand-alone programs to calculate reduced moment-of-inertia matrices by the method of Kilpatrick and Pitzer, to generate conformational structures, to calculate, either analytically or by Monte Carlo sampling, volumes of the torsional subdomains defined by Voronoi tessellation of the conformational subspace, to generate template input files, and to calculate one-dimensional torsional

  17. A finite-element marker-in-cell simulation code for thermo-chemically coupled magma dynamics in a visco-elasto-plastic host rock

    NASA Astrophysics Data System (ADS)

    Keller, Tobias

    2014-05-01

    Many prominent geodynamic scenarios such as subduction zones, plate collision and orogeny formation, mid-ocean ridges, continental rifting, etc. involve a significant amount of active magmatism. However, many numerical simulations used to address related questions focus on the deformation of the solid rock phase only, sometimes taking into account the magma dynamics in the form of some parameterized weakening mechanism. On the other hand, simulations specifically developed for magma dynamics problems have largely been restricted to the context of mantle dynamics and are not designed to deal with brittle tectonic deformation of the rock matrix as it occurs in the lithosphere and crust. Here, a 2-D finite-element marker-in-cell numerical method is presented capable of simulating thermally and compositionally coupled two-phase flow problems in a realistically deforming mantle, lithosphere and crust. The modeling approach is based on a set of well accepted equations for the conservation of mass, momentum, energy and composition, completed by constitutive laws for visco-elasto-plastic shear and compaction stresses and a much simplified yet thermodynamically consistent melting model depending on temperature, pressure and composition. The simulation code is written in Matlab and is capable of solving up to 500k degrees of freedom (requiring 2m marker particles) within few minutes per time step on a standard desktop computer. Long-term simulations of that size require run times of one to three weeks. The non-linear system of equations is solved using a Picard iterative scheme, where the linearized system of equations is solved directly during each iterative step. Sufficient convergence is usually obtained within less than 10 non-linear iterations. Generally, this numerical method is versatile, accessible and efficient enough for a wide range of 2-D problems.

  18. Neutron detector and fabrication method thereof

    DOEpatents

    Bhandari, Harish B.; Nagarkar, Vivek V.; Ovechkina, Olena E.

    2016-08-16

    A neutron detector and a method for fabricating a neutron detector. The neutron detector includes a photodetector, and a solid-state scintillator operatively coupled to the photodetector. In one aspect, the method for fabricating a neutron detector includes providing a photodetector, and depositing a solid-state scintillator on the photodetector to form a detector structure.

  19. Stochastic analog neutron transport with TRIPOLI-4 and FREYA: Bayesian uncertainty quantification for neutron multiplicity counting

    DOE PAGESBeta

    Verbeke, J. M.; Petit, O.

    2016-06-01

    From nuclear safeguards to homeland security applications, the need for the better modeling of nuclear interactions has grown over the past decades. Current Monte Carlo radiation transport codes compute average quantities with great accuracy and performance; however, performance and averaging come at the price of limited interaction-by-interaction modeling. These codes often lack the capability of modeling interactions exactly: for a given collision, energy is not conserved, energies of emitted particles are uncorrelated, and multiplicities of prompt fission neutrons and photons are uncorrelated. Many modern applications require more exclusive quantities than averages, such as the fluctuations in certain observables (e.g., themore » neutron multiplicity) and correlations between neutrons and photons. In an effort to meet this need, the radiation transport Monte Carlo code TRIPOLI-4® was modified to provide a specific mode that models nuclear interactions in a full analog way, replicating as much as possible the underlying physical process. Furthermore, the computational model FREYA (Fission Reaction Event Yield Algorithm) was coupled with TRIPOLI-4 to model complete fission events. As a result, FREYA automatically includes fluctuations as well as correlations resulting from conservation of energy and momentum.« less

  20. Direct Neutron Capture Calculations with Covariant Density Functional Theory Inputs

    NASA Astrophysics Data System (ADS)

    Zhang, Shi-Sheng; Peng, Jin-Peng; Smith, Michael S.; Arbanas, Goran; Kozub, Ray L.

    2014-09-01

    Predictions of direct neutron capture are of vital importance for simulations of nucleosynthesis in supernovae, merging neutron stars, and other astrophysical environments. We calculate the direct capture cross sections for E1 transitions using nuclear structure information from a covariant density functional theory as input for the FRESCO coupled-channels reaction code. We find good agreement of our predictions with experimental cross section data on the double closed-shell targets 16O, 48Ca, and 90Zr, and the exotic nucleus 36S. Extensions of the technique for unstable nuclei and for large-scale calculations will be discussed. Predictions of direct neutron capture are of vital importance for simulations of nucleosynthesis in supernovae, merging neutron stars, and other astrophysical environments. We calculate the direct capture cross sections for E1 transitions using nuclear structure information from a covariant density functional theory as input for the FRESCO coupled-channels reaction code. We find good agreement of our predictions with experimental cross section data on the double closed-shell targets 16O, 48Ca, and 90Zr, and the exotic nucleus 36S. Extensions of the technique for unstable nuclei and for large-scale calculations will be discussed. Supported by the U.S. Dept. of Energy, Office of Nuclear Physics.

  1. Applications of the COG multiparticle Monte Carlo transport code to simulated imaging of complex objects

    SciTech Connect

    Buck, R M; Hall, J M

    1999-06-01

    COG is a major multiparticle simulation code in the LLNL Monte Carlo radiation transport toolkit. It was designed to solve deep-penetration radiation shielding problems in arbitrarily complex 3D geometries, involving coupled transport of photons, neutrons, and electrons. COG was written to provide as much accuracy as the underlying cross-sections will allow, and has a number of variance-reduction features to speed computations. Recently COG has been applied to the simulation of high- resolution radiographs of complex objects and the evaluation of contraband detection schemes. In this paper we will give a brief description of the capabilities of the COG transport code and show several examples of neutron and gamma-ray imaging simulations. Keywords: Monte Carlo, radiation transport, simulated radiography, nonintrusive inspection, neutron imaging.

  2. Compressible Astrophysics Simulation Code

    Energy Science and Technology Software Center (ESTSC)

    2007-07-18

    This is an astrophysics simulation code involving a radiation diffusion module developed at LLNL coupled to compressible hydrodynamics and adaptive mesh infrastructure developed at LBNL. One intended application is to neutrino diffusion in core collapse supernovae.

  3. Current and anticipated uses of thermal hydraulic codes in Korea

    SciTech Connect

    Kim, Kyung-Doo; Chang, Won-Pyo

    1997-07-01

    In Korea, the current uses of thermal hydraulic codes are categorized into 3 areas. The first application is in designing both nuclear fuel and NSSS. The codes have usually been introduced based on the technology transfer programs agreed between KAERI and the foreign vendors. Another area is in the supporting of the plant operations and licensing by the utility. The third category is research purposes. In this area assessments and some applications to the safety issue resolutions are major activities using the best estimate thermal hydraulic codes such as RELAP5/MOD3 and CATHARE2. Recently KEPCO plans to couple thermal hydraulic codes with a neutronics code for the design of the evolutionary type reactor by 2004. KAERI also plans to develop its own best estimate thermal hydraulic code, however, application range is different from KEPCO developing code. Considering these activities, it is anticipated that use of the best estimate hydraulic analysis code developed in Korea may be possible in the area of safety evaluation within 10 years.

  4. Laser generated neutron source for neutron resonance spectroscopy

    SciTech Connect

    Higginson, D. P.; Bartal, T.; McNaney, J. M.; Swift, D. C.; Hey, D. S.; Le Pape, S.; Mackinnon, A.; Kodama, R.; Tanaka, K. A.; Mariscal, D.; Beg, F. N.; Nakamura, H.; Nakanii, N.

    2010-10-15

    A neutron source for neutron resonance spectroscopy has been developed using high-intensity, short-pulse lasers. This technique will allow robust measurement of interior ion temperature of laser-shocked materials and provide insight into material equation of state. The neutron generation technique uses laser-accelerated protons to create neutrons in LiF through (p,n) reactions. The incident proton beam has been diagnosed using radiochromic film. This distribution is used as the input for a (p,n) neutron prediction code which is validated with experimentally measured neutron yields. The calculation infers a total fluence of 1.8x10{sup 9} neutrons, which are expected to be sufficient for neutron resonance spectroscopy temperature measurements.

  5. DRC2: A code with specialized applications for coupling localized Monte Carlo adjoint calculations with fluences from two-dimensional R-Z discrete ordinates air-over-ground calculations

    SciTech Connect

    Slater, C.O.

    1992-01-01

    The DRC2 code, which couples MASH or MASHX adjoint leakages with DORT 2-D discrete ordinates forward directional fluences, is described. The forward fluences are allowed to vary both axially and radially over the coupling surface, as opposed to the strictly axial variation allowed by the predecessor DRC code. Input instructions are presented along with descriptions and results from several sample problems. Results from the sample problems are used to compare DRC2 with DRC, DRC2 with DORT, and DRC2 with itself for the case of x-y dependence versus no x-y dependence of the forward fluence. The test problems demonstrate that for small systems DRC and DRC2 give essentially the same results. Some significant differences are noted for larger systems. Additionally, DRC2 results with no x-y dependence of the forward directional fluences are practically the same as those calculated by DRC.

  6. Radiation transport phenomena and modeling - part A: Codes

    SciTech Connect

    Lorence, L.J.

    1997-06-01

    The need to understand how particle radiation (high-energy photons and electrons) from a variety of sources affects materials and electronics has motivated the development of sophisticated computer codes that describe how radiation with energies from 1.0 keV to 100.0 GeV propagates through matter. Predicting radiation transport is the necessary first step in predicting radiation effects. The radiation transport codes that are described here are general-purpose codes capable of analyzing a variety of radiation environments including those produced by nuclear weapons (x-rays, gamma rays, and neutrons), by sources in space (electrons and ions) and by accelerators (x-rays, gamma rays, and electrons). Applications of these codes include the study of radiation effects on electronics, nuclear medicine (imaging and cancer treatment), and industrial processes (food disinfestation, waste sterilization, manufacturing.) The primary focus will be on coupled electron-photon transport codes, with some brief discussion of proton transport. These codes model a radiation cascade in which electrons produce photons and vice versa. This coupling between particles of different types is important for radiation effects. For instance, in an x-ray environment, electrons are produced that drive the response in electronics. In an electron environment, dose due to bremsstrahlung photons can be significant once the source electrons have been stopped.

  7. A Code For Calculating Self-Shielded Multigroup Neutron Cross Sections and Self-Shielding Factors From Preprocessed ENDF/B Basic Data Files.

    Energy Science and Technology Software Center (ESTSC)

    1990-11-20

    Version 00 REX2-87 is a computer code developed for the calculation of self-shielded multigroup average cross sections, and self-shielding factors for total, elastic, fission and capture processes from an ENDF/B formatted nuclear data file in which the tabulated cross sections follow linear interpolation throughout.

  8. Axion cooling of neutron stars

    NASA Astrophysics Data System (ADS)

    Sedrakian, Armen

    2016-03-01

    Cooling simulations of neutron stars and their comparison with the data from thermally emitting x-ray sources put constraints on the properties of axions, and by extension, of any light pseudoscalar dark matter particles, whose existence has been postulated to solve the strong-C P problem of QCD. We incorporate the axion emission by pair-breaking and formation processes by S - and P -wave nucleonic condensates in a benchmark code for cooling simulations, as well as provide fit formulas for the rates of these processes. Axion cooling of neutron stars has been simulated for 24 models covering the mass range 1 to 1.8 solar masses, featuring nonaccreted iron and accreted light-element envelopes, and a range of nucleon-axion couplings. The models are based on an equation state predicting conservative physics of superdense nuclear matter that does not allow for the onset of fast cooling processes induced by phase transitions to non-nucleonic forms of matter or high proton concentration. The cooling tracks in the temperature vs age plane were confronted with the (time-averaged) measured surface temperature of the central compact object in the Cas A supernova remnant as well as surface temperatures of three nearby middle-aged thermally emitting pulsars. We find that the axion coupling is limited to fa/107 GeV ≥(5 - 10 ) , which translates into an upper bound on axion mass ma≤(0.06 - 0.12 ) eV for Peccei-Quinn charges of the neutron |Cn|˜0.04 and proton |Cp|˜0.4 characteristic for hadronic models of axions.

  9. Simple Coupling of Reactor Physics Effects and Uncertain Nuances

    Energy Science and Technology Software Center (ESTSC)

    2012-08-27

    The "Simple Coupling of Reactor Physics Effects and Uncertain Nuances" (SCORPEUN) code is a simple r-z 1-group neutron diffusion code where each r-mesh is coupled to a single-flow-channel model that represents all flow-channels in that r-mesh. This 1-D model assesses q=m*Cp*deletaT for each z-mesh in that channel. This flow channel model is then coupled to a simple 1-D heat conduction model for ascertaining the peak center-line fuel temperature in a hypothetical pin assigned to thatmore » flow channel. The code has property lookup capability for water, Na, Zirc, HT9, metalic fuel, oxide fuel, etc. It has linear interpolation features for micro-scopic cross-sections with respect to coolant density and fuel temperature. ***This last feature has not been fully tested and may need development***. The interpolated microscopic cross-sections are then combined (using the water density from the T/H calculation) to generate macroscopic diffusion coefficient, removal cross-section and nu-sigmaF for each r-z mesh of the neutron diffusion code.« less

  10. Simple Coupling of Reactor Physics Effects and Uncertain Nuances

    SciTech Connect

    Bays, Samuel

    2012-08-27

    The "Simple Coupling of Reactor Physics Effects and Uncertain Nuances" (SCORPEUN) code is a simple r-z 1-group neutron diffusion code where each r-mesh is coupled to a single-flow-channel model that represents all flow-channels in that r-mesh. This 1-D model assesses q=m*Cp*deletaT for each z-mesh in that channel. This flow channel model is then coupled to a simple 1-D heat conduction model for ascertaining the peak center-line fuel temperature in a hypothetical pin assigned to that flow channel. The code has property lookup capability for water, Na, Zirc, HT9, metalic fuel, oxide fuel, etc. It has linear interpolation features for micro-scopic cross-sections with respect to coolant density and fuel temperature. ***This last feature has not been fully tested and may need development***. The interpolated microscopic cross-sections are then combined (using the water density from the T/H calculation) to generate macroscopic diffusion coefficient, removal cross-section and nu-sigmaF for each r-z mesh of the neutron diffusion code.

  11. Extension of the supercritical carbon dioxide brayton cycle to low reactor power operation: investigations using the coupled anl plant dynamics code-SAS4A/SASSYS-1 liquid metal reactor code system.

    SciTech Connect

    Moisseytsev, A.; Sienicki, J. J.

    2012-05-10

    Significant progress has been made on the development of a control strategy for the supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle enabling removal of power from an autonomous load following Sodium-Cooled Fast Reactor (SFR) down to decay heat levels such that the S-CO{sub 2} cycle can be used to cool the reactor until decay heat can be removed by the normal shutdown heat removal system or a passive decay heat removal system such as Direct Reactor Auxiliary Cooling System (DRACS) loops with DRACS in-vessel heat exchangers. This capability of the new control strategy eliminates the need for use of a separate shutdown heat removal system which might also use supercritical CO{sub 2}. It has been found that this capability can be achieved by introducing a new control mechanism involving shaft speed control for the common shaft joining the turbine and two compressors following reduction of the load demand from the electrical grid to zero. Following disconnection of the generator from the electrical grid, heat is removed from the intermediate sodium circuit through the sodium-to-CO{sub 2} heat exchanger, the turbine solely drives the two compressors, and heat is rejected from the cycle through the CO{sub 2}-to-water cooler. To investigate the effectiveness of shaft speed control, calculations are carried out using the coupled Plant Dynamics Code-SAS4A/SASSYS-1 code for a linear load reduction transient for a 1000 MWt metallic-fueled SFR with autonomous load following. No deliberate motion of control rods or adjustment of sodium pump speeds is assumed to take place. It is assumed that the S-CO{sub 2} turbomachinery shaft speed linearly decreases from 100 to 20% nominal following reduction of grid load to zero. The reactor power is calculated to autonomously decrease down to 3% nominal providing a lengthy window in time for the switchover to the normal shutdown heat removal system or for a passive decay heat removal system to become effective. However, the

  12. The Fundamental Neutron Physics Beamline at the Spallation Neutron Source

    PubMed Central

    Greene, Geoffrey; Cianciolo, Vince; Koehler, Paul; Allen, Richard; Snow, William Michael; Huffman, Paul; Gould, Chris; Bowman, David; Cooper, Martin; Doyle, John

    2005-01-01

    The Spallation Neutron Source (SNS), currently under construction at Oak Ridge National Laboratory with an anticipated start-up in early 2006, will provide the most intense pulsed beams of cold neutrons in the world. At a projected power of 1.4 MW, the time averaged fluxes and fluences of the SNS will approach those of high flux reactors. One of the flight paths on the cold, coupled moderator will be devoted to fundamental neutron physics. The fundamental neutron physics beamline is anticipated to include two beam-lines; a broad band cold beam, and a monochromatic beam of 0.89 nm neutrons for ultracold neutron (UCN) experiments. The fundamental neutron physics beamline will be operated as a user facility with experiment selection based on a peer reviewed proposal process. An initial program of five experiments in neutron decay, hadronic weak interaction and time reversal symmetry violation have been proposed. PMID:27308112

  13. EXTENSION OF THE NUCLEAR REACTION MODEL CODE EMPIRE TO ACTINIDES NUCLEAR DATA EVALUATION.

    SciTech Connect

    CAPOTE,R.; SIN, M.; TRKOV, A.; HERMAN, M.; CARLSON, B.V.; OBLOZINSKY, P.

    2007-04-22

    Recent extensions and improvements of the EMPIRE code system are outlined. They add new capabilities to the code, such as prompt fission neutron spectra calculations using Hauser-Feshbach plus pre-equilibrium pre-fission spectra, cross section covariance matrix calculations by Monte Carlo method, fitting of optical model parameters, extended set of optical model potentials including new dispersive coupled channel potentials, parity-dependent level densities and transmission through numerically defined fission barriers. These features, along with improved and validated ENDF formatting, exclusive/inclusive spectra, and recoils make the current EMPIRE release a complete and well validated tool for evaluation of nuclear data at incident energies above the resonance region. The current EMPIRE release has been used in evaluations of neutron induced reaction files for {sup 232}Th and {sup 231,233}Pa nuclei in the fast neutron region at IAEA. Triple-humped fission barriers and exclusive pre-fission neutron spectra were considered for the fission data evaluation. Total, fission, capture and neutron emission cross section, average resonance parameters and angular distributions of neutron scattering are in excellent agreement with the available experimental data.

  14. Monte Carlo simulation of moderator and reflector in coal analyzer based on a D-T neutron generator.

    PubMed

    Shan, Qing; Chu, Shengnan; Jia, Wenbao

    2015-11-01

    Coal is one of the most popular fuels in the world. The use of coal not only produces carbon dioxide, but also contributes to the environmental pollution by heavy metals. In prompt gamma-ray neutron activation analysis (PGNAA)-based coal analyzer, the characteristic gamma rays of C and O are mainly induced by fast neutrons, whereas thermal neutrons can be used to induce the characteristic gamma rays of H, Si, and heavy metals. Therefore, appropriate thermal and fast neutrons are beneficial in improving the measurement accuracy of heavy metals, and ensure that the measurement accuracy of main elements meets the requirements of the industry. Once the required yield of the deuterium-tritium (d-T) neutron generator is determined, appropriate thermal and fast neutrons can be obtained by optimizing the neutron source term. In this article, the Monte Carlo N-Particle (MCNP) Transport Code and Evaluated Nuclear Data File (ENDF) database are used to optimize the neutron source term in PGNAA-based coal analyzer, including the material and shape of the moderator and neutron reflector. The optimized targets include two points: (1) the ratio of the thermal to fast neutron is 1:1 and (2) the total neutron flux from the optimized neutron source in the sample increases at least 100% when compared with the initial one. The simulation results show that, the total neutron flux in the sample increases 102%, 102%, 85%, 72%, and 62% with Pb, Bi, Nb, W, and Be reflectors, respectively. Maximum optimization of the targets is achieved when the moderator is a 3-cm-thick lead layer coupled with a 3-cm-thick high-density polyethylene (HDPE) layer, and the neutron reflector is a 27-cm-thick hemispherical lead layer. PMID:26325583

  15. ALEPH2 - A general purpose Monte Carlo depletion code

    SciTech Connect

    Stankovskiy, A.; Van Den Eynde, G.; Baeten, P.; Trakas, C.; Demy, P. M.; Villatte, L.

    2012-07-01

    The Monte-Carlo burn-up code ALEPH is being developed at SCK-CEN since 2004. A previous version of the code implemented the coupling between the Monte Carlo transport (any version of MCNP or MCNPX) and the ' deterministic' depletion code ORIGEN-2.2 but had important deficiencies in nuclear data treatment and limitations inherent to ORIGEN-2.2. A new version of the code, ALEPH2, has several unique features making it outstanding among other depletion codes. The most important feature is full data consistency between steady-state Monte Carlo and time-dependent depletion calculations. The last generation general-purpose nuclear data libraries (JEFF-3.1.1, ENDF/B-VII and JENDL-4) are fully implemented, including special purpose activation, spontaneous fission, fission product yield and radioactive decay data. The built-in depletion algorithm allows to eliminate the uncertainties associated with obtaining the time-dependent nuclide concentrations. A predictor-corrector mechanism, calculation of nuclear heating, calculation of decay heat, decay neutron sources are available as well. The validation of the code on the results of REBUS experimental program has been performed. The ALEPH2 has shown better agreement with measured data than other depletion codes. (authors)

  16. A Photo-neutron Source for a Sub-Critical Nuclear Reactor Program

    SciTech Connect

    Reda, M.A.; Harmon, J.F.; Sadineni, S.B.

    2003-08-26

    Experiments to benchmark photo-neutron production calculations for an Accelerator Driven Sub-Critical System (ADS) are described. A photo-nuclear based neutron source with output > 1013 n/sec has been proposed as a driver for a program using the sub-critical assembly at Idaho State University. The program is intended to study ADS control issues arising from coupling an accelerator neutron source with a sub-critical assembly. The experiments were performed using the 20 MeV electron linear accelerator at the Idaho Accelerator Center (IAC). Results of calculations, that were made using ACCEPT, PINP, MCNP, and MCNPX codes to optimize photo-nuclear based neutron conversion targets, are compared to experimental data for a single energy measurement.

  17. Exploration of direct neutron capture with covariant density functional theory inputs

    NASA Astrophysics Data System (ADS)

    Zhang, Shi-Sheng; Peng, Jin-Peng; Smith, M. S.; Arbanas, G.; Kozub, R. L.

    2015-04-01

    Predictions of direct neutron capture are of vital importance for simulations of nucleosynthesis in supernovae, merging neutron stars, and other astrophysical environments. We calculated direct capture cross sections using nuclear structure information obtained from a covariant density functional theory as input for the fresco coupled reaction channels code. We investigated the impact of pairing, spectroscopic factors, and optical potentials on our results to determine a robust method to calculate cross sections of direct neutron capture on exotic nuclei. Our predictions agree reasonably well with experimental cross section data for the closed shell nuclei 16O and 48Ca, and for the exotic nucleus 36S . We then used this approach to calculate the direct neutron capture cross section on the doubly magic unstable nucleus 132Sn which is of interest for the astrophysical r-process.

  18. Neutronic reactor

    DOEpatents

    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.

  19. FAA Smoke Transport Code

    SciTech Connect

    Domino, Stefan; Luketa-Hanlin, Anay; Gallegos, Carlos

    2006-10-27

    FAA Smoke Transport Code, a physics-based Computational Fluid Dynamics tool, which couples heat, mass, and momentum transfer, has been developed to provide information on smoke transport in cargo compartments with various geometries and flight conditions. The software package contains a graphical user interface for specification of geometry and boundary conditions, analysis module for solving the governing equations, and a post-processing tool. The current code was produced by making substantial improvements and additions to a code obtained from a university. The original code was able to compute steady, uniform, isothermal turbulent pressurization. In addition, a preprocessor and postprocessor were added to arrive at the current software package.

  20. Fast ignitor coupling physics

    SciTech Connect

    Mason, R.J.; Tabak, M.

    1997-10-01

    The Fast Ignitor is an alternate approach to ICF in which short pulse lasers are used to initiate burn at the surface of the compressed DT fuel. The aim is to avoid the need for careful central focusing of final shocks, and possibly to lower substantially the energy requirements for ignition. Ultimately, both goals may prove crucial to Science Based Stockpile Stewardship (SBSS). This will be the case should either emerging energetic needs, or funding difficulties render the presently planned radiative fusion approach to ignition with the NIF impractical. Ignition is a first step towards the achievement of substantial energy and neutron outputs for such Stewardship. For success with the Fast Ignitor, the laser energy must be efficiently deposited into megavolt electrons (suprathermal), which must, in turn, couple to the background ions within an alpha particle range. To understand the electron fuel coupling, we have used ANTHEM plasma simulation code to model the transport of hot electrons generated by an intense short pulse laser into plasma targets over a broad range of densities. Our study will spell out the acceleration and transport mechanisms active in the Fast Ignitor environment.

  1. Advanced multiphysics coupling for LWR fuel performance analysis

    SciTech Connect

    Hales, J. D.; Tonks, M. R.; Gleicher, F. N.; Spencer, B. W.; Novascone, S. R.; Williamson, R. L.; Pastore, G.; Perez, D. M.

    2015-10-01

    Even the most basic nuclear fuel analysis is a multiphysics undertaking, as a credible simulation must consider at a minimum coupled heat conduction and mechanical deformation. The need for more realistic fuel modeling under a variety of conditions invariably leads to a desire to include coupling between a more complete set of the physical phenomena influencing fuel behavior, including neutronics, thermal hydraulics, and mechanisms occurring at lower length scales. This paper covers current efforts toward coupled multiphysics LWR fuel modeling in three main areas. The first area covered in this paper concerns thermomechanical coupling. The interaction of these two physics, particularly related to the feedback effect associated with heat transfer and mechanical contact at the fuel/clad gap, provides numerous computational challenges. An outline is provided of an effective approach used to manage the nonlinearities associated with an evolving gap in BISON, a nuclear fuel performance application. A second type of multiphysics coupling described here is that of coupling neutronics with thermomechanical LWR fuel performance. DeCART, a high-fidelity core analysis program based on the method of characteristics, has been coupled to BISON. DeCART provides sub-pin level resolution of the multigroup neutron flux, with resonance treatment, during a depletion or a fast transient simulation. Two-way coupling between these codes was achieved by mapping fission rate density and fast neutron flux fields from DeCART to BISON and the temperature field from BISON to DeCART while employing a Picard iterative algorithm. Finally, the need for multiscale coupling is considered. Fission gas production and evolution significantly impact fuel performance by causing swelling, a reduction in the thermal conductivity, and fission gas release. The mechanisms involved occur at the atomistic and grain scale and are therefore not the domain of a fuel performance code. However, it is possible to use

  2. Advanced multiphysics coupling for LWR fuel performance analysis

    DOE PAGESBeta

    Hales, J. D.; Tonks, M. R.; Gleicher, F. N.; Spencer, B. W.; Novascone, S. R.; Williamson, R. L.; Pastore, G.; Perez, D. M.

    2015-10-01

    Even the most basic nuclear fuel analysis is a multiphysics undertaking, as a credible simulation must consider at a minimum coupled heat conduction and mechanical deformation. The need for more realistic fuel modeling under a variety of conditions invariably leads to a desire to include coupling between a more complete set of the physical phenomena influencing fuel behavior, including neutronics, thermal hydraulics, and mechanisms occurring at lower length scales. This paper covers current efforts toward coupled multiphysics LWR fuel modeling in three main areas. The first area covered in this paper concerns thermomechanical coupling. The interaction of these two physics,more » particularly related to the feedback effect associated with heat transfer and mechanical contact at the fuel/clad gap, provides numerous computational challenges. An outline is provided of an effective approach used to manage the nonlinearities associated with an evolving gap in BISON, a nuclear fuel performance application. A second type of multiphysics coupling described here is that of coupling neutronics with thermomechanical LWR fuel performance. DeCART, a high-fidelity core analysis program based on the method of characteristics, has been coupled to BISON. DeCART provides sub-pin level resolution of the multigroup neutron flux, with resonance treatment, during a depletion or a fast transient simulation. Two-way coupling between these codes was achieved by mapping fission rate density and fast neutron flux fields from DeCART to BISON and the temperature field from BISON to DeCART while employing a Picard iterative algorithm. Finally, the need for multiscale coupling is considered. Fission gas production and evolution significantly impact fuel performance by causing swelling, a reduction in the thermal conductivity, and fission gas release. The mechanisms involved occur at the atomistic and grain scale and are therefore not the domain of a fuel performance code. However, it is

  3. Interfacing MCNPX and McStas for simulation of neutron transport

    NASA Astrophysics Data System (ADS)

    Klinkby, Esben; Lauritzen, Bent; Nonbøl, Erik; Kjær Willendrup, Peter; Filges, Uwe; Wohlmuther, Michael; Gallmeier, Franz X.

    2013-02-01

    Simulations of target-moderator-reflector system at spallation sources are conventionally carried out using Monte Carlo codes such as MCNPX (Waters et al., 2007 [1]) or FLUKA (Battistoni et al., 2007; Ferrari et al., 2005 [2,3]) whereas simulations of neutron transport from the moderator and the instrument response are performed by neutron ray tracing codes such as McStas (Lefmann and Nielsen, 1999; Willendrup et al., 2004, 2011a,b [4-7]). The coupling between the two simulation suites typically consists of providing analytical fits of MCNPX neutron spectra to McStas. This method is generally successful but has limitations, as it e.g. does not allow for re-entry of neutrons into the MCNPX regime. Previous work to resolve such shortcomings includes the introduction of McStas inspired supermirrors in MCNPX. In the present paper different approaches to interface MCNPX and McStas are presented and applied to a simple test case. The direct coupling between MCNPX and McStas allows for more accurate simulations of e.g. complex moderator geometries, backgrounds, interference between beam-lines as well as shielding requirements along the neutron guides.

  4. The National Ignition Facility neutron imaging system.

    PubMed

    Wilke, Mark D; Batha, Steven H; Bradley, Paul A; Day, Robert D; Clark, David D; Fatherley, Valerie E; Finch, Joshua P; Gallegos, Robert A; Garcia, Felix P; Grim, Gary P; Jaramillo, Steven A; Montoya, Andrew J; Moran, Michael J; Morgan, George L; Oertel, John A; Ortiz, Thomas A; Payton, Jeremy R; Pazuchanics, Peter; Schmidt, Derek W; Valdez, Adelaida C; Wilde, Carl H; Wilson, Doug C

    2008-10-01

    The National Ignition Facility (NIF) is scheduled to begin deuterium-tritium (DT) shots possibly in the next several years. One of the important diagnostics in understanding capsule behavior and to guide changes in Hohlraum illumination, capsule design, and geometry will be neutron imaging of both the primary 14 MeV neutrons and the lower-energy downscattered neutrons in the 6-13 MeV range. The neutron imaging system (NIS) described here, which we are currently building for use on NIF, uses a precisely aligned set of apertures near the target to form the neutron images on a segmented scintillator. The images are recorded on a gated, intensified charge coupled device. Although the aperture set may be as close as 20 cm to the target, the imaging camera system will be located at a distance of 28 m from the target. At 28 m the camera system is outside the NIF building. Because of the distance and shielding, the imager will be able to obtain images with little background noise. The imager will be capable of imaging downscattered neutrons from failed capsules with yields Y(n)>10(14) neutrons. The shielding will also permit the NIS to function at neutron yields >10(18), which is in contrast to most other diagnostics that may not work at high neutron yields. The following describes the current NIF NIS design and compares the predicted performance with the NIF specifications that must be satisfied to generate images that can be interpreted to understand results of a particular shot. The current design, including the aperture, scintillator, camera system, and reconstruction methods, is briefly described. System modeling of the existing Omega NIS and comparison with the Omega data that guided the NIF design based on our Omega results is described. We will show NIS model calculations of the expected NIF images based on component evaluations at Omega. We will also compare the calculated NIF input images with those unfolded from the NIS images generated from our NIS numerical

  5. Computational radiology and imaging with the MCNP Monte Carlo code

    SciTech Connect

    Estes, G.P.; Taylor, W.M.

    1995-05-01

    MCNP, a 3D coupled neutron/photon/electron Monte Carlo radiation transport code, is currently used in medical applications such as cancer radiation treatment planning, interpretation of diagnostic radiation images, and treatment beam optimization. This paper will discuss MCNP`s current uses and capabilities, as well as envisioned improvements that would further enhance MCNP role in computational medicine. It will be demonstrated that the methodology exists to simulate medical images (e.g. SPECT). Techniques will be discussed that would enable the construction of 3D computational geometry models of individual patients for use in patient-specific studies that would improve the quality of care for patients.

  6. NEUTRONIC REACTOR CONTROL

    DOEpatents

    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.

  7. Clinical coding. Code breakers.

    PubMed

    Mathieson, Steve

    2005-02-24

    --The advent of payment by results has seen the role of the clinical coder pushed to the fore in England. --Examinations for a clinical coding qualification began in 1999. In 2004, approximately 200 people took the qualification. --Trusts are attracting people to the role by offering training from scratch or through modern apprenticeships. PMID:15768716

  8. Recent development and applications of the MORSE Code

    SciTech Connect

    Cramer, S.N.

    1993-06-01

    Several recent analyses using the multigroup MORSE Monte Carlo code are presented. In the calculation of a highly directional-dependent neutron streaming experiment it is shown that P{sub 7} cross section representation produces results virtually identical with those from an analog code. Use has been made here of a recently released ENDF/B-VI data set. In the analysis of neutron distributions inside the water-cooled ORELA accelerator target and positron source, an analytic hydrogen scattering model is incorporated into the otherwise multigroup treatment. The radiation from a nuclear weapon is analyzed in a large concrete building in Nagasaki by coupling MORSE and the DOT discrete ordinates code. The spatial variation of the DOT-generated free-field radiation is utilized, and the building is modeled with the array feature of the MORSE geometry package. An analytic directional biasing, applicable to the discrete scattering angle procedure in MORSE, is combined with the exponential transform. As in more general studies, it is shown that the combined biasing is more efficient than either biasing used separately. Other tracking improvements are included in a difficult streaming and penetration radiation analysis through a concrete structure. Proposals are given for the code generation of the required biasing parameters.

  9. Neutronics activities for next generation devices

    SciTech Connect

    Gohar, Y.

    1985-01-01

    Neutronic activities for the next generation devices are the subject of this paper. The main activities include TFCX and FPD blanket/shield studies, neutronic aspects of ETR/INTOR critical issues, and neutronics computational modules for the tokamak system code and tandem mirror reactor system code. Trade-off analyses, optimization studies, design problem investigations and computational models development for reactor parametric studies carried out for these activities are summarized.

  10. Accurate determination of chlorine, bromine, and iodine in sedimentary rock reference samples by radiochemical neutron activation analysis and a detailed comparison with inductively coupled plasma mass spectrometry literature data.

    PubMed

    Sekimoto, Shun; Ebihara, Mitsuru

    2013-07-01

    Trace amounts of three halogens (chlorine, bromine, and iodine) were determined using radiochemical neutron activation analysis (RNAA) for nine sedimentary rocks and three rhyolite samples. To obtain high-quality analytical data, the radiochemical procedure of RNAA was improved by lowering the background in gamma-ray spectrometry and completing the chemical procedure more rapidly than in conventional procedures. A comparison of the RNAA data of Br and I with corresponding inductively coupled plasma mass spectrometry (ICPMS) literature data revealed that the values obtained by ICPMS coupled with pyrohydrolysis preconcentration were systematically lower than the RNAA data for some reference samples, suggesting that the quantitative collection of Br and I cannot always be achieved by the pyrohydrolysis for some solid samples. The RNAA data of three halogens can classify sedimentary rock reference samples into two groups (the samples from inland water and those from seawater), implying the geochemical significance of halogen data. PMID:23710630

  11. Multidimensional optimization of fusion reactors using heterogenous codes and engineering software

    NASA Astrophysics Data System (ADS)

    Hartwig, Zachary; Olynyk, Geoffrey; Whyte, Dennis

    2012-10-01

    Magnetic confinement fusion reactors are tightly coupled systems. The parameters under a designer's control, such as magnetic field, wall temperature, and blanket thickness, simultaneously affect the behavior, performance, and components of the reactor, leading to complex tradeoffs and design optimizations. In addition, the engineering analyses require non-trivial, self-consistent inputs, such as reactor geometry, to ensure high fidelity between the various physics and engineering design codes. We present a framework for analysis and multidimensional optimization of fusion reactor systems based on the coupling of heterogeneous codes and engineering software. While this approach is widely used in industry, most code-coupling efforts in fusion have been focused on plasma and edge physics. Instead, we use a simplified plasma model to concentrate on how fusion neutrons and heat transfer affect the design of the first wall, breeding blanket, and magnet systems. The framework combines solid modeling, neutronics, and engineering multiphysics codes and software, linked across Windows and Linux clusters. Initial results for optimizing the design of a compact, high-field tokamak reactor based on high-temperature demountable superconducting coils and a liquid blanket are presented.

  12. Advanced Neutron Source Cross Section Libraries (ANSL-V): ENDF/B-V based multigroup cross-section libraries for advanced neutron source (ANS) reactor studies

    SciTech Connect

    Ford, W.E. III; Arwood, J.W.; Greene, N.M.; Moses, D.L.; Petrie, L.M.; Primm, R.T. III; Slater, C.O.; Westfall, R.M.; Wright, R.Q.

    1990-09-01

    Pseudo-problem-independent, multigroup cross-section libraries were generated to support Advanced Neutron Source (ANS) Reactor design studies. The ANS is a proposed reactor which would be fueled with highly enriched uranium and cooled with heavy water. The libraries, designated ANSL-V (Advanced Neutron Source Cross Section Libraries based on ENDF/B-V), are data bases in AMPX master format for subsequent generation of problem-dependent cross-sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, DORT, TORT, and MORSE. Included in ANSL-V are 99-group and 39-group neutron, 39-neutron-group 44-gamma-ray-group secondary gamma-ray production (SGRP), 44-group gamma-ray interaction (GRI), and coupled, 39-neutron group 44-gamma-ray group (CNG) cross-section libraries. The neutron and SGRP libraries were generated primarily from ENDF/B-V data; the GRI library was generated from DLC-99/HUGO data, which is recognized as the ENDF/B-V photon interaction data. Modules from the AMPX and NJOY systems were used to process the multigroup data. Validity of selected data from the fine- and broad-group neutron libraries was satisfactorily tested in performance parameter calculations.

  13. A Search for Exotic Spin-Dependent Interactions of the Neutron using Neutron Spin Rotation

    NASA Astrophysics Data System (ADS)

    Haddock, Chris; Nsr Collaboration

    2016-03-01

    Many theories beyond the Standard Model lead at low energy to spin-dependent, weakly-coupled interactions of mesoscopic range. Laboratory constraints on such interactions are quite poor. We describe an experiment in progress at the LANSCE spallation neutron source at Los Alamos to search for exotic axial couplings of neutrons to matter from light vector boson exchange. The experiment makes use of a slow neutron polarimeter and a target with an oscillating mass density. Neutron Spin Rotation.

  14. Applications guide to the MORSE Monte Carlo code

    SciTech Connect

    Cramer, S.N.

    1985-08-01

    A practical guide for the implementation of the MORESE-CG Monte Carlo radiation transport computer code system is presented. The various versions of the MORSE code are compared and contrasted, and the many references dealing explicitly with the MORSE-CG code are reviewed. The treatment of angular scattering is discussed, and procedures for obtaining increased differentiality of results in terms of reaction types and nuclides from a multigroup Monte Carlo code are explained in terms of cross-section and geometry data manipulation. Examples of standard cross-section data input and output are shown. Many other features of the code system are also reviewed, including (1) the concept of primary and secondary particles, (2) fission neutron generation, (3) albedo data capability, (4) DOMINO coupling, (5) history file use for post-processing of results, (6) adjoint mode operation, (7) variance reduction, and (8) input/output. In addition, examples of the combinatorial geometry are given, and the new array of arrays geometry feature (MARS) and its three-dimensional plotting code (JUNEBUG) are presented. Realistic examples of user routines for source, estimation, path-length stretching, and cross-section data manipulation are given. A deatiled explanation of the coupling between the random walk and estimation procedure is given in terms of both code parameters and physical analogies. The operation of the code in the adjoint mode is covered extensively. The basic concepts of adjoint theory and dimensionality are discussed and examples of adjoint source and estimator user routines are given for all common situations. Adjoint source normalization is explained, a few sample problems are given, and the concept of obtaining forward differential results from adjoint calculations is covered. Finally, the documentation of the standard MORSE-CG sample problem package is reviewed and on-going and future work is discussed.

  15. Potential for improvement of a neutron producing target for time-of-flight measurements

    NASA Astrophysics Data System (ADS)

    Flaska, M.; Lathouwers, D.; Plompen, A. J. M.; Mondelaers, W.; van der Hagen, T. H. J. J.; van Dam, H.

    2005-12-01

    The Geel electron linear accelerator (GELINA) white neutron source is designed for time-of-flight (TOF) measurements with a high neutron energy resolution. A project has been launched in order to improve the accuracy of the high-resolution neutron cross-section measurements. The experimental accuracy is highly dependent on the neutron intensity and the time spread of the neutrons leaving the target. The main project objective is not only to design a new neutron producing target that reduces the time spread of the neutrons of a given energy, but also to maintain or possibly enhance the neutron production. The present GELINA target has been simulated recently with coupled electron-photon-neutron MCNP4C3 calculations and, subsequently, compared with experimental results. Based on the very good benchmarking results, this code is now used to design and optimise the new target and to assess all relevant parameters influencing the neutron flux and resolution functions. In this paper, the reader will find an explanation of the methodology on which the research of a new target is based. We compared the neutron fluxes of the compact geometries of various materials for different angles with the fluxes of the existing target. Further, we also made a comparison of the resolution functions for the present design versus a compact design. Finally, a dedicated figure of merit was used to make a qualitative comparison of various designs. The results presented in this paper reveal that there is a possibility to design a new target with superior characteristics so that all project objectives will be fulfilled.

  16. A Monte Carlo model system for core analysis and epithermal neutron beam design at the Washington State University Radiation Center

    SciTech Connect

    Burns, T.D. Jr.

    1996-05-01

    The Monte Carlo Model System (MCMS) for the Washington State University (WSU) Radiation Center provides a means through which core criticality and power distributions can be calculated, as well as providing a method for neutron and photon transport necessary for BNCT epithermal neutron beam design. The computational code used in this Model System is MCNP4A. The geometric capability of this Monte Carlo code allows the WSU system to be modeled very accurately. A working knowledge of the MCNP4A neutron transport code increases the flexibility of the Model System and is recommended, however, the eigenvalue/power density problems can be run with little direct knowledge of MCNP4A. Neutron and photon particle transport require more experience with the MCNP4A code. The Model System consists of two coupled subsystems; the Core Analysis and Source Plane Generator Model (CASP), and the BeamPort Shell Particle Transport Model (BSPT). The CASP Model incorporates the S({alpha}, {beta}) thermal treatment, and is run as a criticality problem yielding, the system eigenvalue (k{sub eff}), the core power distribution, and an implicit surface source for subsequent particle transport in the BSPT Model. The BSPT Model uses the source plane generated by a CASP run to transport particles through the thermal column beamport. The user can create filter arrangements in the beamport and then calculate characteristics necessary for assessing the BNCT potential of the given filter want. Examples of the characteristics to be calculated are: neutron fluxes, neutron currents, fast neutron KERMAs and gamma KERMAs. The MCMS is a useful tool for the WSU system. Those unfamiliar with the MCNP4A code can use the MCMS transparently for core analysis, while more experienced users will find the particle transport capabilities very powerful for BNCT filter design.

  17. Development of a dynamic simulation mode in Serpent 2 Monte Carlo code

    SciTech Connect

    Leppaenen, J.

    2013-07-01

    This paper presents a dynamic neutron transport mode, currently being implemented in the Serpent 2 Monte Carlo code for the purpose of simulating short reactivity transients with temperature feedback. The transport routine is introduced and validated by comparison to MCNP5 calculations. The method is also tested in combination with an internal temperature feedback module, which forms the inner part of a multi-physics coupling scheme in Serpent 2. The demo case for the coupled calculation is a reactivity-initiated accident (RIA) in PWR fuel. (authors)

  18. Shifting scintillator neutron detector

    SciTech Connect

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

    2014-03-04

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

  19. The FLUKA Code: an Overview

    SciTech Connect

    Ballarini, F.; Battistoni, G.; Campanella, M.; Carboni, M.; Cerutti, F.; Empl, A.; Fasso, A.; Ferrari, A.; Gadioli, E.; Garzelli, M.V.; Lantz, M.; Liotta, M.; Mairani, A.; Mostacci, A.; Muraro, S.; Ottolenghi, A.; Pelliccioni, M.; Pinsky, L.; Ranft, J.; Roesler, S.; Sala, P.R.; /Milan U. /INFN, Milan /Pavia U. /INFN, Pavia /CERN /Siegen U. /Houston U. /SLAC /Frascati /NASA, Houston /ENEA, Frascati

    2005-11-09

    FLUKA is a multipurpose Monte Carlo code which can transport a variety of particles over a wide energy range in complex geometries. The code is a joint project of INFN and CERN: part of its development is also supported by the University of Houston and NASA. FLUKA is successfully applied in several fields, including but not only, particle physics, cosmic ray physics, dosimetry, radioprotection, hadron therapy, space radiation, accelerator design and neutronics. The code is the standard tool used at CERN for dosimetry, radioprotection and beam-machine interaction studies. Here we give a glimpse into the code physics models with a particular emphasis to the hadronic and nuclear sector.

  20. The FLUKA Code: An Overview

    NASA Technical Reports Server (NTRS)

    Ballarini, F.; Battistoni, G.; Campanella, M.; Carboni, M.; Cerutti, F.; Empl, A.; Fasso, A.; Ferrari, A.; Gadioli, E.; Garzelli, M. V.; Lantz, M.; Liotta, M.; Mairani, A.; Mostacci, A.; Muraro, S.; Ottolenghi, A.; Pelliccioni, M.; Pinsky, L.; Ranft, J.; Roesler, S.; Sala, P. R.; Scannicchio, D.; Trovati, S.; Villari, R.; Wilson, T.

    2006-01-01

    FLUKA is a multipurpose Monte Carlo code which can transport a variety of particles over a wide energy range in complex geometries. The code is a joint project of INFN and CERN: part of its development is also supported by the University of Houston and NASA. FLUKA is successfully applied in several fields, including but not only, particle physics, cosmic ray physics, dosimetry, radioprotection, hadron therapy, space radiation, accelerator design and neutronics. The code is the standard tool used at CERN for dosimetry, radioprotection and beam-machine interaction studies. Here we give a glimpse into the code physics models with a particular emphasis to the hadronic and nuclear sector.

  1. The New 1999 National Electrical Code Coupled with New Standards Clarify Requirements for Installations of Photovoltaic Systems in the U.S.

    SciTech Connect

    Bower, W.

    1999-01-08

    The National Electrical Code@ (NEC@) focuses primarily on electrical system installation requirements in the U.S. The NEC addresses both fire and personnel safety. This paper will describe recent efforts of the PV industry in the U.S. and the resulting requirements in the 1999 National Electrical Code-- Article 690 --Solar Photovoltaic Systems. The Article 690 requirements spell out the PV-unique requirements for safe installations of PV systems in the U.S.A. This paper provides an overview of the most significant changes that appear in Article 690 of the 1999 edition of the NEC. The related and coordinated efforts of the other standards- making groups will also be briefly reviewed.

  2. Boltzmann-Fokker-Planck calculations using standard discrete-ordinates codes

    SciTech Connect

    Morel, J.E.

    1987-01-01

    The Boltzmann-Fokker-Planck (BFP) equation can be used to describe both neutral and charged-particle transport. Over the past several years, the author and several collaborators have developed methods for representing Fokker-Planck operators with standard multigroup-Legendre cross-section data. When these data are input to a standard S/sub n/ code such as ONETRAN, the code actually solves the Boltzmann-Fokker-Planck equation rather than the Boltzmann equation. This is achieved wihout any modification to the S/sub n/ codes. Because BFP calculations can be more demanding from a numerical viewpoint than standard neutronics calculations, we have found it useful to implement new quadrature methods ad convergence acceleration methods in the standard discrete-ordinates code, ONETRAN. We discuss our BFP cross-section representation techniques, our improved quadrature and acceleration techniques, and present results from BFP coupled electron-photon transport calculations performed with ONETRAN. 19 refs., 7 figs.

  3. Evaluated Nuclear Data Library for Transport Calculations Involving Incident Neutrons and Protons of Energy Up to 100 MeV.

    Energy Science and Technology Software Center (ESTSC)

    1993-08-09

    Version 00 This data base was developed for use in Monte Carlo or discrete ordinate transport codes, for example, the general Monte Carlo code MCNP. Various modules of the NJOY processing code system have been enhanced to permit processing of the ENDF/B-VI formatted evaluations into both continuous-energy and multi-group format. The transport data files for all 18 projectile-plus-target systems have been processed through NJOY, and coupled multi-particle, multi-group transport libraries for MCNP now exist. Inmore » addition, pointwise MCNP libraries to 100 MeV for incident neutrons have been prepared for the nine targets. The production version of the MCNP code is being modified to handle the new pointwise libraries. The production version of MCNP already supports the use of coupled multi-group libraries.« less

  4. Neutron-image intensifier

    NASA Technical Reports Server (NTRS)

    Berger, H.

    1970-01-01

    Electronic intensifier tube with a demagnification ratio of 9-1 enhances the usefulness of neutron-radiographic techniques. A television signal can be obtained by optical coupling of a small-output phosphor-light image to a television camera.

  5. Neutron energy spectrum adjustment using deposited metal films on Teflon in the miniature neutron source reactor.

    PubMed

    Nassan, L; Abdallah, B; Omar, H; Sarheel, A; Alsomel, N; Ghazi, N

    2016-01-01

    The focus of this article was on the experimental estimation of the neutron energy spectrum in the inner irradiation site of the miniature neutron source reactor (MNSR), using, for the first time, a selected set of deposited metal films on Teflon (DMFTs) neutron detectors. Gold, copper, zinc, titanium, aluminum, nickel, silver, and chromium were selected because of the dependence of their neutron cross-sections on neutron energy. Emphasis was placed on the usability of this new type of neutron detectors in the total neutron energy spectrum adjustment. The measured saturation activities per target nucleus values of the DMFTs, and the calculated neutron spectrum in the inner irradiation site using the MCNP-4C code were used as an input for the STAY'SL computer code during the adjustment procedure. The agreement between the numerically calculated and experimentally adjusted spectra results was discussed. PMID:26562448

  6. Artificial Neural Networks in Spectrometry and Neutron Dosimetry

    SciTech Connect

    Vega-Carrillo, H. R.; Martinez-Blanco, M. R.; Ortiz-Rodriguez, J. M.; Hernandez-Davila, V. M.

    2010-12-07

    The ANN technology has been applied to unfold the neutron spectra of three neutron sources and to estimate their dosimetric features. To compare these results, neutron spectra were also unfolded with the BUNKIUT code. Both unfolding procedures were carried out using the count rates of a Bonner sphere spectrometer. The spectra unfolded with ANN result similar to those unfolded with the BUNKIUT code. The H*(10) values obtained with ANN agrees well with H*(10) values calculated with the BUNKIUT code.

  7. Slow neutron leakage spectra from spallation neutron sources

    SciTech Connect

    Das, S.G.; Carpenter, J.M.; Prael, R.E.

    1980-02-01

    An efficient technique is described for Monte Carlo simulation of neutron beam spectra from target-moderator-reflector assemblies typical of pulsed spallation neutron sources. The technique involves the scoring of the transport-theoretical probability that a neutron will emerge from the moderator surface in the direction of interest, at each collision. An angle-biasing probability is also introduced which further enhances efficiency in simple problems. These modifications were introduced into the VIM low energy neutron transport code, representing the spatial and energy distributions of the source neutrons approximately as those of evaporation neutrons generated through the spallation process by protons of various energies. The intensity of slow neutrons leaking from various reflected moderators was studied for various neutron source arrangements. These include computations relating to early measurements on a mockup-assembly, a brief survey of moderator materials and sizes, and a survey of the effects of varying source and moderator configurations with a practical, liquid metal cooled uranium source Wing and slab, i.e., tangential and radial moderator arrangements, and Be vs CH/sub 2/ reflectors are compared. Results are also presented for several complicated geometries which more closely represent realistic arrangements for a practical source, and for a subcritical fission multiplier such as might be driven by an electron linac. An adaptation of the code was developed to enable time dependent calculations, and investigated the effects of the reflector, decoupling and void liner materials on the pulse shape.

  8. Combined neutron imaging techniques for cultural heritage purpose

    SciTech Connect

    Materna, T.

    2009-01-28

    This article presents the different new neutron techniques developed by the Ancient Charm collaboration to image objects of cultural heritage importance: Prompt-gamma-ray activation imaging (PGAI) coupled to cold/thermal neutron transmission tomography, Neutron Resonance Capture Imaging (NRCI) and Neutron Resonance Tomography.

  9. Nucleon-nucleon momentum-correlation function as a probe of the density distribution of valence neutrons in neutron-rich nuclei

    NASA Astrophysics Data System (ADS)

    Cao, X. G.; Cai, X. Z.; Ma, Y. G.; Fang, D. Q.; Zhang, G. Q.; Guo, W.; Chen, J. G.; Wang, J. S.

    2012-10-01

    Proton-neutron, neutron-neutron, and proton-proton momentum-correlation functions (Cpn,Cnn, and Cpp) are systematically investigated for 15C and other C-isotope-induced collisions at different entrance channel conditions within the framework of the isospin-dependent quantum-molecular-dynamics model complemented by the correlation after burner (crab) computation code. 15C is a prime exotic nucleus candidate due to the weakly bound valence neutron coupling with closed-neutron-shell nucleus 14C. To study density dependence of the correlation function by removing the isospin effect, the initialized 15C projectiles are sampled from two kinds of density distribution from the relativistic mean-field (RMF) model in which the valence neutron of 15C is populated in both 1d5/2 and 2s1/2 states, respectively. The results show that the density distributions of the valence neutron significantly influence the nucleon-nucleon momentum-correlation function at large impact parameters and high incident energies. The extended density distribution of the valence neutron largely weakens the strength of the correlation function. The size of the emission source is extracted by fitting the correlation function by using the Gaussian source method. The emission source size as well as the size of the final-state phase space are larger for projectile samplings from more extended density distributions of the valence neutron, which corresponds to the 2s1/2 state in the RMF model. Therefore, the nucleon-nucleon momentum-correlation function can be considered as a potentially valuable tool to diagnose exotic nuclear structures, such as the skin and halo.

  10. Neutron/gamma dose separation by the multiple-ion-chamber technique

    SciTech Connect

    Goetsch, S.J.

    1983-01-01

    Many mixed n/..gamma.. dosimetry systems rely on two dosimeters, one composed of a tissue-equivalent material and the other made from a non-hydrogenous material. The paired chamber technique works well in fields of neutron radiation nearly identical in spectral composition to that in which the dosimeters were calibrated. However, this technique is drastically compromised in phantom due to the degradation of the neutron spectrum. The three-dosimeter technique allows for the fall-off in neutron sensitivity of the two non-hydrogenous dosimeters. Precise and physically meaningful results were obtained with this technique with a D-T source in air and in phantom and with simultaneous D-T neutron and /sup 60/Co gamma ray irradiation in air. The MORSE-CG coupled n/..gamma.. three-dimensional Monte Carlo code was employed to calculate neutron and gamma doses in a water phantom. Gamma doses calculated in phantom with this code were generally lower than corresponding ion chamber measurements. This can be explained by the departure of irradiation conditions from ideal narrow-beam geometry. 97 references.

  11. Verification of ARES transport code system with TAKEDA benchmarks

    NASA Astrophysics Data System (ADS)

    Zhang, Liang; Zhang, Bin; Zhang, Penghe; Chen, Mengteng; Zhao, Jingchang; Zhang, Shun; Chen, Yixue

    2015-10-01

    Neutron transport modeling and simulation are central to many areas of nuclear technology, including reactor core analysis, radiation shielding and radiation detection. In this paper the series of TAKEDA benchmarks are modeled to verify the critical calculation capability of ARES, a discrete ordinates neutral particle transport code system. SALOME platform is coupled with ARES to provide geometry modeling and mesh generation function. The Koch-Baker-Alcouffe parallel sweep algorithm is applied to accelerate the traditional transport calculation process. The results show that the eigenvalues calculated by ARES are in excellent agreement with the reference values presented in NEACRP-L-330, with a difference less than 30 pcm except for the first case of model 3. Additionally, ARES provides accurate fluxes distribution compared to reference values, with a deviation less than 2% for region-averaged fluxes in all cases. All of these confirms the feasibility of ARES-SALOME coupling and demonstrate that ARES has a good performance in critical calculation.

  12. Speech coding

    NASA Astrophysics Data System (ADS)

    Gersho, Allen

    1990-05-01

    Recent advances in algorithms and techniques for speech coding now permit high quality voice reproduction at remarkably low bit rates. The advent of powerful single-ship signal processors has made it cost effective to implement these new and sophisticated speech coding algorithms for many important applications in voice communication and storage. Some of the main ideas underlying the algorithms of major interest today are reviewed. The concept of removing redundancy by linear prediction is reviewed, first in the context of predictive quantization or DPCM. Then linear predictive coding, adaptive predictive coding, and vector quantization are discussed. The concepts of excitation coding via analysis-by-synthesis, vector sum excitation codebooks, and adaptive postfiltering are explained. The main idea of vector excitation coding (VXC) or code excited linear prediction (CELP) are presented. Finally low-delay VXC coding and phonetic segmentation for VXC are described.

  13. [Identification of hashish samples with inductively coupled high-frequency plasma emission spectrometry and neutron activation analysis and data handling with neuronal networks. 1. Methods for the quantitative determination of characteristic trace elements].

    PubMed

    Lahl, H; Henke, G

    1997-11-01

    Neutron activation analysis (NAA) and inductively coupled plasma emission spectrometry (ICP-AES) were used to quantify the relative contents of Fe, Sc, Ce, Pa, Cr, Co, respectively the absolute contents of Cr, Zn, Mn, Fe, Mg, Al, Cu, Ti, Ca, Sr in hashish samples, seized in different countries. The samples were processed after dry ashing by means of instrumental NAA and after wet mineralization by means of ICP-AES. For determination of the sampling and measurement errors, one of the samples was analyzed repeatedly with both methods. Classifying hashish samples with regard to concentration of certain elements could be done by artificial neural networks with a modified backpropagation algorithm. By this way, identity and non identity of one unknown sample with one of many different samples as data pool can be ascertained, on principle. PMID:9446107

  14. Neutronics design

    SciTech Connect

    Moir, R.

    1984-10-01

    Initial scoping calculations were done by Lee at LLNL with the TART code and ENDL data to determine the tritium breeding potential of this blanket type. A radially zoned cylindrical nucleonics model was used and is described. Results, local (100% blanket coverage) T and M vs Be zone thickness, are shown. The tritium breeding ratio, T, is seen to vary between 0.5 with no Be to 1.7 with a 60-cm Be zone. Correspondingly, energy multiplication, M, varies between 1.1 and 1.4. The effects of less than 100% blanket coverage on T is shown. For example, if the effective coverage is only 80, a 15-cm Be zone is needed for T = 1.01 compared to 10 cm at full coverage. Higher T can be achieved, of course, by increasing the Be zone thickness. Another possibly attractive use of the excess neutrons generated in Be is for higher M. While this was not the objective here it is clearly possible to include material in the blanket with significantly higher Q's than 4.8 MeV for the Li6(n,t) reaction. Also enriching the Li in Li6 can increase T.

  15. Atmospheric neutrons

    NASA Technical Reports Server (NTRS)

    Korff, S. A.; Mendell, R. B.; Merker, M.; Light, E. S.; Verschell, H. J.; Sandie, W. S.

    1979-01-01

    Contributions to fast neutron measurements in the atmosphere are outlined. The results of a calculation to determine the production, distribution and final disappearance of atmospheric neutrons over the entire spectrum are presented. An attempt is made to answer questions that relate to processes such as neutron escape from the atmosphere and C-14 production. In addition, since variations of secondary neutrons can be related to variations in the primary radiation, comment on the modulation of both radiation components is made.

  16. An Improved Neutron Transport Algorithm for Space Radiation

    NASA Technical Reports Server (NTRS)

    Heinbockel, John H.; Clowdsley, Martha S.; Wilson, John W.

    2000-01-01

    A low-energy neutron transport algorithm for use in space radiation protection is developed. The algorithm is based upon a multigroup analysis of the straight-ahead Boltzmann equation by using a mean value theorem for integrals. This analysis is accomplished by solving a realistic but simplified neutron transport test problem. The test problem is analyzed by using numerical and analytical procedures to obtain an accurate solution within specified error bounds. Results from the test problem are then used for determining mean values associated with rescattering terms that are associated with a multigroup solution of the straight-ahead Boltzmann equation. The algorithm is then coupled to the Langley HZETRN code through the evaporation source term. Evaluation of the neutron fluence generated by the solar particle event of February 23, 1956, for a water and an aluminum-water shield-target configuration is then compared with LAHET and MCNPX Monte Carlo code calculations for the same shield-target configuration. The algorithm developed showed a great improvement in results over the unmodified HZETRN solution. In addition, a two-directional solution of the evaporation source showed even further improvement of the fluence near the front of the water target where diffusion from the front surface is important.

  17. Uplink Coding

    NASA Technical Reports Server (NTRS)

    Pollara, Fabrizio; Hamkins, Jon; Dolinar, Sam; Andrews, Ken; Divsalar, Dariush

    2006-01-01

    This viewgraph presentation reviews uplink coding. The purpose and goals of the briefing are (1) Show a plan for using uplink coding and describe benefits (2) Define possible solutions and their applicability to different types of uplink, including emergency uplink (3) Concur with our conclusions so we can embark on a plan to use proposed uplink system (4) Identify the need for the development of appropriate technology and infusion in the DSN (5) Gain advocacy to implement uplink coding in flight projects Action Item EMB04-1-14 -- Show a plan for using uplink coding, including showing where it is useful or not (include discussion of emergency uplink coding).

  18. Neutron guide

    DOEpatents

    Greene, Geoffrey L.

    1999-01-01

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

  19. Neutron dosimetry

    DOEpatents

    Quinby, Thomas C.

    1976-07-27

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

  20. Dual-mode operation of a neutron source, a concept

    NASA Technical Reports Server (NTRS)

    Givens, W. W.; Mills, W. R., Jr.

    1969-01-01

    Pulsed neutron source operates in conjunction with a photomultiplier tube coupled to a gamma ray scintillation crystal. This allows measurements of gamma radiation from both inelastic scattering and thermal neutron capture in a single experiment.

  1. Coupled simulation of the reactor core using CUPID/MASTER

    SciTech Connect

    Lee, J. R.; Cho, H. K.; Yoon, H. Y.; Jeong, J. J.

    2012-07-01

    The CUPID is a component-scale thermal hydraulics code which is aimed for the analysis of transient two-phase flows in nuclear reactor components such as the reactor vessel, steam generator, containment. This code adopts a three-dimensional, transient, two-phase and three-field model, and includes physical models and correlations of the interfacial mass, momentum, and energy transfer for the closure. In the present paper, a multi-physics simulation was performed by coupling CUPID with a three dimensional neutron kinetics code, MASTER. MASTER is merged into CUPID as a dynamic link library (DLL). The APR1400 reactor core during a control rod drop/ejection accident was simulated as an example by adopting a porous media approach to employ a fuel assembly. The following sections present the numerical modeling for the reactor core, coupling of the kinetics code, and the simulation results. And also, a preliminary study for multi-scale simulation between CUPID and system-scaled thermal hydraulics code, MARS will be introduced as well. (authors)

  2. Accelerator shield design of KIPT neutron source facility

    SciTech Connect

    Zhong, Z.; Gohar, Y.

    2013-07-01

    Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the design development of a neutron source facility at KIPT utilizing an electron-accelerator-driven subcritical assembly. Electron beam power is 100 kW, using 100 MeV electrons. The facility is designed to perform basic and applied nuclear research, produce medical isotopes, and train young nuclear specialists. The biological shield of the accelerator building is designed to reduce the biological dose to less than 0.5-mrem/hr during operation. The main source of the biological dose is the photons and the neutrons generated by interactions of leaked electrons from the electron gun and accelerator sections with the surrounding concrete and accelerator materials. The Monte Carlo code MCNPX serves as the calculation tool for the shield design, due to its capability to transport electrons, photons, and neutrons coupled problems. The direct photon dose can be tallied by 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 less than 0.01 neutron per electron. This causes difficulties for Monte Carlo analyses and consumes tremendous computation time for tallying with acceptable statistics the neutron dose outside the shield boundary. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were developed for the study. The generated neutrons are banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron and secondary photon doses. The weight windows variance reduction technique is utilized for both neutron and photon dose calculations. Two shielding materials, i.e., heavy concrete and ordinary concrete, were considered for the shield design. The main goal is to maintain the total

  3. European Neutron Activation System.

    Energy Science and Technology Software Center (ESTSC)

    2013-01-11

    Version 03 EASY-2010 (European Activation System) consists of a wide range of codes, data and documentation all aimed at satisfying the objective of calculating the response of materials irradiated in a neutron flux. The main difference from the previous version is the upper energy limit, which has increased from 20 to 60 MeV. It is designed to investigate both fusion devices and accelerator based materials test facilities that will act as intense sources of high-energymore » neutrons causing significant activation of the surrounding materials. The very general nature of the calculational method and the data libraries means that it is applicable (with some reservations) to all situations (e.g. fission reactors or neutron sources) where materials are exposed to neutrons below 60 MeV. EASY can be divided into two parts: data and code development tools and user tools and data. The former are required to develop the latter, but EASY users only need to be able to use the inventory code FISPACT and be aware of the contents of the EAF library (the data source). The complete EASY package contains the FISPACT-2007 inventory code, the EAF-2003, EAF-2005, EAF-2007 and EAF-2010 libraries, and the EASY User Interface for the Window version. The activation package EASY-2010 is the result of significant development to extend the upper energy range from 20 to 60 MeV so that it is capable of being used for IFMIF calculations. The EAF-2010 library contains 66,256 reactions, almost five times more than in EAF-2003 (12,617). Deuteron-induced and proton-induced cross section libraries are also included, and can be used with EASY to enable calculations of the activation due to deuterons and proton [2].« less

  4. Optimization of a coupling scheme between MCNP5 and SUBCHANFLOW for high fidelity modeling of LWR reactors

    SciTech Connect

    Ivanov, A.; Sanchez, V.; Imke, U.; Ivanov, K.

    2012-07-01

    In order to increase the accuracy and the degree of spatial resolution of core design studies, coupled Three-Dimensional (3D) neutronics (deterministic and Monte Carlo) and 3D thermal hydraulics (CFD and sub-channel) codes are being developed worldwide. In this paper the optimization of a coupling between MCNP5 code and an in-house development thermal-hydraulics code SUBCHANFLOW is presented. Various improvements of the coupling methodology are presented. With the help of novel interpolation tool a consistent methodology for the preparation of thermal scattering data library have been developed, ensuring that inelastic scattering from bound nuclei is treated at the correct moderator temperature. Trough the utilization of a hybrid coupling with discrete energy Monte-Carlo code KENO a methodology for acceleration of the coupled calculation is being demonstrated. In this approach an additional coupling between KENO and SUBCHANFLOW was developed, the converged results of which are used as initial conditions for the MCNP-SUBCHANFLOW coupling. Acceleration of fission source distribution convergence, by sampling fission source distribution from the power distribution obtained by KENO is also demonstrated. (authors)

  5. High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems

    PubMed Central

    Mahadevan, Vijay S.; Merzari, Elia; Tautges, Timothy; Jain, Rajeev; Obabko, Aleksandr; Smith, Michael; Fischer, Paul

    2014-01-01

    An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework. PMID:24982250

  6. High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems.

    PubMed

    Mahadevan, Vijay S; Merzari, Elia; Tautges, Timothy; Jain, Rajeev; Obabko, Aleksandr; Smith, Michael; Fischer, Paul

    2014-08-01

    An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework. PMID:24982250

  7. Research on Fast-Doppler-Broadening of neutron cross sections

    SciTech Connect

    Li, S.; Wang, K.; Yu, G.

    2012-07-01

    A Fast-Doppler-Broadening method is developed in this work to broaden Continuous Energy neutron cross-sections for Monte Carlo calculations. Gauss integration algorithm and parallel computing are implemented in this method, which is unprecedented in the history of cross section processing. Compared to the traditional code (NJOY, SIGMA1, etc.), the new Fast-Doppler-Broadening method shows a remarkable speedup with keeping accuracy. The purpose of using Gauss integration is to avoid complex derivation of traditional broadening formula and heavy load of computing complementary error function that slows down the Doppler broadening process. The OpenMP environment is utilized in parallel computing which can take full advantage of modern multi-processor computers. Combination of the two can reduce processing time of main actinides (such as {sup 238}U, {sup 235}U) to an order of magnitude of 1{approx}2 seconds. This new method is fast enough to be applied to Online Doppler broadening. It can be combined or coupled with Monte Carlo transport code to solve temperature dependent problems and neutronics-thermal hydraulics coupled scheme which is a big challenge for the conventional NJOY-MCNP system. Examples are shown to determine the efficiency and relative errors compared with the NJOY results. A Godiva Benchmark is also used in order to test the ACE libraries produced by the new method. (authors)

  8. New precision measurements of free neutron beta decay with cold neutrons

    SciTech Connect

    Baeßler, Stefan; Bowman, James David; Penttilä, Seppo I.; Počanić, Dinko

    2014-10-14

    Precision measurements in free neutron beta decay serve to determine the coupling constants of beta decay, and offer several stringent tests of the standard model. This study describes the free neutron beta decay program planned for the Fundamental Physics Beamline at the Spallation Neutron Source at Oak Ridge National Laboratory, and finally puts it into the context of other recent and planned measurements of neutron beta decay observables.

  9. DIRECTIONAL DETECTION OF A NEUTRON SOURCE.

    SciTech Connect

    VANIER, P.E.; FORMAN, L.

    2006-10-23

    Advantages afforded by the development of new directional neutron detectors and imagers are discussed. Thermal neutrons have mean free paths in air of about 20 meters, and can be effectively imaged using coded apertures. Fission spectrum neutrons have ranges greater than 100 meters, and carry enough energy to scatter at least twice in multilayer detectors which can yield both directional and spectral information. Such strategies allow better discrimination between a localized spontaneous fission source and the low, but fluctuating, level of background neutrons generated by cosmic rays. A coded aperture thermal neutron imager will be discussed as well as a proton-recoil double-scatter fast-neutron directional detector with time-of-flight energy discrimination.

  10. Application of a Java-based, univel geometry, neutral particle Monte Carlo code to the searchlight problem

    SciTech Connect

    Charles A. Wemple; Joshua J. Cogliati

    2005-04-01

    A univel geometry, neutral particle Monte Carlo transport code, written entirely in the Java programming language, is under development for medical radiotherapy applications. The code uses ENDF-VI based continuous energy cross section data in a flexible XML format. Full neutron-photon coupling, including detailed photon production and photonuclear reactions, is included. Charged particle equilibrium is assumed within the patient model so that detailed transport of electrons produced by photon interactions may be neglected. External beam and internal distributed source descriptions for mixed neutron-photon sources are allowed. Flux and dose tallies are performed on a univel basis. A four-tap, shift-register-sequence random number generator is used. Initial verification and validation testing of the basic neutron transport routines is underway. The searchlight problem was chosen as a suitable first application because of the simplicity of the physical model. Results show excellent agreement with analytic solutions. Computation times for similar numbers of histories are comparable to other neutron MC codes written in C and FORTRAN.

  11. Coupling a Reactive Transport Code with a Global Land Surface Model for Mechanistic Biogeochemistry Representation: 1. Addressing the Challenge of Nonnegativity

    DOE PAGESBeta

    Tang, Guoping; Yuan, Fengming; Bisht, Gautam; Hammond, Glenn E.; Lichtner, Peter C.; Collier, Nathaniel O.; Kumar, Jitendra; Mills, Richard T.; Xu, Xiaofeng; Andre, Ben; et al

    2016-01-01

    Reactive transport codes (e.g., PFLOTRAN) are increasingly used to improve the representation of biogeochemical processes in terrestrial ecosystem models (e.g., the Community Land Model, CLM). As CLM and PFLOTRAN use explicit and implicit time stepping, implementation of CLM biogeochemical reactions in PFLOTRAN can result in negative concentration, which is not physical and can cause numerical instability and errors. The objective of this work is to address the nonnegativity challenge to obtain accurate, efficient, and robust solutions. We illustrate the implementation of a reaction network with the CLM-CN decomposition, nitrification, denitrification, and plant nitrogen uptake reactions and test the implementation atmore » arctic, temperate, and tropical sites. We examine use of scaling back the update during each iteration (SU), log transformation (LT), and downregulating the reaction rate to account for reactant availability limitation to enforce nonnegativity. Both SU and LT guarantee nonnegativity but with implications. When a very small scaling factor occurs due to either consumption or numerical overshoot, and the iterations are deemed converged because of too small an update, SU can introduce excessive numerical error. LT involves multiplication of the Jacobian matrix by the concentration vector, which increases the condition number, decreases the time step size, and increases the computational cost. Neither SU nor SE prevents zero concentration. When the concentration is close to machine precision or 0, a small positive update stops all reactions for SU, and LT can fail due to a singular Jacobian matrix. The consumption rate has to be downregulated such that the solution to the mathematical representation is positive. A first-order rate downregulates consumption and is nonnegative, and adding a residual concentration makes it positive. For zero-order rate or when the reaction rate is not a function of a reactant, representing the availability limitation

  12. Coupling a Reactive Transport Code with a Global Land Surface Model for Mechanistic Biogeochemistry Representation: 1. Addressing the Challenge of Nonnegativity

    SciTech Connect

    Tang, Guoping; Yuan, Fengming; Bisht, Gautam; Hammond, Glenn E.; Lichtner, Peter C.; Collier, Nathaniel O.; Kumar, Jitendra; Mills, Richard T.; Xu, Xiaofeng; Andre, Ben; Hoffman, Forrest M.; Painter, Scott L.; Thornton, Peter E.

    2016-01-01

    Reactive transport codes (e.g., PFLOTRAN) are increasingly used to improve the representation of biogeochemical processes in terrestrial ecosystem models (e.g., the Community Land Model, CLM). As CLM and PFLOTRAN use explicit and implicit time stepping, implementation of CLM biogeochemical reactions in PFLOTRAN can result in negative concentration, which is not physical and can cause numerical instability and errors. The objective of this work is to address the nonnegativity challenge to obtain accurate, efficient, and robust solutions. We illustrate the implementation of a reaction network with the CLM-CN decomposition, nitrification, denitrification, and plant nitrogen uptake reactions and test the implementation at arctic, temperate, and tropical sites. We examine use of scaling back the update during each iteration (SU), log transformation (LT), and downregulating the reaction rate to account for reactant availability limitation to enforce nonnegativity. Both SU and LT guarantee nonnegativity but with implications. When a very small scaling factor occurs due to either consumption or numerical overshoot, and the iterations are deemed converged because of too small an update, SU can introduce excessive numerical error. LT involves multiplication of the Jacobian matrix by the concentration vector, which increases the condition number, decreases the time step size, and increases the computational cost. Neither SU nor SE prevents zero concentration. When the concentration is close to machine precision or 0, a small positive update stops all reactions for SU, and LT can fail due to a singular Jacobian matrix. The consumption rate has to be downregulated such that the solution to the mathematical representation is positive. A first-order rate downregulates consumption and is nonnegative, and adding a residual concentration makes it positive. For zero-order rate or when the reaction rate is not a function of a reactant, representing the availability limitation of each

  13. Measurement of Neutron Yields from UF4

    SciTech Connect

    Bell, Zane W; Ziock, Klaus-Peter; Ohmes, Martin F; Xu, Yunlin; Downar, Thomas J; Pozzi, Sara A

    2010-01-01

    We have performed measurements of neutron production from UF{sub 4} samples using liquid scintillator as the detector material. Neutrons and gamma rays were separated by a multichannel digital pulse shape discriminator, and the neutron pulse-height spectra were unfolded using sequential least-squares optimization with an active set strategy. The unfolded spectra were compared to estimates calculated with the SOURCES 4C code.

  14. Toward a New Evaluation of Neutron Standards

    NASA Astrophysics Data System (ADS)

    Carlson, A. D.; Pronyaev, V. G.; Capote, R.; Hale, G. M.; Hambsch, F.-J.; Kawano, T.; Kunieda, S.; Mannhart, W.; Nelson, R. O.; Neudecker, D.; Schillebeeckx, P.; Simakov, S.; Smith, D. L.; Talou, P.; Tao, X.; Wallner, A.; Wang, W.

    2016-02-01

    Measurements related to neutron cross section standards and certain prompt neutron fission spectra are being evaluated. In addition to the standard cross sections, investigations of reference data that are not as well known as the standards are being considered. Procedures and codes for performing this work are discussed. A number of libraries will use the results of this standards evaluation for new versions of their libraries. Most of these data have applications in neutron dosimetry.

  15. Neutron dosimetry in solid water phantom

    SciTech Connect

    Benites-Rengifo, Jorge Luis; Vega-Carrillo, Hector Rene

    2014-11-07

    The neutron spectra, the Kerma and the absorbed dose due to neutrons were estimated along the incoming beam in a solid water phantom. Calculations were carried out with the MCNP5 code, where the bunker, the phantom and the model of the15 MV LINAC head were modeled. As the incoming beam goes into the phantom the neutron spectrum is modified and the dosimetric values are reduced.

  16. Neutron Transport Characteristics of a Nuclear Reactor Based Dynamic Neutron Imaging System

    SciTech Connect

    Khaial, Anas M.; Harvel, Glenn D.; Chang, Jen-Shih

    2006-07-01

    An advanced dynamic neutron imaging system has been constructed in the McMaster Nuclear Reactor (MNR) for nondestructive testing and multi-phase flow studies in energy and environmental applications. A high quality neutron beam is required with a thermal neutron flux greater than 5.0 x 10{sup 6} n/cm{sup 2}-s and a collimation ratio of 120 at image plane to promote high-speed neutron imaging up to 2000 frames per second. Neutron source strength and neutron transport have been experimentally and numerically investigated. Neutron source strength at the beam tube entrance was evaluated experimentally by measuring the thermal and fast neutron fluxes, and simple analytical neutron transport calculations were performed based upon these measured neutron fluxes to predict facility components in accordance with high-speed dynamic neutron imaging and operation safety requirements. Monte-Carlo simulations (using MCNP-4B code) with multiple neutron energy groups have also been used to validate neutron beam parameters and to ensure shielding capabilities of facility shutter and cave walls. Neutron flux distributions at the image plane and the neutron beam characteristics were experimentally measured by irradiating a two-dimensional array of Copper foils and using a real-time neutron radiography system. The neutron image characteristics -- such as neutron flux, image size, beam quality -- measured experimentally and predicted numerically for beam tube, beam shutter and radiography cave are compared and discussed in detail in this paper. The experimental results show that thermal neutron flux at image plane is nearly uniform over an imaging area of 20.0-cm diameter and its magnitude ranges from 8.0 x 10{sup 6} - 1.0 x 10{sup 7} n/cm{sup 2}-sec while the neutron-to-gamma ratio is 6.0 x 10{sup 5} n/cm{sup 2}-{mu}Sv. (authors)

  17. The RealGas and RealGasH2O options of the TOUGH+ code for the simulation of coupled fluid and heat flow in tight/shale gas systems

    NASA Astrophysics Data System (ADS)

    Moridis, George J.; Freeman, Craig M.

    2014-04-01

    We developed two new EOS additions to the TOUGH+ family of codes, the RealGasH2O and RealGas. The RealGasH2O EOS option describes the non-isothermal two-phase flow of water and a real gas mixture in gas reservoirs, with a particular focus in ultra-tight (such as tight-sand and shale gas) reservoirs. The gas mixture is treated as either a single-pseudo-component having a fixed composition, or as a multicomponent system composed of up to 9 individual real gases. The RealGas option has the same general capabilities, but does not include water, thus describing a single-phase, dry-gas system. In addition to the standard capabilities of all members of the TOUGH+ family of codes (fully-implicit, compositional simulators using both structured and unstructured grids), the capabilities of the two codes include coupled flow and thermal effects in porous and/or fractured media, real gas behavior, inertial (Klinkenberg) effects, full micro-flow treatment, Darcy and non-Darcy flow through the matrix and fractures of fractured media, single- and multi-component gas sorption onto the grains of the porous media following several isotherm options, discrete and fracture representation, complex matrix-fracture relationships, and porosity-permeability dependence on pressure changes. The two options allow the study of flow and transport of fluids and heat over a wide range of time frames and spatial scales not only in gas reservoirs, but also in problems of geologic storage of greenhouse gas mixtures, and of geothermal reservoirs with multi-component condensable (H2O and CH4) and non-condensable gas mixtures. The codes are verified against available analytical and semi-analytical solutions. Their capabilities are demonstrated in a series of problems of increasing complexity, ranging from isothermal flow in simpler 1D and 2D conventional gas reservoirs, to non-isothermal gas flow in 3D fractured shale gas reservoirs involving 4 types of fractures, micro-flow, non-Darcy flow and gas

  18. The RealGas and RealGasH2O Options of the TOUGH+ Code for the Simulation of Coupled Fluid and Heat Flow in Tight/Shale Gas Systems

    SciTech Connect

    Moridis, George; Freeman, Craig

    2013-09-30

    We developed two new EOS additions to the TOUGH+ family of codes, the RealGasH2O and RealGas . The RealGasH2O EOS option describes the non-isothermal two-phase flow of water and a real gas mixture in gas reservoirs, with a particular focus in ultra-tight (such as tight-sand and shale gas) reservoirs. The gas mixture is treated as either a single-pseudo-component having a fixed composition, or as a multicomponent system composed of up to 9 individual real gases. The RealGas option has the same general capabilities, but does not include water, thus describing a single-phase, dry-gas system. In addition to the standard capabilities of all members of the TOUGH+ family of codes (fully-implicit, compositional simulators using both structured and unstructured grids), the capabilities of the two codes include: coupled flow and thermal effects in porous and/or fractured media, real gas behavior, inertial (Klinkenberg) effects, full micro-flow treatment, Darcy and non-Darcy flow through the matrix and fractures of fractured media, single- and multi-component gas sorption onto the grains of the porous media following several isotherm options, discrete and fracture representation, complex matrix-fracture relationships, and porosity-permeability dependence on pressure changes. The two options allow the study of flow and transport of fluids and heat over a wide range of time frames and spatial scales not only in gas reservoirs, but also in problems of geologic storage of greenhouse gas mixtures, and of geothermal reservoirs with multi-component condensable (H2O and CH4) and non-condensable gas mixtures. The codes are verified against available analytical and semi-analytical solutions. Their capabilities are demonstrated in a series of problems of increasing complexity, ranging from isothermal flow in simpler 1D and 2D conventional gas reservoirs, to non-isothermal gas flow in 3D fractured shale gas reservoirs involving 4 types of fractures, micro-flow, non-Darcy flow and gas

  19. Simulation of the neutron spectrum from the 7Li(p,n) reaction with a liquid-lithium target at Soreq Applied Research Accelerator Facility

    NASA Astrophysics Data System (ADS)

    Friedman, M.; Cohen, D.; Paul, M.; Berkovits, D.; Eisen, Y.; Feinberg, G.; Giorginis, G.; Halfon, S.; Krása, A.; Plompen, A. J. M.; Shor, A.

    2013-01-01

    The 7Li(p,n)7Be reaction has been used for the last 25 years to produce quasi-Maxwellian neutrons in order to measure Maxwellian-Averaged Cross-Sections in the relevant temperatures for stellar nucleosynthesis. A liquid-lithium target at the Soreq Applied Research Accelerator Facility is expected to allow us to perform such measurements at higher neutron intensities. Here we describe a Monte Carlo tool, SimLiT, developed to evaluate neutron spectra, intensities and angular distributions resulting from this reaction. We also demonstrate the feasibility to couple SimLiT with an advanced transport code, resulting in a powerful tool for planning and analysis of experiments using the 7Li(p,n) reaction as a neutron source.

  20. Monte Carlo Neutronics and Thermal Hydraulics Analysis of Reactor Cores with Multilevel Grids

    NASA Astrophysics Data System (ADS)

    Bernnat, W.; Mattes, M.; Guilliard, N.; Lapins, J.; Zwermann, W.; Pasichnyk, I.; Velkov, K.

    2014-06-01

    Power reactors are composed of assemblies with fuel pin lattices or other repeated structures with several grid levels, which can be modeled in detail by Monte Carlo neutronics codes such as MCNP6 using corresponding lattice options, even for large cores. Except for fresh cores at beginning of life, there is a varying material distribution due to burnup in the different fuel pins. Additionally, for power states the fuel and moderator temperatures and moderator densities vary according to the power distribution and cooling conditions. Therefore, a coupling of the neutronics code with a thermal hydraulics code is necessary. Depending on the level of detail of the analysis, a very large number of cells with different materials and temperatures must be regarded. The assignment of different material properties to all elements of a multilevel grid is very elaborate and may exceed program limits if the standard input procedure is used. Therefore, an internal assignment is used which overrides uniform input parameters. The temperature dependency of continuous energy cross sections, probability tables for the unresolved resonance region and thermal neutron scattering laws is taken into account by interpolation, requiring only a limited number of data sets generated for different temperatures. The method is applied with MCNP6 and proven for several full core reactor models. For the coupling of MCNP6 with thermal hydraulics appropriate interfaces were developed for the GRS system code ATHLET for liquid coolant and the IKE thermal hydraulics code ATTICA-3D for gaseous coolant. Examples will be shown for different applications for PWRs with square and hexagonal lattices, fast reactors (SFR) with hexagonal lattices and HTRs with pebble bed and prismatic lattices.